ML20078G750
| ML20078G750 | |
| Person / Time | |
|---|---|
| Issue date: | 09/30/1983 |
| From: | NRC OFFICE FOR ANALYSIS & EVALUATION OF OPERATIONAL DATA (AEOD) |
| To: | |
| References | |
| NUREG-0090, NUREG-0090-V06-N01, NUREG-90, NUREG-90-V6-N1, NUDOCS 8310130133 | |
| Download: ML20078G750 (49) | |
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l NUREG-0090 Vol. 6, No.1 l
Report to Congress on Abnormal Occurrences January - March 1983 Date Published: September 1983 Office for Analysis and Evaluation of Operational Data U.S. Nuclear Regulatory Commission Washington, D.C. 20555 y
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U.S. NUCLEAR REGULATORY COMMISSION I
Previous Reports in Series for the Report to Congress on Abnormal Occurrences NUREG 75/090, January-June 1975, NUREG-0090, Vol.2, No. 2, April-June 1979, published October 1975 ^
published November 1979 NUREG-0090-1,. July-September 1975, NUREG-0090, Vol.2, No.3, July-September 1979, published March 1976 published February 1980 NUREG-0090-2, October-December 1975, NUREG-0090, Vol.2, No.4, OctobenDecember 1979 published March 1976 published April 1980 j
NUREG-0090-3, January-March 1976, NUREG-0990, Vol. 3, No.1, January-March 1980, published July 1976 published September 1980 NUREG-0090-4, April-June 1976, NUREG-0090, Vol. 3, No. 2, April-June 1980, published March 1977 published November 1980 NUREG-0090-5, July-September 1976, NUREG-0090, Vol. 3, No. 3,, July-September 1980, published March 1977 published February 1981 NUREG-0090-6, October-December 1976, NUREG-0090, Vol.3, No.4, October-December 1980, published June 1977 published May 1981 NUREG-0090-7, Januard-March 1977, NUREG-0090, Vol.4, No.1, January-March 1981, published June 1977 published July 1981 NUREG-0090-8, April-June 1977,
. NUREG-0090, Vol. 4, No. 2, April-June 1981,
. published September 1977 published October 1981 NUREG-0090-9, July-September 1977, RUREG-0090, Vol.4, No.3, July-September 1981, published November 1977 published January 1982 NUREG-0090-10, October-December 1977, NUREG-0090, Vol.4, No.4, October-December 1981,,
published March 1978 published May 1982 NUREG-0090,- Vol.1, No.1, Januarg-March 1978, NUREG-0090, Vol.5, No.1, January-March 1982, published June 1978 published August 1982 NUREG-0090, Vol.1, No. 2, April-June 1978, NUREG-0090, Vol.5, No.2, April-Jime 1982, published September 1978 published December 1982 NUREG-0090, Vol.1, No. 3, July-September 1978, NUREG-0090, Vol.5, No.3, July-Septcmber 1982, published December 1978 published January 1983 NUREG-0090, Vol.1, No.4, October-December 1978, NUREG-0090, Vol.S, No.4, October-Deccmber 1982, publishad March 1979 published May 1983
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NUREG-0090, Vol. 2, No.
1, January-March 1979, published July 1979 i,
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7 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 January 1 to March 31, 1983.
The report states that for this report period, there were three abnormal occurrences at the nuclear power plants licensed by the NRC to operate.
The first involved a main feedwater line break due to water hammer.
The second involved management and procedural control deficiencies.
The third involved failure of the automatic reactor trip system.
There were no abnormal occur-rences for the other NRC licensees.
There were six abnormal occurrences at Agreement State licensees.
One involved an individual who ingested and was contaminated by radioactive material.
Four involved lost or stolen radio-active sources.
One involved radioactive contamination of a metals production facility.
The report also contains information updating some previously reported abnormal occurrences.
iii
7 5
Y r T F
f CONTENTS Page iii ABSTRACT vii PREFACE................
vii INTRODUCTION.
vii THE REGULATORY SYSTEM viii REPORTABLE OCCURRENCES.........
ix AGREEMENT STATES..
FOREIGN INFORMATION................
ix REPORT TO CONGRESS ON ABNORMAL OCCURRENCES, JANUARY-MARCH 1983.
1 1
NUCLEAR POWER PLANTS......
83-1 Main Feedwater Line Break Due to Water Hammer....
1 83-2 Deficiencies in Management and Procedural 4
Controls.......
83-3 Failure of Automatic Reactor Trip System....
7 FUEL CYCLE FACILITIES (Other than Nuclear Power Plants).....
15 OTHER NRC LICENSEES (Industrial Radiographers, Medical Institutions, Industrial Users, Etc.)..
15 AGREEMENT STATE LICENSEES...............
15 AS83-1 Contamination By and Ingestion of 15 Radioactive Material.
AS83-2 Lost Radioactive Source........
17 18 AS83-3 Stolen Radioactive Source....
19 AS83-4 Lost Radioactive Source....
20 AS83-5 Stolen Radioactive Source...
AS83-6 Radioactive Contamination of a Metals 21 Production Facility..
REFERENCES........................
25 APPENDIX A - ABNORMAL OCCURRENCE CRITERIA........
27 APPENDIX B - UPDATE OF PREVIOUSLY REPORTED ABNORMAL OCCURRENCES..
31 NUCLEAR POWER PLANTS.............................
31 75-5 Cracks in Pipes at Boiling Water Reactors 31 (BWRs).......
77-9 Environmental Qualification of Safety-Related Electrical Equipment Inside Containment.
32 32 79-3 Nuclear Accident at Three Mile Island..........
OTHER NRC LICENSEES....
35 82-6 Radiological Contamination From Well Logging Operations...........
35 APPENDIX C - OTHER EVENTS OF INTEREST......
37 43 REFERENCES (FOR APPENDICES)................................
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f PREFACE INTRODUCTION i
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The Nuclear Regulatory Commission reports to the Congress each quarter under provisions of Section 208 of the Energy Reorganization Act of 1974 on any abnormal occurrences involving facilities and activities regulated by the NRC.
An abnormal occurrence is defined in Section 208 as an unscheduled incident or event which the Commission determines is significant from the standpoint of public health or safety.
Events are currently identified as abnormal occurrences for this report by the NRC using the criteria delineated in Appendix A.
These criteria were promul-gated in an NRC policy statement which was published in the Federal Register on February 24, 1977 (Vol. 42, No. 37, pages 10950-10952).
In order to provide wide dissemination of information to the public, a Federal Register notice is issued on each abnormal occurrence with copies distributed to the NRC Public Document Room and all local public document rooms.
At a minimum, each such notice contains the date and place of the occurrence and describes its nature ano probable consequences.
The NRC has reviewed Licensee Event Reports, licensing and enforcement actions (e.g., notices of violations, civil penalties, license modifications, etc.),
generic issues, significant inventory differences involving special nuclear material, and other categories of information available to the NRC.
The NRC has determined that only those events, including those submitted by the Agree-ment States, described in this report meet the criteria for abnormal occurrence reporting.
This report covers the period between January 1 to March 31, 1983.
Information reported on each event includes:
date and place; nature and probable 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 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 precess and the protcetion 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 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 vii
y licensed by NRC.
An inspection and enforcement program helps assure compliance with the regulations.
Requirements for reporting incidents or events exist which help identify deficiencies early and aid in assuring that corrective 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 incident.
4 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 j
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.
l 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 informa-i tion 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 viii
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 i
safety.
Many are discovered during routine inspection and surveillance testing
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and are corrected upcn discovery.
Typically, they concern single malfunctions l
of components or parts of systems, with redundant cperable components or i
systems continuing to be available to perform the design function.
Information concerning reportable occurrences at facilities licensed or other-wise 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 document rooms, special notifications to licensees and other affected or interested groups, and public announcements.
In addition, a computer printout containing information on reportable 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 relinquishes and the States assume regulatory authority over byproduct, souice and special nuclear materials (in quantities not capable of sustaining a chain reaction).
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.
NRC prepares a semiannual summary of this and other information in a document entitled, " Licensing Statistics and Other Data," which is pub-licly 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.
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 occurrerices are reported.
ix
REPORT TO CONGRESS ON ABNORMAL OCCURRENCES JANUARY-MARCH 1983 NUCLEAR POWER PLANTS The NRC is reviewing events reported at the nuclear power plants licensed to operate during the first calendar quarter of 1983.
As of the date of this report, the NRC had determined that the following were abnormal occurrences.
83-1 Main Feedwater Line Break Due To Water Hammer The following information pertaining to this event is also being reported in the Federal Register (Ref. 1).
Appendix A (Example 10 of "For All Licensees")
of this report notes that a major deficiency in design having safety implica-tions requiring immediate remedial action can be considered an abnormal occurrence.
Date and Place - On January 25, 1983, the Maine Yankee Nuclear Power Plant experienced a reactor trip followed by loss of main feedwater and automatic initiation of the auxiliary feedwater system.
The auxiliary feedwater initia-tion resulted in a water hammer transient in the feedwater lines for two of the three steam generators with a resultant feedwater pipe rupture.
The Maine Yankee plant is a Combustion Engineering designed pressurized water reactor (PWR), operated by Maine Yankee Atomic Power Company (the licensee), and located in Lincoln County, Maine.
Nature and Probable Consequences - The Maine Yankee plant uses a three loop reactor coolant system design.
Each loop contains a steam generator (SG), a reactor coolant pump and associated connecting piping.
The steam generator employs a vertical U-tube design.
Feedwater is normally provided to the secondary side of each steam generator by operating one of two motor-driven main feedwater pumps when the plant power level is less than 50%, or a turbine-driven feedwater pump for power levels greater than 50%.
(Note:
This feed-water arrangement is a new configuration.
Until recently, the plant did not have a turbine-driven feedwater pump; main feedwater was provided only by motor-driven pumps for all power levels.) If all sources of main feedwater are lost, an auxiliary feedwater system is available to supply the three SGs using portions of the normal feedwater lines.
On January 25, 1983, the plant was operating near 100% power in its new configu-ration, with only its new turbine-driven feedwater pump supplying feedwater.
Both motor-driven feedwater pumps were out of service for maintenance.
While operators were attempting to is'olate an electrical ground in the control rod drive systems, a reactor trip occurred.
As a result, both the main turbine and the turbine-driven feedwater pump tripped.
This resulted in a complete loss of normal feedwater flow, followed by a normal reduction in SG water level associated with reactor trips. When the SG water level reached approximately 30% (on the narrew range indicator), auxiliary feedwater flow automatically initiated, as designed.
Approximately 15 minutes after the trip, a loud noise was heard in the plant machine shop.
The shop is just below the main feedwater lines.
Additionally, a containment fire detector (temperature sensitive) alarmed and containment humidity began to rise.
The containment was entered for inspection and the feedwater line for SG No. 2 was found to be leaking severely near the inlet nozzle.
The leak rate was estimated to be a maximum of 100 gpm.
Feedwater flow to SG No. 2 was terminated and its level maintained by inter-connecting the tube sheet drains for all three SGs.
Normal station cooldown was initiated to facilitate inspection and repair.
Reactor coolant system parameters were stable and within normal ranges for this operating condition.
During the cooldown period of about twelve hours, leakage continued through the feedwater line.
This spillage was removed and disposed of through the containment auxiliary sump drain system.
Subsequent inspection of the SG No. 2 feedwater piping showed that the through-wall crack had occurred in the pipe adjacent to the weld joining the pipe and SG safe end.
It is believed that the water hammer had caused an existing crack to propagate on through the pipe wall.
The location of the crack coincided with a stress area in the pipe where previous PWR experience had identified the likelihood of thermal stress cracking (Ref. 2).
Radiographic examinations showed similar cracking had also begun on the SG No. 1 and No. 3 feedwater pipes.
Further examinations showed various degrees of damage to some snubbers and other supports for the SG No. 2 and No. 3 feedwater lines, a feed ring support for SG No. 3, and the safe ends and nozzles (including distorted thermal sleeves) for SG No. 2 and No. 3.
Cause or Causes - The cause of the event is attributed to incomplete considera-tion, in ongoing design and operational plant upgrading, of previous generic safety concerns related to steam generator water hammer and feedline thermal stress cracking.
The installation of a steam turbine-driven main feed pump and automatic initiation of the cold water auxiliary feedwater system without the addition of SG J-tubes and operational procedures to alleviate these concerns increased the potential for feedwater piping thermal shock and water hammer at Maine Yankee.
The water hammer probably occurred when the outlet nozzle at the bottom of the SG feed ring became submerged in the rising SG water level and the steam in contact with the cold feedwater within the ring suddenly collapsed.
Before the design and operational changes were made, the plant had apparently not experienced any water hammers.
Review of plant operating history indicated that all full load trips were followed by continuous main feedwater flow to the SGs via the bypass valves.
(These valves bypass the normal feedwater regulating valves at low flow rates.) This feedwater was provided by the motor-driven main feedwater pumps that continued to operate, feeding about 5% of full flow to each SG through the bypass valves, following the trip.
Hence, warm feedwater was supplied, minimizing the potential for thermal shock and water hammer.
In the two prior trips in which main feedwater was lost due to loss of electric power, power levels were below 50%, so SG level shrinkage was less.
Also, the auxiliary feedwater system was manually started and the flow rate controlled by the plant operators.
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In the January 25, 1983 event, the trip was from full power and since the turbine-driven feedwater pump tripped, all warn feedwater was lost.
Subse-l quently, auxiliary feedwater automatically initiated.
However, the auxiliary feedwater was drawn from the demineralized water storage tank (DWST) at 60 F.
By comparison, normal feedwater temperature is about 440 F.
The extreme tem-perature differential between the normal and auxiliary feedwater can cause two problems.
First, it can cause potentially high thermal stresses and cracking in the feedwater piping.
Second, it may rapidly condense any steam in the feed lines, leading to a higher possibility of water hammer in the feed ring and feedwater line, i
Actions Taken to Prevent Recurrence Licensee - Repairs were made to components damaged by this event.
This included replacing cracked (and/or broken) feedwater piping, and the replacement or repair of damaged piping supports and SG internals.
A design change was implemented adding J-tubes to the top of the SG feed rings.
This change increases the area for pressure equalization.
It also reduces the rate at which the feed rings drain when the SG water level drops below the feed ring after a trip; thus, on early initiation of auxiliary feedwater flow, the feedwater line and feed ring are expected to remain full.
The design change has been used at other plants, such as Millstone Unit 2 and Calvert Cliffs Units 1 and 2, to reduce the possibility of water hammer events.
A number of operational changes also were made, for various niodes of plant operation, to reduce the differential in temperature between the main feedwater and the auxiliary feedwater.
These design and operational changes minimize the potential for future water hammer; some of the changes will also minimize the thermal cycling of the feed-water lines and steam generator feedwater nozzles, and minimize the potential for any future thermal fatigue failures.
The licensee conducted a series of tests to verify the integrity of the feed-water lines, and to verify that the design and operational changes were effec-tive.
The licensee reported that the tests were successful.
The licensee also plans to perform a longer range program to determine what further changes should be made to minimize the potential for recurrence of the event.
NRC - The NRC held meetings with the licensee and by means of inspections and analyses evaluated the licensee's corrective actions.
The NRC prepared a safety evaluation report, dated March 18, 1983 (Ref. 3), justifying the plant's return to power operation.
This incident is closed for purposes of this report.
3
83-2 Deficiencies in Management and Procedural Controls I
The following information pertaining to this event is also being reported in the Federal Register (Ref. 4).
Appendix A (Example 11 of "For All Licensees")
of this report notes that serious deficiency in management or procedural con-trolc in major areas can be considered an abnormal occurrence.
Date and Place - On February 18, 1983, the NRC issued a notice of violation and proposed imposition of civil penalties (for $600,000) to Caroline Power and Light Company, licensee for Brunswick Units 1 and 2 (Ref. 5).
The action was based on violations involving technical specification surveillance requirements.
Brunswick Units 1 and 2 are General Electric designed boiling water reactors located in Brunswick County, North Carolina.
Nature and Probable Consequences - Inspection findings indicated that the Brunswick facility had been oparated, in some cases since the issuance of the operating licenses (December 1974 for Unit 2 and September 1976 for Unit 1),
without certain surveillance procedures and verification by surveillance test-ing of a number of safety systems and components.
In addition, it was found that the licensee's quality assurance program failed to correct the problem once the lack of one of the surveillance procedures was identified.
Even though testing performed subsequent to the identification of the missed sur-veillances demonstrated the affected equipment to be operable, the deficiencies were of serious safety concern, i.e.,
(a) the facility had been operated for an extended period of time without the necessary assurance that the equipment would function properly if called upon, and (b) the violations, when viewed collectively, and in light of later identified examples of t;ilures to meet limiting conditions for operation and surveillance requirement., suggested a programmec failure that unless corrected could lead to more serious events.
A summary of the history leading to the February 18, 1983 NRC action follows.
On June 28, 1982 Unit 1 reactor lost voltage to certain emergency electrical busses and tripped.
It was returned to power on June 29.
The licensee's post-trip evaluation of the event revealed that certain relays associated with the energency electrical busses of Units 1 and 2, although they functioned properly, had not been tested or calibrated as required by technical specifications.
Action statements for the relevant limiting conditions for operation required shutdown of Unit 1 (Unit 2 was shut down for refueling) until test and calibra-tion of the relays were accomplished.
On June 30, the licensee recuested and was granted NRC approval for continued operation of Unit 1 while the required tests and calibrations were being per-formed.
On July 2, NRC Region II issued a Confirmation of Action Letter (Ref. 6) confirming the licensee's commitment to review all technical specifi-cation surveillance requirements and the administrative control system for assuring that surveillance requirements were met.
On July 15, NRC Region II was informed that the licensee review of technical specification surveillance requirements had revealed additional missed surveil-lance requirements that were not covered by procedures.
The tests involved limiting conditions for operation and required implementation of action state-ments for continued operation of Unit 1.
Upon discovery of these missed sur-veillances, the licensee wrote the necessary procedures and conducted the 4
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i required tests.
The test results showed that the equipment would have func-I tioned if called upon.
On July 16, NRC inspectors informed NRC Region II and Brunswick management that
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containment leakage tests of certain penetrations and valves had not been con-I ducted at the required frequency.
Although the licensee had implemented appro-priate procedure changes on Unit 2 in June 1982, no procedme change had been implemented for Unit I which had similar requirements.
On receipt of this information, Unit 1 was shut down.
The NRC Region II Administrator and the Executive Vice President of the licensee discussed the situation by telephone.
l It was agreed that neither unit would be operated until the licensee had com-pleted a comprehensive review of technical specification surveillance require-i ments, corrected such violations as might be disclosed by the review, identified the root causes of the violations, and presented the Commission a proposed revision of its management control program to prevent recurrence of similar violations.
On July 20, NRC Region II issued a Confirmation of Action Letter (Ref. 7) detailing broad commitments made by the licensee in several previous tele-communcations.
The letter covered certain specific assignments of review responsibility for the corporate Nuclear Safety and corporate Quality Assurance staffs, implementation of an extensive training program, assignment of a full-time operationally qualified corporate representative on site, assignment of a special corporate panel to review the adequacy of corrective actions which the licensee had committed to take, and formal notification of NRC Region II prior to resumption of Unit 1 or Unit 2 power operation.
On August 24 an enforcement conference was held at the NRC Region II office.
The NRC Region II Administrator reviewed NRC inspection findings relating to facts disclosed since June 28, expressed NRC concerns about the failure of corporate and facility management controls to prevent the violations indicated by the findings, and asked the licensee what actions had been taken or were planned to reestablish satisfactory management control of licensad activities.
The Senior Vice President of the licensee presented recommendations and con-clusions furnished to the licensee by a panel of senior mcn nement officers from the nuclear power industry, retained by the licensee t review the ade-quacy and completeness of actions taken, and to recommend additional management actions needed to assure future compliance with the Brunswick technical speci-fications.
The Senior Vice President detailed the actions taken or planned to meet each item identifed in Region II Confirmation of Action Letters dated July 2 and July 20.
Beyond the commitments previously made, the licensee described an improvement program involving extensive assignments of corporate and facility staff rcsponsibilities designed to achieve basic improvement in management, operations, and quality assurance performance.
The licensee stated that commitments made during the conference would be incorporated in its improvement program which would be submitted in a comprehensive report to the NRC Region II Administrator by November 1, 1982.
This report was submitted on October 29, 1982 (Ref. 8).
The actions described in the licensee's long-range improvement program were the subject of an NRC Confirmatory Order issued by the Director of the NRC Office of Inspection and Enforcement on December 22, 1982 (Ref. 9).
On that same date another enforcement conference was held in the NRC Region II Office with senior 5
managers of the licensee.
During that meeting, additional events similar in nature to those identified previously and which had recently been revealed at the licensee's Brunswick facility, were discussed.
On January 23, 1983, still another event occurred.
These events heightened the NRC's concerns regarding the safe operation of the licensee's facility.
Cause or Causes - The cause of the violations was attributed to a breakdown in corporate and facility management controls in the areas of corporate oversight, I
tacility management and operations, and problem identification and correction.
The violations, particularly considering the length of time the violations con-tinued undetected and the failure to take action to correct problems that were identified, raised serious concerns about the adequacy of the safety of opera-
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tion of the facility in regard to properly protecting the health and safety of
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the public.
Actions Taken to Prevent Recurrence Licensee - In response to the issues raised by the NRC during 1982, the licensee undertook a self-appraisal program, commencing in July 1982.
In addition to short term corrective actions, a Brunswick Improvement Program was developed and submitted to the NRC on October 29, 1982 (Ref. 8).
The Improvement Program had previously been reviewed by an External Safety Review Panel composed of nuclear power industry professionals; the Panel was chaired by a vice president of the Institute of Nuclear Power Operations (INP0).
The Improvement Program contained seven objectives, together with the associated action items, plans for implementing each action item, and an identification of personnel accounta-bilities.
The seven objectives are as follows.
1.
Ensure full and timely compliance to all surveillance requirements, regulatory commitments, and regulatory requirements.
2.
Ensure that all necessary procedures (including those resulting from plant modifications and new requirements) exist and are clear, unam-biguous, precise, complete, and of high technical quality.
3.
Increase frequency and scope of QC surveillance and corporate auditing
. program activities.
4.
Ensure the maintenance activities do not degrade or render inoperable any component, system, or instrument.
5.
Increase the proficiency of plant personnel by means of expanded training programs.
6.
More effectively utilize the technical expertise of the on-site and corporate nuclear safety personnel in enhancing the safety and reli-
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ability of plant operations.
7.
Undertake actions to enhance and strengthen the management control and organizational discipline necessary to provid-for safe and reli-able operation The details of the long-range Improvement Program and the results of the near-term corrective actions were discussed with the licensee in a meeting at the 6
NRC Region II Office on November 10, 1982, in a meeting at the Brunswick site on January 6, 1983, and in a meeting with the NRC Executive Director for Opera-tions in Bethesda, Maryland on January 19, 1983.
In addition to inauguration of the Improvement Program, other actions taken included, (a) placement of a senior corporate official at the Brunswick site, (b) visiting several other utilities to examine their programs, (c) assuring that lestons learned at Brunswick would also be applied at the licensee's other plants (Robinson and Harris facilities), (d) identifying every technical specification surveillance requirement and assuring that an updated, written procedure exists for each, and (e) establishing a computerized system for monitoring technical specifica-tion compliance.
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Most of the licensee's action items for the Improvement Program were :cheduled for completion during the last quarter of 1982 and first quarter of 1383; some items would not be complete until the last quarter of 1983.
NRC - The NRC has taken numerous actions to convey the NRC's concern for the lack of diligence demonstrated by the licensee in implementation of technical specification requirements, and to assure that the licensee has taken and will continue to take thorough and effective corrective actions.
As discussed pre-viously, these actions included inspections, telephone communications, enforce-ment and other meetings, and Confirmation of Action Letters.
In addition, on February 18, 1983, the NRC issued a letter (Ref. 5) with a Notice of Violation and Proposed Imposition of Civil Penalties (for $600,000) to the licensee.
The letter contained a chronology of the events leading to the enforcement action and again emphasized the NRC's concern with the identi-fied deficiencies.
The violations cited the missed surveillance testing of certain safety systems and the failure to correct the problem once the lack of one of the surveillance procedures was identified.
The enforcen.ent letter also stated that it was vital that effective communica-tions with and between all segments of the licensee's staff be established and that all segments of the operations staff be involved in identifying program-matic deficiences and in developing procedures to remedy those deficiencies.
Accordingly, in response to the Notice of Violation and Proposed Imposition of Civil Penalties, the licensee was directed to describe the efforts taken and to be taken to ensure that effective communcations between management and staff are established and maintained.
The licensee later requested additional time in which to respond.
The NRC subsequently agreed.
The NRC will continue to closely monitor the licensee's actions to ensure that they are thorough and effective.
Further reports will be made as appropriate.
83-3 Fsilure of Automatic Reactor Trip System The following information pertaining to this event is also being reported in the Federal Register (Ref. 10).
Appendix A (Example 10 of "For All Licensee")
notes that a major deficiency ir. design, construction, or operation having 7
\\
l safety implications requiring immediate remedial action can be considered an abnormal occurrence.
Date and Place - On February 22, 1983 and again on February 25, 1983, the Salem Unit 1 reactor control rods failed to insert upon receipt of an auto-matic trip signal from the reactor protection system.
However, the rods did insert and shutdown the plant upon receipt of a manually initiated trip signal.
Salem Unit 1 is a Westinghouse designed, pressurized water nuclear power plant located in Salem County, New Jersey.
The plant is operated by Public Service Electric and Gas Company (the licensee).
Nature and Probable Consequences - Nuclear plants have safety and control systems to limit the consequences of abnormal operating conditions.
"Antici-pated transients" are defined as abnormal operating conditions (e.g., loss of feedwater, loss of off-site power, tripping of the turbine generator), which are likely to occur one or more times during the life of the nuclear power plant.
In some such cases, a rapid shutdown of the nuclear reactor (fast insertion of the control rods into the reactor core - a reactor trip) is an important safety measure to assure that acceptable fuel design limits are not exceeded.
If there were a potentially severe transient, and the reactor shut-down system did not function as designed, then an " anticipated transient with-out scram," or ATWS, would have occurred.
ATWS safety issues have been under study by the AEC, NRC, and the nuclear industry for a number of years.
A pro-posed ATWS technical position and ATWS rule are presently being developed by the NRC.
The reactor protection system (RPS) is a safety-related system which encompasses all electrical and mechanical devices and circuitry (from sensors to actuation device input terminals) involved in generating those signals associated with the protective function.
These signals include those that actuate reactor trip.
The reactor trip system (RTS) is part of the RPS and includes those power sources, sensors, initiation circuits, logic matrices, bypasses, inter-locks, racks, panels and control boards, and actuation and actuated devices, that are required to initiate reactor shutdown.
The RTS is designed to initiate automatically the reactivity control system (control rods) to shut down the reactor, thereby assuring that acceptable fuel design limits are not exceeded, and is designed to fail safe for most internal component failures.
The RTS can also be actuated manually by operator action.
Westinghouse designed plants use two redundant reactor trip breakers (RTBs) in series in the RTS.
For the Salem Unit 1 plant, each RTB includes an under-voltage (UV) trip attachment and a shunt trip attachment to actuate (open) the trip breaker. The UV device initiates a breaker trip when de-energized, while l
the shunt device initiates a breaker trip when energized.
For an automatic trip, only the UV device is actuated; initiation of the UV devices in either or both RTBs will actuate the control rods.
A manual trip signal operates both the UV device and the separate shunt device.
Either device i:, designed to cause the RTBs to open.
Salem Unit 1 uses Westinghouse 08-50 type RT8s.
At 12:21 a.m. on February 25, 1983, a low-low water level condition in one of the four steam generators at Salem Unit 1 initiated a reactor trip signal in the RPS.
The reactor was at 12% rated thermal power at the time preparatory to power escalation after a recently completed refueling outage.
Upon receipt of 8
the valid reactor trip signal, both of the redundant RTBs failed to open (open-ing of either RTB would have caused the reactor to trip).
About 25 seconds later, operators manually initiated a reactor trip from the control room.
The RTBs opened as a result of the manual trip signal and this resulted in insertion of all control rods and shutdown of the reactor.
Following the manual trip, the plant was stabilized in the hot standby condition.
All other systems func-tioned as designed.
Latec that morning when the cause of the failure had been determined by the licensee, the plant was placed in cold shutdown at the request of the NRC.
During investigation of this incident on February 26, 1983 by the NRC, it was found that a similar failure had occurred on February 22, 1983 at Salem Unit 1.
At 9:55 p.m. on February 22, with the reactor at 20% power, operators were attempting to transfer the 4160 volt group electrical busses from the station power transformers to the auxiliary power transformers, a routine evolution during power escalation.
During the transfer attempt, one of the 4160 busses failed to transfer and deenergized, resulting in the loss of one reactor coolant pump and power for the operating main feed pump control and indication.
At 9:56 p.m., a low-low level condition occurred in one steam generator (due to the loss of the main feed pump), initiating a reactor trip signal.
Due to the abnormal conditions created by the loss of the 4160 volt bus and in anticipation of loss of steam generator water levels, the operator was directed at about the same time to manually initiate a reactor trip.
It was understood by plant personnel and was reported to the NRC that the automatic reactor trip signal due to the low-low level in one steam generator had, in fact, caused the reactor to trip.
On February 26, 1983, as a result of NRC queries, the sequence of events computer printout for February 22 was reviewed in detail and it revealed that the RTBs actually opened in response to the operator's manual trip signal.
Consequently, it is now evident that on February 22 (as on February 25) the two RTBs failed to open upon receipt of an automatic trip signal from the RPS.
Since the operators initiated a manual reactor trip shortly after receipt of the automatic trip signals on both February 22 and February 25, no adverse con-sequences occurred and the reactor was in a safe condition.
On February 22, the operators initiated a manual trip even though they were unaware that the automatic trip had failed.
These events at Salem Unit 1 were of major safety concern since all redundancy was lost to automatically trip the reactor when plant operating conditions required a fast shutdown to protect the integrity of the reactor core.
Safe control of anticipated operating transients is strongly dependent on the reli-able and fast operation of reactor trip, either automatically or manually.
Other pressurized water reactors (PWRs) have experienced RTB failures, both before and after the February 1983 Salem Unit 1 events.
None of them however, involved an ATWS event.
With few exceptions, all PWR plants designed by the three nuclear steam system suppliers (Westinghouse, Babcck & Wilcox, and Combustion Engineering) use an 'RTS design requiring circuit breakers to cpen to trip the reactor.
Although the basic designs of the RTSs arid the number of RTBs per plant differ considerably among the plant designus, each RTB generally includes a UV trip attachment and a shunt trip attachment to actuate the circuit breaker.
Westinghouse designed plants use a Westinghouse type breaker (DB type being the most common) while the other two PWR designers use a General Electric type breaker (AK type).
9
The RTB failures prior to the February 1983 events at Salem Unit 1 have been the subject of several actions taken since 1971 by the AEC/NRC, Westinghouse, and General Electric.
The AEC issued IE Bullet.in No. 71-2 on December 9, 1971 (Ref. 11) as a result of three failures of Westinghouse DB-50 type RTBs at Robinson Unit 2 during j
1971, and two failures of similar breakers at Haddam Neck (Connecticut Yankee) on December 2, 1971.
The Bulletin informed operating PWR licensees of the RTB failures and requested information on the results of testing, inspections, and corrective actions taken, or planned, by other facilities using similar RTBs.
Following further failures at Robinson Unit 2 on December 21, 1973 Westing-house issued Technical Bulletin NSD-TB-74-1 on January 11, 1974, and NSD Data Letter 74-2 on February 19, 1974, regarding recommendations for inspection /
maintenance of DB type breakers.
Because of failures of General Electric AK type reactor breakers at Arkansas Unit 1, Crystal River Unit 3, Oconee Units 1 and 3, and Three Mile Island Unit 1, between April 25, 1975, and January 31, 1979, (and additional failures of similar breakers used in other safety related applications), the NRC issued IE Bulletin No. 79-09 on April 17, 1979 (Ref. 12).
The Bulletin, which included General Electric Service Ad9 ice Letter (SAL) No. 187 (CPDD) 9.3 regarding inspection / maintenance of AK type breakers, stipulated requirements for estab-lishing and performance of an acceptable preventive maintenance program for the t
RTBs.
Further, due to a November 30, 1980 breaker event at St. Lucie dua to an adjust-ment problem, the NRC issued IE Circular No. 81-12 on July 22, 1981 (Ref. 13) in regard to recommended procedures for independently testing the UV and shunt trip devices.
Cause or Causes - On February 25, 1983, approximately two hours after the Salem Unit 1 event, the cause of the failure to trip was determined by licensee instrumentation technicians to be failure of the UV trip device in both RTBs to function as designed.
The same problem had occurred on February 22,1983, but it was not recognized by the licensee.
As previously discussed, the plant on both occasions was shut down by manual operator action.
Possible contributors to the failure of the UV trip devices are 1) dust and dirt; 2) lack of lubrica-l tion; 3) wear; 4) more frequent operation than intended by design; and 5) nick-Based ing of latch surfaces caused from repeated operation of the breakers.
on an independent evaluation of the failed UV trip devices identified by the licensee, the NRC staff concluded that, while the Salem Unit 1 breaker failures occurred as a result of several possible contributors, the predominant cause was excessive wear accelerated by lack of lubrication and improper maintenance.
It appears that no preventive maintenance was conducted on the Salem Unit 1 DB-50 circuit breakers until January 1983.
Additionally, the lubrication recommendations of the Westinghouse 1974 Technical Bulletin and Data Letter were not implemented during the January 1983 maintenance, since personnel per-forming the maintenance (including a Westinghouse service representative) were not aware of this information.
The January maintenance was performed because In of a breaker problem which occurred at Salem Unit 2 on January 6, 1983.
this event, a reactor trip occurred due to a low-low water level condition in one steam generator and only one RTB operated.
The second RTB finally opened 10
25 minutes later, although the reactor had already tripped from opening of the other RTB.
The failure of this RTB was concluded by the licensee to be due to dirt and corrosion interfering with proper operation of the UV trip device.
As I
a result of this event, maintenance was conducted on all Unit 1 RTBs, at least f
one of which involved supervision of the RTB vendor, Westinghouse.
The licensee also reported that all reactor trip breakers were tested after maintenance per plant procedures.
The licensee had also experienced a similar breaker problem at Salem Unit 2 on August 20, 1982; during surveillance, a RTB did not trip due to binding of the f
h UV coil solenoid.
As noted previously, the licensee failed to recognize on February 22, 1983 that t
E an ATW5 event had occurred.
This was due to a lack of a thocough and system-atic review to achieve the necessary understanding of the event.
This, and j
certain previously identified problems at Salem, indicated that a number of corrective actions were necessary before the plants would be allowed to return to operation.
Actions Taken to Prevent Recurrence Licensee - The licensee has completed or committed to complete many corrective actions to address various issues concerning RTBs; operator procedures, training and response; and management issues.
Actions taken by the licensee concerning the RTBs include installing new UV trio devices on Salem Units 1 and 2 which incorporate all design changes made to the devices, augmenting surveillance test requirements, developing a comprehensive maintenance procedure, and incor-porating Westinghouse recommendations regarding maintenance and testing.
Actions concerning operator procedures, training, and response issues include revising emergency procedures to identify actions to be taken in the event a reactor trip signal is received, conducting additional operator training, and evaluating certain aspects of the control room design.
Actions concerning management issues include reviewing past maintenance and procurement documents to ensure that the problems associated with the RTBs did not extend to other safety sys-tems; strengthening administrative controls over maintenance, procurement and post-maintenance testing activities; establishing additional safety review groups within the company; developing a formal post-trip review procedure; instituting a program to update vendor-supplied information; and subjecting the company to independent management assessment by external consulting organizations.
As discussed later, the NRC has accepted the licensee's corrective action program.
Vendors - Westinghouse formed an intercompany task force to conduct an internal review of their procedures for dissemination of technical information to utili-ties.
In addition, they reviewed the testing program for the breakers.
Since there were generic implications associated with the Salem Unit 1 events, Westinghouse worked with the Owners Group (licensees of Westinghouse designed plants) to review operating and emergency procedures, to review for similar type failures in other plant systems, and to assure that the owners have a current listing of current Westinghouse technical information.
11
_l
~
e l
Westinghouse also developed updated maintenance procedures for RTBs, which will be given to the licensees with DB type breakers.
In addition, based on Westinghouse's review of problems experienced with their DS type breakers at the Farley and McGuire facilities, they have alerted the i
NRC and appropriate licensees to potential deficiencies involving clearance and dimensional problems and retaining ring seating defects which could create con-ditions under which the RTBs might not open automatically on demand from the RPS.
Westinghouse also developed procedures for the affected licensees to follow.
DS type breakers are presently used in five operating plants (Farley
{
Units 1 and 2, McGuire Units 1 and 2, and Summer) and specified for use in L
24 plants still under construction.
Combustion Engineering and Babcock & Wilcox also made similar reviews, and in f
cooperation with General Electric, developed updated maintenance procedures for RTBs, which will be given to the licensees with AK-2 type breakers.
Other Licensees - In response to NRC Inspection and Enforcement Bulletin Nos. 83-01 and 83-04 (discussed later), the licensees either performed the RTB testing required (or stated why they were exempt) or provided a schedule for when it would be performed, and took the other actions required by the Bu!1e-tins regarding maintenance, operating procedures, etc.
During, and subsequent to the required testing, additional cases of RTB failures occurred.
NRC - Following the February 25, 1983 event at Salem Unit 1, the NRC sent engi-As a neers to the plant to review the circumstances associated with the event.
result of questions asked by the NRC on February 26, 1983 the licensee again reviewed the circumstances associated with the February 22, 1983 reactor trip and it was discovered that the reactor tripped because of the operator's manual trip signal, rather than from an automatic trip signal as originally reported by the licensee.
Meanwhile, due to the serious nature of the known automatic trip failure in both redundant RTBs on February 25, 1983, the NRC issued Inspection and Enforce-ment Bulletin No. 83-01 (Ref. 14) on the same day to all pressurized water nuclear power reactor facilities holding an operating license for action and to other nuclear power reactor facilities for information.
The Bulletin informed the licensees of the Salem Unit 1 February 25, 1983 event in which both Westing-house DB type breakers failed to open autorhotically (the similarity of the February 22, 1983 event had not yet been ascertained) and mentioned that fail-ures involving only one of the two breakers have previously occurred at Salem Unit 2, Robinson Unit 2, Connecticut Yankee, and St. Lucie.
The Bulletin referenced the two previously discussed pertinent NRC documents; i.e.,
Inspec-tion and Enforcement Bulletin No. 71-02 and Circular No. 81-12, which were issued on December 9, 1971 and July 20, 1981, respectively.
Also mentioned were the previously discussed Westinghouse issued technical information on their breakers; i.e., Technical Bulletin No. NSD-TB-74-1 dated January 11, 1974 and NSD Data Letter 74-2 dated February 14, 1974.
Action items required of licensees using Westinghouse DB type breakers by Bulletin No. 83-01 included, a) testing of the Westinghouse DB type breakers, (b) assuring maintenance is in accord with the recommended Westinghouse program, (c) notifying licensed operators of tne h bm Unit 1 events, (d) reviewing with the operators the procedures to follow in the event of failure of trip, and (e) reporting the results to the NRC.
12
On February 28, 1983, the NRC Executive Director for Operations (EDO) directed that NRC Region I was to develop a detailed report of the Salem Unit 1 events; this report was subsequently issued as NUREG-0977 (Ref. 15).
The EDO further directed that a special NRC task force be formed to evaluate the generic impli-cations of the events.
During the testing required by Bulletin No. 83-01, no further failures of Westinghouse DB type RTBs occurred.
However, even though not required to do so j
by Bulletin No. 83-01, Southern California Edison decided to test the General Electric type AK-2 breakers on their Combustion Engineering designed San Onofre Units 2 and 3.
On March 1, 1983, one of eight RTBs in Unit 3 failed to trip on e
On March 8, 1983, three of eight RTBs in Unit 2 failed to trip on undervoltage.
(Note:
Contrary to the Salem design in which an automatic J
trip signal is fed only to the UV trip devices, the signal is fed to both the UV and shunt trip devices for the San Onofre Units 2 and 3 design.
The shunt devices were satisfactorily tested; therefore, the RTBs would have tripped from an automatic trip signal during operations.) During the investigations of these events, it was found that previous failures had occurred at these units during 1982 but had not been reported to the NRC.
Accordingly, Inspection and Enforcement Bulletin No. 83-04 (Ref. 16) was issued on March 11, 1983 to all pressurized water nuclear power reactor facilities holding an operating license except those with Westinghouse DB type breakers for action and to other nuclear power reactor facilties for information.
The Bulletin described the San Onofre events and mentioned that similar events involving the General Electric AK-2 type breakers had previously occurred at Arkansas Unit 1, Crystal River Unit 3, Oconee Units 1 and 3, Three Mile Island Unit 1, St. Lucie Unit 1, and Rancho Seco Unit 1.
The Bulletin referenced the two previously discussed pertinent NRC documents; i.e.,
Inspection and Enforce-ment Bulletin No. 79-09 (which included General Electric recommended data on the AK-2 type breakers) and Circular No. 81-12 which were issued on April 17.
1979 and July 20, 1981, respectively.
Action items to be taken included, (a) actions similar to those required by Bulletin No. 83-01, (b) licensees to provide a descriptica of all RPS breaker malfunctions not previously reported to the NRC, and (c) licensees to verify that procurement, testing, and main-tenance activities treat the RFS breaker and UV devices as safety related.
In response to Bulletin No. 83-04, additional cases of RTB failures were reported to the NRC.
In addition, other failures occurred after the testing required by Bulletin Nos. 83-01 and 83-04.
In all cases, the NRC closely monitored the corrective actions taken by the licensees to assure that the plants were safe for continued operation.
On April 1, 1983, the NRC issued Inspection and Enforcement Information Notice No. 83-18 (Ref. 17) to all nuclear power reactor facilities holding an operating license or construction permit.
The Notice included the following:
(a) the results of the testing required by Bulletin Nos. 83-01 and 83-04 and that the results show that the breakers may not be achieving the performance reliability expected of them, apparently due to the UV trip attachments.
(b) the problem has ramifications not only for RTBs, but similar breakers used in other plant applications.
13
(c) the importance of regular, careful maintenance of RTBs.
Updated maintenance procedures have been developed by all PWR vendors.
(d) there may be limitations associated with the design life of UV devices such that periodic replacement may be necessary due to wear.
(e) the torque available to trip the General Electric AK-2 breaker by the UV device is critical to proper operation so that certain mea-surements are needed periodically to detect the onset of problems.
(f) the importance of a thorough post-trip analysis, including close a
scrutiny of the events recorder.
The special NRC task force prepared a report (NUREG-1000, Vol. 1) in regard to
]
the generic implications of the Salem events (Ref. 18).
A second report will document the NRC actions to be taken based on the work of the task force.
The f
Commission is being kept informed of progress.
The results of the task force may also affect the ATWS position and rule, which as previously discussed, is being developed by the NRC.
Several briefings were given to the Commissioners by the staff in regard to the Salem problems and other ongoing studies.
The Commission did not permit the Salem plants to restart until not only short term corrective actions were satis-factory, but also until certain management improvements had been satisfactorily addressed.
As noted abcve, the licensee agreed to permit outside consulting firms to evaluate management effectiveness at the Salem plants.
After reviewing the consultants' recommendations, the licensee will generate a plan to incorpo-rate them.
On April 26, 1983, the Commission agreed that the plants could be returned to service, after the NRC staff is satisfied with the licensee's com-mitment to meet certain restart conditions.
The licensee's commitments for both restart and long term conditions, including schedule for implementation, were submitted to the NRC on April 28, 1983.
The NRC incorporated these commitments into an Order dated May 6, 1983.
The licens-ee's commitment for restart was:
"Before entering any new mode, all systems and components required to be operable for that mode, in accordance with Tech-nical Specifications, shall be reviewed to con (irm operability.
If all or part of a system or component has not been shown to be operable within 30 days prior to April 28, 1983, a review shall be conducted to determine if maintenance or other activity has taken place on such system or component since the last opera-bility confirmation.
If such maintenance or other activity has taken place, operability shall be verified by applicable surveillance testing and/or prepara-tion of a written analysis, available for NRC inspection, demonstrating that This the system or component is capable of performing its intended function.
procedure will be followed until the first entry into Mode 1 for each unit r
subsequent to April 28, 1983." The review of documentation and additional surveillance testing necessitated by the licensee's committment delayed plant startup until May 20, 1983.
l The NRC safety evaluation related to plant restart was forwarded to the licensee by a letter dated April 29, 1983 (Ref. 19).
On May 5, 1983, the NRC fowarded to the Salem licensee (Ref. 20) a Notice of Violation and Proposed Imposition of Civil Penalties (for $850,000).
Violations 14
included operation of the reactor even though the RPS could not be considered operable, and several significant deficiencies which contributed to the inoper-ability of the RTBs.
Region I has instituted an augmented inspection program at Salem to monitor the licensee's progress towards completion of longer term corrective actions, including the consultants' recommendations as noted above.
The general issues associated with RTB failures remains under active review by the nuclear industry and the NRC.
Further reports will be made as appropriate.
FUEL CYCLE FACILITIES l
(Other than Nuclear Power Plants)
The NRC is reviewing events reported by these licensees during the first 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 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 licensees such as radiographers, medical institutions, and byproduct material users.
The NRC is reviewing events reported by these licensees during the first calendar quarter of 1983.
As of the date of this report, the NRC had not determinad that any events were abnormal occurrences.
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 first calendar quarter of 1983, the Agreement States reported the following abnormal occurrences to the NRC.
AS83-1 Contamination by and Ingestion of Radioactive Material Appendix A (the general criterion) of this report notes that a major reduction in the degree of protection of the public health or safety can be considered an abnormal occurrence.
Date and Place - On February 5, 1982, authorities at Brown University, located in Providence, Rhode Island reported to the Rhode Island Radiation Control Agency (RCA) by telephone that a research worker had become contaminated and may have ingested radioactive material.
15
f f
Nature and Probable Consequences - A microbiologist discovered that she was contaminated at approximately 3:00 p.m. on February 5, 1982 when she turned on a survey meter prior to starting a laboratory procedure involving the use of phosphorus-32 (P-32). The laboratory has ten research workers whose total use of P-32 labeled organic compounds is several millicuries per month.
Licensee personnel surveyed the entire third floor of the J.W. Wilson Building and determined that the P-32 contamination was limited to the individual's lab coat. a piece of bread found on the individual's desk in her office, and a sheet of paper in that office.
The contaminated individual had eaten two pieces of bread from her lurich at approximately 2:00 p.m.
Another individual who is a researcher in another laboratory had also eaten two pieces of bread from the same lunch at approximately 12:00 p.m.
Bioassay of the contaminated individual indicated an uptake of P-32.
No uptake was indicated for the other individual.
The licensee conducted an inventory of the laboratory's stock P-32 labeled com-pounds and determined that approximately 350 microcuries of material was missing.
Subsequent analysis of the contaminated items accounted for the following activ-ities:
remaining piece of bread, 50 microcuries; lab coat, 22 microcuries; sheet of paper, 20 microcuries.
Calculation of the individual's uptake, based I
on initial urine bioassays, indicated a range of 49-94 microcuries had been ingested.
The licensee estimated that the contaminated laboratory coat had been worn for approximately 35 hours4.050926e-4 days <br />0.00972 hours <br />5.787037e-5 weeks <br />1.33175e-5 months <br />.
Conservatively assuming that the contamination was on the skin rather than on the coat, the licensee estimated that the individual could have received a naximum skin dose of 1100-1500 rads to a small skin area.
On February 17, 1982 the licensee reported to the state that bioassay of the remainder of the research group identified a second individual with an uptake.
When questioned, this individual recalled having taken one piece of candy from the first individual's desk on February 4, 1982.
On March 17 and April 2, 1982, the licensee reported on further studies involving whole body counting of the two individuals with P-32 uptakes.
Estimates of the amounts ingested based upon these studies were 157 microcuries for the first individual and 25 microcuries for the second.
The licensee also reported that clinical exami-nations have been negative and that both individuals were under continuing observation as a precautionary measure.
Cause or Causes - The investigation produced no evidence that the incident was related to or caused by routine laboratory use of P-32.
University police were notified on February 5,1982, and they investigated the matter, with the assist-ance of state and local police.
The University reported to the Agency that their investigation concluded that the contamination was deliberate, but did not produce a suspect.
The State RCA notified the Attorney General of the event early in the course of its investigation.
Actions Taken to Prevent Recurrence Licensee - The University temporarily suspended the use of radioactive materials in the microbiology laboratory until a determination was made that 16
s unsafe operating conditions did not exist.
The licensee's radiation safety committee required improvements in security, survey procedures and records, which the laboratory has implemented.
The committee also required the first individual to abstain from further radioactive materials use pending final i
dosimetry results.
The second individual does not work with radioactive materials.
State Radiation Control Agency - The State RCA conducted an investigation and inspection on February 5, 10, and 18, 1982.
Two immediate action letters were I
issued, confirming steps to be taken by the licensee to reduce the possibility
)
of any recurrence.
The licensee's responses to these letters were reviewed and the licensee's actions were examined by RCA inspectors to assess the adequacy of implementation.
Several items of noncompliance with regulations and license conditions were discovered during the investigation and inspection.
The licensee took action to 'orrect these items, and items relating to the ingestion inci-dents, in accoroance with RCA enforcement procedures.
This incident is closed for purposes of this report.
)
=
I The following four incidents of lost or stolen radioactive sources during 1982 were reported to the NRC by the State of Texas Bureau of Radiation Control (referred to as Agency below) during the first calendar quarter of 1983.
After reviewing the incidents, it was apparent that choosing a specific example (or examples) from Appendix A of this report as the basis of abnormal occurrence reporting may not be appropriate due to the following:
(a) The possibility of theft cannot be definitely ruled out for any of the incidents, (b) The degree of hazard to people in unrestricted areas depends upon the strength of the sources and their ultimate location.
In addition, for those which were stolen, the thieves' knowledge of handling sources would be a factor in the degree of hazard to public health and safety.
Therefore, all four incidents are being reported under the general criterion of Appendix A for abnormal occurrences; i.e., events which involve a moderate or more severe impact on the public health or safety can be considered abnormal occurrences. A description of each event follows.
AS83-2 Lost Radioactive Source Date and Place - Gear - Tex Well. Service, Inc. of Corpus Christi, Texas, reported to the Agency that on March 15, 1982, a 125 millicurie cesium-137 source was found to be missing from its storage container.
Nature and Possible Consequences - While cleaning its facility, the licensee noticed that the storage container containing the 125 millicurie cesium-137 17
q l
source, serial No. CSV-75, was lying on its side with the top off.
When the inner container was removed and opened, the source was found to be missing.
A survey of the facility was performed, but the source was not located.
The licensee then contacted their consultant and was instructed to remove any other sources from the facility and to perform another survey, being sure to stress the area in which the container was found to be lying on its side and the down-hole storage area.
This survey also failed to locate the source.
All of the employees were questioned concerning the source.
None of the employees admitted to having removed the source from the container or remember knocking the con-tainer over.
Later surveys were performed of the facility, the truck used to pickup the trash, and the waste dump by the licensee and representatives of the Agency.
The licensee was required to publish a public notice in local newspapers con-cerning the source.
Due to the small caount of material present in the source, the possibility of l
harm to the general pubiic would be small unless an individual handles the j
source or is near it for a long period of time.
j Cause or Causes - The principle cause was that the source was not stored in its proper downhcle storage container.
A contributing cause was that the top of the container in which the source was stored was not secured with the bolts provided to ensure that the top would not come off.
Actions Taken to Present Recurrence Licensee - The licensee instructed its personnel on the proper procedures for storing radioactive material sources not in use.
Agency - During the investigation of the incident, one item of noncompliance was noted concerning the improper storage of the source.
The licensee satis-factorily addressed this item.
This incident is closed for purposes of this report.
AS83-3 Stolen Radioactive Source Date and Place - On March 26, 1982, Huytech Corporation of Wake Forest, North Carolina reported the theft of a gauge containing a 25 millicurie americium-241 The gauge was stolen from a locked vehicle parked at a hotel in source.
Houston, Texas the night of March 25, 1982.
Nature and Probable Consequences - Huytech Corporation was in Texas under The reciprocity agreement, performing measurement tests for a Texas Company.
Corporation has a North Carolina radioactive materials license and a NRC license. On March 25, 1982, a licensee operator locked the gauge inside his vehicle, which was parked at the hotel in which he was staying.
The following morning he discovered the door locks on the vehicle had been forced and the gauge and other materials had been taken.
The gauge model is 104RPD and the 18
serial number is 380.
The source, manufactured by Amersham, has a model number j
AMCP1 and serial number 0865LA.
The source was last leak tested on February 2, 1982.
Due to the small amount of material present in the source, the possibility of harm is small unless it is mishandled.
Cause or Causes - As described above, the gauge and other material were delib-L erately stolen from the locked vehicle by an unknown individual, or individuals.
)
Actions Taken to Prevent Recurrence Licensee - The licensee notified the Harris County (Texas) Sheriff Department.
f In addition, the licensee notified the North Carolina Department of Human Resources and the NRC.
The licensee will keep the Agency informed of any fur-ther information.
l Agency - The Agency notified the Houston City Police Department of the theft
}
and requested the licensee to report to the Agency any further information con-cerning the theft.
This incident is closed for purposes of this report.
AS83-4 Lost Radioactive Source Date and Place - Dresser Atlas of Houston, Texas, reported to the Agency on July 9, 1982 that a 2 curie cesium-137 source (serial number 350) was missing from its downhole storage location.
Nature and Probable Consequences - While preparing to test a logging tool, a technician proceeaed to the downhole storage to get the source.
The source is a National Bureau of Standards calibrated source used to test new logging equip-ment.
He discovered the source handling tool in the storage hole without the source.
The rest of that day, a Friday, was spent by the employee searching for the source.
The technician failed to notify the Radiation Safety Officer (RS0) until the following Monday morning.
At that time, the RSO attempted to find the source by surveying the grounds of the Houston facility, including all of the storage locations.
After normal work hours, all of the buildings on the site were surveyed to ensure the source was not in areas which could be occupied by the general public, or by personnel not involved with radiation work.
The downhole storages were surveyed for a third time; in addition, the bore holes used to test equipment were surveyed again.
Prior to the survey of the bore holes, a 2 curie cesium-137 source was placed downhole and the licensee had no difficulty detecting the source in that hole or in adjacent holes.
The licensee required all field camps, having vehicles at the Houston facility, to perform an inventory of all sources.
When that inventory failed to locate 19
1 the source, personnel from Houston performed an inspection of camps in Texas, Louisiana and Oklahoma, but without success.
{
There is a possibility of hazard to the public due to the strength of the source and the fact that the source appears to have been taken or missplaced without being ir a storage container.
If an individual were to be in contact or close proximity to the source for a short period of time, an excessive
-exposure could be received.
Cause or Causes - The loss of the source may have been avoided had stronger administrative controls been in effect.
Actions Taken to Prevent Recurrence Licensee - The source security has b'een changed for downhold storage by uti-lizing locks with a limited number of keys.
The licensee has also revised the' utilization logging system to provide greater administrative control.
s s
Agency - The Agency required the licensee to publish a notice to the public, concerning the loss of the source and a description of the source,' including instructions for notifying authoritites with a reward of $250.00 offered for i
information leading to its recovery.
In addition, during an investigation by the Agency, two items of noncompliance were identified.
The licensee subsequently addressed these items satisfactorify.
This incident is closed for purposes of this report.
AS83-5 Stolen Radioactive Source Date and Place - Magnaflux Quality Services of Houston, Texas telephoned the Agency on August 2, 1982 to report the theft of a radiography source and camera.
Nature and Consequences - The stolen source is a 121 Ci, iridium-192 Gulf Nuclear Industries model RG-13 source, serial #6241.
The radiography camera is a Spec. 2-T, serial #208.
The source was being stored at & temporary work site storage building at the Brown and Root Marine Yard in Houston, Texas.
On Friday, July 30, 1982, at approximately 4:30 p.m., the source was signed out by two radiographers and an assistant radiographer.
At approximately 8:00 p.m., after performing their assignment, they returned to the work site and secured the camera by chaining it to the wall inside the storage building and locking the door.
The radiographers processed their film until approximately 9:45 p.m. then left the site.
No other personnel were known to be at the site for the weekend.
At 8:00 a.m. on Monday, August 2, 1982, thecsite supervisor arrived at work and found the storage shed door removed from the' frame and leaning against the lab trailer.
He checked inside the shed and found the Spec. 2-T camera missing.
Checking the utilization log and calling all the technicians failed to locate 20
_1_
3-the camera.
The site sypervisor then called the field supervisor who rechecked all the technicians, with the same negative results.
At this point, the site s$pervisor began an indepth investigation and search for the camera.
All radiographers were again called and interviewed and it was verified the source had not been taken to any ocner site.
The Brown and Root safety supervisor was notified and an extensive search was performed at the Marine Yard, with negative results.
At this point, the-Agency and the Houston Police Department were notitied of the theft.
Subsequently, it was reported to the Harris County (Texas) Sheriff's Department.
No further information is available at this time, f
)
The source would be particularly hazardous if it were removed from the camera.
If an individual were to be in contact or close proximity to the unshie.1ded source for a short period of time, an excessive exposure could be received.
Cause or Causes - As described above, the camera was~ removed from inside a locked storage shed where it was chained to an inside wall, by an unknown individual or individuals.
i l
Actions Taken to Prevent Recurrence Licensee - An extensive investigation and search for the source were performed and when the source could not be found, it was reported to the proper authorities.
Agency - The Agency and the Agency's special investigator have been cooperating with local-authorities in attempting to recover the stolen source.
This incident is closed for purposes of this report.
The following item concerns an incident in New York State (an Agreement State).
The item was prepared by the NRC based on information developed during the State's investigation of the event and followup efforts.
AS83-6 Radioactive Contamination of a Metals Production Facility Appendix A (Example 5 of "For All Licensees") notes that any loss of licensed material in such quantities and under such circumstances that substantial hazard may result to persons in unrestricted areas can be considered an abnor-mal occurrence.
Cate and Place - On February 21, 1983, Auburn Steel Company of Auburn, New York, discovered that a batch of molten steel and some recently cast rods were radioactively contaminated.
Nature and Probabie Consequences - Auburn Steel Company (a New York licensee) manufacturee steel rods for concrete reinforcement.
The rods aie made by melt-ing a "chargo" composed primarily of scrap steel and then loading the melted steel into a casting machine for continuous casting of the rods.
A level gauge, consisting of a sealed radioactive source and a radiation detector, is used to 21
l assure that the level of steel in the casting machine is maintained at the proper level.
On February 21, 1983, the level gauge responded abnormally after a charge was loaded into the casting machine.
The company Radiation Safety Officer (R50) closed the shutter on the level gauge, which shields the radioactive source; however, the radiation detector continued to respond.
The R5G performed sur-veys with a geiger counter and found that the rods produced from this charge had radiation levels of about 24 mr/hr at three feet; in addition, the ladle, casting machine, and the "baghouse" (used to trap airborne particles) were contaminated.
The licensee notified the New York Department of Health, which is responsible for emergency response; they sent a representative to perform onsite surveys.
They also investigated the incident and subsequently prepared a report.
The New York Department of Labor, which licensed the level gauge, was also notified. They sent an inspector to take offsite samples and evaluate offsite releases.
They also audited the licensee's decontamination efforts, and assisted in the efforts to determine the source of the contamination.
The licensee took urine samples and nose wipes from workers who were present in the area to evaluate possible internal exposures, and evaluated external exposures from time and distance information.
The licensee shut down the facility and retained Chem-Nuclear Corporation as the contractor to assist them in performing the necessary decontamination of the facility.
Measurements taken by the New York Department of Health inspectors showed that the steel contained about 4.2 x 105 picocuries per gram of cobalt-60 (Co-60) distributed over more than 130 tons of steel.
It was estimated that the con-tamination resulted from the addition of up to 40-50 curies of Co-60 to the scrap steel.
No other radionuclides were found.
Additional measurements taken at the facility showed 700, 2900, and 540 picocuries per gram of Co-60 in a dust sample from the ventilation system, in a dust sample from inside the plant, and in a composite sample of wipes near the outlet vent of the ventila-tion system located in the plant roof, respectively.
It was determined that the contamination was confined to within the property boundaries of the steel plant and did not present a threat to public health or safety. There was no evidence of worker contamination or overexposures.
The Nes York Department of Labor reviewed and approved the decontamination plan submitted by the licensee's contractor.
The NRC Region I also reviewed and commented on the plan.
The cost of decontamination was estimated to be in excess of one million dollars.
Decontamination of the plant was completed in All materials early April 1983 and the plant returned to normal operations.
and equipment removed from the plant were stored in an onsite, secured shed.
Cause or Causes - The most likely cause was the presence of a sealed Co-60 source in some scrap steel shipment received by the licensee.
The origin of the source and how it became commingled in scrap steel, have not been deter-mined.
As discussed below, an extensive search for these answers was made.
22
Actions Taken to Prevent Recurrence Licensee /New York State Agencies - The licensee noted that scrap steel is obtained from about 100 sources in the northeast United States and in Canada.
Such scrap is usually processed within 10-14 days.
New York State Police are investigating the records of scrap shipments to the plant.
The State Agencies also were investigating companies authorized to manufacture, distribute, or possess Co-60 sealed sources in New York State.
The licensee also contacted its suppliers of scrap steel to attempt to determine the source of the contaci-nation.
As of the date of this report, the results of the investigation have not been conclusive.
The New York Department of Labor plans to issue a report regarding the incident.
NRC - The NRC Region I offered assistance to the New York agencies, if they desired. The NRC Office of International Programs was notified of the possible international aspects.
NRC Region I furnished the New York State Police a list of companies authorized to manufacture or distribute Co-60 sealed sources, and a list of NRC licensees in New York authorized to possess Co-60.
In addition, Inspection and Enforcement Information Notice No. 83-16 (Ref. 21) i
{
was issued on March 30, 1983 to all material licensees.
The Notice described the event and suggested the licensees review the provisions of the regulations and their licenses that deal with control of licensed materials and reporting lost or stolen material.
For those licensees licensed to possess a Co-60 source, it was suggested that an inventory be conducted; if any source was not accounted for, the NRC should be notified immediately in accordance with the regulations.
Unless additional significant information becomes available, this incident is closed for purposes of this report.
23
REFERENCES 1.
U.S. Nuclear Regulatory Commission, " Abnormal Occurrence:
Main Feedwater Line Break Due to Water Hammer," Federal Register.
(Item is being pub-lished in the Federal Register concurrently with this report.)
2.
U.S. Nuclear Regulatory Commission, Inspection and Enforcement Bulletin No. 79-13, " Cracking in Feedwater System Piping," June 25, 1979,* with Revision 1 issued on August 30, 1979,* and Revision 2 issued on October 16, 1979.*
3.
Letter from Robert A. Clark, Chief, Operating Reactors Branch #3, Division of Licensing, NRC Office of Nuclear Reactor Regulation, to John H. Garrity, Senior Director, Nuclear Engineering and Licensing, Maine Yankee Atomic f
Power Company, transmitting a safety evaluation dated March 18, 1983, Docket No. 50-309, March 18, 1983.*
4.
U.S. Nuclear Regulatory Commission, " Abnormal Occurence:
Deficiencies in Management and Procedural Controls," Federal Register.
(Item is being published in the Federal Register concurrently with this report.)
5.
Letter from Richard C. DeYoung, Director, NRC Office of Inspection and Enforcement, to E.E. Utley, Executive Vice President, Carolina Power and Light Company, transmitting a Notice of Violation and Proposed Imposition of Civil Penalties, Docket Nos. 50-324 and 50-325, February 18, 1983.*
6.
Confirmation of Action letter from James P. O'Reilly, Regional Administra-tor, NRC Region II, to J. A. Jones, Senior Executive Vice President and Chief Operating Officer, Carolina Power and Light Company, Docket Nos.
50-324 and 50-325, July 2, 1982.*
7.
Cor.firmation of Action letter from James P. O'Reilly, Regional Administra-tor, NRC Region II, to J. A. Jones, Vice Chairman, Carolina Power and Lighc Company, Docket Nos. 50-324 and 50-325, July 20,1982.*
8.
Letter from E. E. Utley, Executive Vice President, Carolina Power and Light Company, to James P. O'Reilly, Regional Administrator, NRC Region II, forwarding Brunswick Improvement Program dated November 1982, Docket Nos. 50-324 and 50-325, October 29, 1982.*
9.
Letter from Richard C. DeYoung, Director, NRC Office of Inspection and Enforcement, to E. E. Utley, Executive Vice President, Carolina Power and Light Company, transmitting Confirmatory Order EA-82-106 dated December 22, 1982, Docket Nos. 50-324 and 50-325, December 22, 1982.*
- Available in NRC Public Document Room, 1717 H Street, NW., Washington, D.C.
20o55, for inspection and copying (for a fee).
- Available to purchase from NRC-GPO Sales Program, Division of Technical Information and Document Control, U.S. Nuclear Regulatory Commission, Washington, DC 20555.
25
10.
U.S. Nuclear Regulatory Commission, " Abnormal Occurence:
Failure of Reactor Trip System," Federal Register.
(Item is bEing publiLhed in the Federal Register concurrently with this report. )
11.
U.S. Nuclear Regulatory Commission, Bulletin No. 71-2, regarding Westinghouse DB-50 reactor scram circu t breakers, December 9, 1971.*
i 12.
U.S. Nuclear Regulatory Commission, Inspection and Enforcement Bulletin No. 79-09, Failures of GE Type AX-2 Circuit Breakers in Safety Related Systems," April 17, 1979.*
13.
U.S. Nuclear Regulatory Commission, Inspection and Enforcement Circular No. 81-12, " Inadequate Pericdic Test Procedure of PWR Protection System,"
July 22, 1981.*
14.
U.S. Nuclear Regulatory Commission, Inspection and Enforcement Bulletin No. 83-01, " Failure of Reactor Trip Breakers (Westinghouse 0B-50) to Open on Automatic Trip Signal," February 25, 1981
- 15.
U.S. Nuclear Regulatory Commission, "NRC Fact-Finding Task Force on the ATWS Events at Salem Nuclear Generating Station, Unit 1, on February 22 and 25, 1983," USNRC Report NUREG-0977, published March 1983.**
16.
U.S. Nuclear Regulatory Commission, Inspecticn and Enforcement Bulletin No. 83-04, " Failure of the Undervoltage Trip Function of Reactor Trip Breakers," March 11, 1983.*
17.
U.S. Nuclear Regalatory Commission, Inspection and Enforcement Information Notice No. 83-18, " Failures of the Undervoltage Trip Function of Reactor 1 rip System Breakers," April 1, 1983.*
18.
U.S. Nuclear Regulatory Commission, " Generic Implications of ATWS Events at the Salem Nuclear Power Plant," USNRC Report NUREG-1000, Vol. 1, published April 1983.**
19.
Letter from D. G. Eisenhut, Director, Division of Licensing, NRC Office of Nuclear Reactor Regulation, to R. A. Uderitz, Vice President - Nuclear, Public Service Electric and Gas Company, transmitting "NRC Safety Evalua-tion Related to Plant Restart (dated April 28, 1983)," Docket Nos. 50-272 and 50-311, April 29,1983.
- The safety evaluation report is to be pub-lished as NUREG-0995.
)
20.
Letter from Richard C. DeYoung, Director, NRC Office of Inspection and Enforcement, to Robert Smith, Chairman of the Board, Public Service and Gas Company, transmitting a Notice of Violation and Proposed Imposition of Civil Penalties, Docket Nos. 50-272 and 50-311, May 5,1983.*
21 U.S. Nuclear Regulatory Commission, Inspection and Enforcement Information Notice No. 83-16, " Contamination of the Auburn Steel Company Property with Cobalt-60," March 30, 1983.*
26 C
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:
l 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, constraction, use of, or management controis 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 radiation; 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 S 20.403(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 S 20.105(a)).
3.
The release of radioactive material to an unrestricted area in concentrations 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 S 20 (10 CFR S 20.403(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 L S 71.36(a)).
27
5.
Any loss of licensed material in such quantities and under such circumstances 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.
7.
Any substantiated loss of special nuclear material or any substan-tiated inventory discrepancy which is judged to De 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 S 70.52(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.
l 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 Commerical Nuclear Power Plants l
1.
Exceeding a safety limit of license Technical Specifications (10 CFR S 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 S 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 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 S 100 guide-lines could result from a postulated transient or accident (e.g.,
loss of emergency core cooling system, loss of control rod system).
28
f l
For fuel Cycle Licenses 1.
A safety limit of license Technical Specifications is exceeded and a plant shutdown is required (10 CFR S 50.36(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.
29
~
APPENDIX B UPDATE OF PREVIOUSLY REPORTED ABNORMAL OCCURRENCES During the January through March 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 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 75-5 Cracks in Pipes at Boiling Water Reactors (BWRs)
This abnormal occurrence was originally reported in NUREG-75/090, " Report to the Congress on Abnormal Occurrences:
January-June 1975," and updated in subsequent reports in this series, i.e., NUREG-0090-1; 0090-2; 0090-3; 0090-9; Vol. 1, No. 3; Vol. 2, No. ?; Vol. 2, No. 4; Vol. 3, No. 2; Vol. 3, No. 4; Vol. 5, No. 2; and Vol. 5, No. 4.
It is further updated as follows.
NUREG-0090, Vol. 5, No. 4, included an update to describe cracks detected in the recirculation system piping at Monticello, Hatch Unit 1 and Browns Ferry Unit 2.
The flaws were either repaired or were evaluated by the licensees and the NRC staff and judged to be sufficiently small that the plants could be safely operated with the flaws and with implementation of augmented leak detection equipment and procedures.
As was also described in the last report, NRC Inspection and Enforcement Bulletin No. 82-03 (Ref. B-1) was issued on October 14, 1982, with Revision 1 on October 28, 1982 (Ref. B-2), requiring all BWRs that were shut down or scheduled to be shut down by January 31, 1983 to augment the normal inspections of the recirculation system piping; the licensees were also required to demonstrate the capability of their personnel and procedures for detecting very small cracks in pipe samples taken from Nine Mile Point Unit 1.
As a result of the significantly larger number of welds which licensees were required to examine and the much more sensitive detection techniques, indica-tions of cracks were found in the recirculation system piping at three addi-tional plants which had conducted inspections during the first quarter of 1983.
These plants were Brunswick Unit 1, Dresden Unit 2 and Oyster Creek.
Evalua-tion of these flaws is in progress.
As a follow-on to IE Bulletin No. 82-03 discussed above, IE Bulletin 83'02 (Ref. B-3) was issued on March 4, 1983 to extend the augmented inspection requirements to all BWRs scheduled to shut down by January 1984 and continuing the requirement that licensees demonstrate the capability to detect small, tight cracks.
31
Based on the examinations required by Bulletin No. 83-02, pipe cracks in the reactor recirculation system were found at Vermont Yankee during April 1983.
Cracks were found in several large diameter (22 and 28 inch) and smaller diam-eter (12 inch) pipe welds.
The large diameter pipe cracks were evaluated by the licensee to be in the range of 10-15% through-wall in depth and were judged acceptable for continued operation.
However, the smaller diameter pipe cracks exceeded acceptance criteria and required repair.
Also, during April 1983, crack indications were found at Hatch Unit 2 and Peach Bottom Unit 3.
At Hatch Unit 2, the indications were in the endcap of a 22 inch recirculation jet pump riser header.
At Peach Bottom Unit 3, preliminary results found indications in some recirculation pump riser welds.
The NRC is closely monitoring all of the affected licensees' inspection programs and, along with the licensees, evaluating any indications found, and repairs made to assure that the plants are safe to resume power operation.
Further reports will be made as appropriate.
77-9 Environmental Qualification of Safety-Related Electrical Equipment Inside Containment lhis abnormal occurrence was originally reported in NUREG-0090-10, " Report to Congress on Abnormal Occurrences:
October - December 1977," and updated in subsequent reports in this series, i.e.,
NUREG-0090; Vol. 1, No. 1; Vol. 1, No. 2: Vol. 2, No. 2; Vol. 3, No. 2; Vol. 4, No. 2, and Vol 5, No. 2.
It is further updated as follows.
A new rule, " Environmental Qualification of Electrical Equipment Important to Safety for Nuclear Power Plants," Section 50.49 to 10 CFR Part 50, was published in the Federal Register on.lanuary 21, 1983 and became effective on February 22, 1983 (Ref. B-4).
The scopa of the rule covers electrical equipment important to safety as defined by the rule.
This covers electrical equipment important to safety both inside and outside containment that would be subject to a harsh environment following design basis events.
The rule specifies the qualification parameters and methods that an equipment qualification program should include.
In addition, the rule specifies a deadline by which all electrical equipment covered by the scope of the rule should be qualified.
Regulatory Guide 1.89, Revision 1 is currently scheduled for issuance ir.
mid-1983.
(Public comments are now being resolved).
The Regulatory Guide clarifies the new rule and specifies acceptable methods for compliance with the rule.
Unless significant new information becomes available, this item is closed for purposes of this report.
79-3 Nuclear Accident at Three Mile Island This abnormal occurrence was originally reported in NUREG-0090, Vol. 2, No. 1,
" Report to Congress on Abnormal Occurrences:
January-March 1979," and updateo 32
/
in subsequent reports in this series, i.e.,
NUREG-0090, Vol. 2, No. 2; Vol. 2, 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, 1
No. 1; Vol. 5, No. 2; Vol. 5, No. 3; and Vol. 5, No. 4.
It is further updated l
as follows:
l
, Reactor Building Entries During the first quitrter of 1983, fifty-seven entries were made into contain-ment, with a total exposure of 135.5 man-rem.
Major activities included dose reduction efforts on the 305' and 347' elevations and decontamination of the reactor vessel service structure, the refueling pool, the D-rings and the reactor building air coolers.
Another major task was continuation of polar crane refurbishment.
Other activities included the removal of the neutron shield tanks from around the reactor vessel and " quick scan" measurements of radiological conditions under the reactor vessel head.
During the quick scan, it was learned that radiation levels above the reactor plenum may be higher than expected (gamma dose in excess of 500 R/hr.), and that the contamination is more firmly fixed than previously anticipated.
Dose Reduction Program In March 1983, the licensee began work inside containment in an attempt to reduce the general area dose rates on the 305' and 347' elevations.
The reactor buildir.g air coolers, which are a significant source of radiation near the personnel entry point, were decontaminated externally and internally by high temperature, high pressure water sprays.
There was no apparent decrease in air cooler dose rates; however, shielding placed on the open stairwell and on the equipment hatch decreased dose rates in these areas by over 50 percent.
Shielding placed around the "B"
core flood tank and the enclosed stairwell was also very effective in reducing dose rates in these areas.
EPICOR-II Submerged Demineralizer System (SDS) Processing Approximately 127,800 gallons of water were processed by the EPICOR-II system during the first quarter of 1983.
Sources of the water included the contami-nated drain tank, the A-once-through steam generator, the chemical cleaning building sump, and the SDS.
The SDS processed approximately 203,200 gallons of water with its sources being the reactor building sump and the reactor coolant system.
EPICOR-II Prefilter and SDS Liner Shipments Seventeen of the remaining EPICOR-II prefilters were shipped from TMI to the Idaho National Engineering Laboratory (INEL) in Scoville, Idaho during the first quarter of 1983.
It is anticipated that the remaining sixteen will be shipped before the end of August 1983.
A total of four recombiner loaded SDS waste liners were shipped to the DOE facility in Richland, Washington.
There are currently six additional SDS liners in storage awaiting shipment.
33
Advisory Panel for the Decontamination of TMI-2 On February 2,1983, the Three Mile Island Advisory Panel held a meeting in Harrisburg, PA.
Members of the GPU TMI-2 Safety Advisory Board (SA8) provided an overview of their functinns and findings to date.
Dr. J. Fletcher, the Chairman of the SAB, discussed the role of the board and its interaction with GPU Nuclear Corporation (GPUN).
Dr. N. Rasmussen provided a technical descrip-tion of some of the activities at the site.
Dr. R. Friedman, Chairman of the External Affairs Panel of the SAB, discussed the Panel's role in communicating information between GPUN and the local communities.
Representatives from GPUN provided an overview of the latest TMI-2 Recovery Program Estimate.
Five different alternatives, based on different levels and schedules of funding, were presented.
The status of the cleanup was summarized by Mr. Bahman Kanga, the Director of TMI-2.
Mr. Robert C. Arnold, President of GPUN, provided an update on funding, summarizing the expected sources of revenues for covering the cost of environmental radiological monitoring in the vicinity of the site and the region.
On March 17, 1983, the Three Mile Island Advisory Panel held another meeting in Harris'urg.
Representatives from the NRC, EPA and DOE provided an update of o
their respective agency's activities.
Dr. Bixby from DOE presented a short video tape summarizing the agency's accomplishments at the TMI-2 facility over the past year.
Mr. Gerusky, Director, Pennsylvania Bureau of Radiation Protection, provided a detailed summary of the current status of the proposed low-level radioactive waste compact for the Northeast.
Under the auspices of the coalition of North-east Governor's (CONEG), the CONEG Policy Working Group completed a draft com-pact on February 18, 1983, and it was sent to the states for approval by the State legislators.
The implications of adoption of a compact, or failure to adopt a compact, on the TMI-2 cleanup effort were discussed by Mr. Arnold from GPUN.
Mr. Arnold stated that the amount of low-level radioactive wastes gener-ated at the TMI-2 facility was approximately equal to that from an operating plant.
Contractor Employee Allegations On March 23, 1983, the Government Accountability Project (GAP) forwarded an affidavit dated March 21, 1983 from a GPUN contractor (Bechtel) employee to the Chairman of the NRC, Nunzio Palladino.
The contractor employee made alle-gations about the safety of the reactor building polar crane, general mismanage-ment of the cleanup, and NRC collusion in these activities.
As a result of the allegations made in the affidavit, Chairman Palladino asked the NRC's Office of Investigations to review the conduct of the licensee (GPU Nuclear Corporation) and its contractors (primarily Bechtel North American) to evaluate the merits of the allegations.
The NRC's Office of Inspector and Auditor is also reviewing the allegations regarding the conduct of NRC employees.
Further reports will be made as appropriate.
A A
34
OTHER NRC LICENSEES 82-6 Radiological Contamination from Well Logging Operations This abnormal occurrence was originally reported in NUREG-0090, Vol. 5, No. 3,
" Report to Congrets on Abnormal Occurrences:
July-September 1982," and updated in a subsequent report in this series, i.e., NUREG-0090, Vol. 5, No. 4.
It is further updated as follows.
The licensee completed decontamination of the second drilling site near Pine Bank, Pennsylvania, and moved contaminated equipment and waste to the original drilling site near Jollytown, Pennsylvania.
The licensee released the Pine 3
Bank site for unrestricted use on November 15, 1982, following a verification survey by an NRC Region I inspector.
All vehicles and equipment at the Jolly-town site were subsequently decontaminated and released for unrestricted use, leaving only contaminated soil to be removed to complete the cleanup.
NRC Region I made spot checks during periodic site visits to confirm the accepta-bility of the licensee's decontamination efforts.
Inclement weather has hampered cleanup of the remaining contamination in soil, but the licensee plans to complete decontamination by the summer of 1983.
The licensee had generated approximately 1,000 55 gallon drums of americium-241 contaminated waste as of January 1983 and estimates that they may generate up to a total of 3000 drums.
They have received a permit from the State of Washington to dispose of this waste at the Richland, Washington burial site, but have made no waste shipments as of April 1983.
The total estimated cost of decontamina-tion has been revised upward to 1.5 million dollars.
As a clarification to the original abnormal occurrence report in NUREG-0090, Vol. 5, No. 3, the licensee notified the NRC Region I office on August 27, 1982 that they were in the prucess of recovering a well logging device from a well hole and had identified radiation levels greater than background at the well head.
The information supplied by the licensee led the Region I staff to agree with the licensee staff's conclusion that they had moved the intact device to near the top of the well head.
This agreement was based on a misunderstanding of the operation in progress by the Region I staff.
The radiation levels were apparently the result of rupture of the source.
In the original report, a contributing cause was identified as " inadequate use of survey instrumentation." The licensee did make continuous radiological surveys with the instrument which was approved in their NRC license.
- However, they did not correctly interpret the positive readings by that instrument.
A more sensitive instrument would have given a positive reading earlier and would have given results which were easier to interpret.
No further significant information is expected in regard to this event.
The NRC Region I office will contin'ue to monitor the licensee's followup efforts.
On May 26, 1983, the NRC issued Inspection and Enforcement Information Notice No. 83-32 (Ref. B-5) to all NRC licensees holding a specific license to possess and use sealed sources containing by product or special nuclear material in 35
l well logging tools.
The Notice described the event and contained suggestions regarding procedures.
This incident is closed for purposes of this report.
36
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1
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APPENDIX C OTHER EVENTS OF INTEREST The following events are described below because they may possibly be perceived by the public to be of public health significance.
None of the events involved a major reduction in the level of protection provided for public health or safety; therefore, they are not reportable as abnormal occurrences.
1.
Radioactive Release On January 16, 1983, a radioactive release to the Tennessee River occurred from Browns Ferry Unit 3.
At the time, Unit 3 was in cold shutdown for maintenance; Unit 2 was down for refueling; and Unit 1 continued to operate during the event without interruption.
Browns Ferry Units 1, 2, and 3 are General Electric designed boiling water reactors and are located in Limestone County, Alabama.
The Units are operated by the Tennessee Valley Authority (the licensee).
At 12:14 a.m. on January 16, 1983, the licensee placed the 3B pump and heat exchanger on Unit 3's residual heat removal system (RHR) in service and removed the 3D pump and heat exchanger after a routine RHR service water sample indi-cated high activity (1.9 times maximum permissible concentrations).
- However, no alarm had been received from the service water effluent monitor.
The licensee declared an Unusual Event at 2:04 a.m.
At this time, only the 3B RHR was operable since both the 3A and 3C RHRs were previously out of service for valve maintenance.
In the proc, of pressurizing the 3D heat exchanger to verify a leak, a radia-tion monitor on the service water effluent alarmed at 8:25 a.m.
The radiation monitor is common to both the 3D and 3B heat exchangers; therefore, until the cause of the alarm could be determined, the licensee isolated the 3B pump and heat exchanger.
Since this resulted in a complete loss of normal shutdcwn cooling capability for Unit 3, the licensee then declared an Alert at 8:30 a.m.
in accordance with their procedures.
Until the licensee could confirm that the 3B heat exchanger was not leaking, an alternate means of core cooling was established by use of the condenser, control rod drive, and reactor water cleanup systems.
Reactor coolant temperature increased from 185 F to a peak of 205 F, and then started decreasing.
In addi-tion, a cross tie to the Unit 2 RHR could have been established for cooling had it been necessary.
Pressure testing indicated that the 3B heat exchanger was not leaking.
The 3B pump and heat exchanger were returned to service and the licensee cancelled the plant Alert at 7:17 p.m.
Both sides of the 3D heat exchanger were isolated and drained; however, until this could be accomplished, the defective 3D heat exchanger leaked reactor coolant into the RHR service water system which dis-charges into the Wheeler Reservoir and on into the Tennesssee River.
The amount i
of radioactivity released, about 0.015 curies in over 200,000 gallons of water, j
37
was in excess of technical specification limits; however, no significant envi-ronmental effects would be expected, particularly considering the large dilution of the radioactivity.
It was later determined that the 3D heat exchanger contained 12 dented tubes, with one dented tube leaking.
All 12 tubes were plugged to remove them from service.
An investigation will continue during the next refueling outage to determine the cause.
The impact of the event on public health and safety was minimal; therefore, the event is not considered reportable as an abnormal occurrence.
It is being reported as an Appendix C item since the event received both local and national media interest following issuance of several news releases by the licensee during the course of the event.
2.
Uranium in Cloisonne Jewelry On January 25, 1983, the New York State Department of Health issued a press release announcing that it had discovered pieces of Cloisonne jewelery to be l
radioactive from small amounts of uranium contained in the enamel glaze on the jewelry.
The press release advised people that although the jewelry was not an immediate health hazard, it should be discarded or returned to the place of purchase to avoid unnecessary low-level radiation exposure.
Based on the press release, an article appeared in the New York Times on January 25, 1983.
The article resulted in widespread public interest and the several States and the NRC answered hundreds of telephone calls from concerned citizens over the fol-lowing weeks.
In response to the many requests from members of the public, including jewelry merchants and distributors, the States and the NRC examined thousands of pieces of the jewelry using radiation detection instruments and found that only a small fraction, about 10 percent, contained the radioactive glass enamel.
The primary source of the radioactive jewelry was Taiwan.
The Taiwan govern-ment was apparently not aware that uranium-bearing glass enamel material was being imported and used in the making of jewelry for export.
The primary source of the uranium colored glass enamel material used in Taiwan was a company in the United Kingdom.
The regulatory body of the United Kingdom was apparently unaware that the material was being exported for use in jewelry.
Both Taiwan and the United Kingdom have informed NRC of actions being taken which would stop the use of uranium-bearing glaze in Cloisonne jewelry in Taiwan.
NRC had samples of the jewelry analyzed by a DOE laboratory.
Some gold, yel-low, and beige enamels used in the jewelry were found to contain between 3 and 7 percent by weight of uranium.
Small amounts of uranium have been used for years to provide coloring-in glass enamels and glazed ceramics.
An exemption in NRC regulations permits this use of uranium if it is less than 10 percent by weight.
In its radiological assessment, NRC determined that radiation measurements at contact with enameled faces of the jewelry showed no alpha radiation, very low gamma radiation, beta radiation dose rates between 3 and 6 millirads per hour, 38
- ~ _ _ _ - _
and a depth dose at 1 centimeter that was below detection limits.
While still small, the highest risk would be to a person wearing a piece of the jewelry I
with a radioactive enamel face in contact with their skin.
A piece of jewelry worn in contact with the skin for 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> per week, 52 weeks per year, would result in a dose to the local area of the skin of 2 to 4 rads, corresponding to a risk of getting skin cancer of ? to 4 in one million.
The radiation dose and resultant risk potentially received by jewelry merchants, shoppers, and other members of the public, including children, are less than for someone wearing the jewelry.
The level of dose is within the limits recommended by the International Com-mission on Radiation Protection for members of the general public.
While wearing the jewelry for excessive periods of time could result in a radiation dose to the skin which might be considered unacceptable by members of the
-ger.eral public, the level of risk from normal use and wear of the jewelry is not any greater than that normally accepted by individuals in usual daily activities.
This was not an event involving a sajor reduction in the degree of protection of public health and safety, and is therefore not reportable as an abnormal occurrence.
It is described because it received widespread media coverage and may be perceived by the public to be of public health significance.
3.
Damaged Fuel Cladding On January 15, 1983, Alabama Power Company (the licensee) began the Cycle 4 refueling outage of Farley Unit 1.
On January 29, 1983, the licensee reported several damaged fuel assemblies based on visual examination.
Farley Unit 1 is a Westinghouse designed pressurized water reactor located in Houston County, Alabama.
Damaged fuel cladding became evident during Cycle 4 operation since the reactor coolant iodine radioactivity gradually increased to about 45% of technical specifications allowable limits.
The cause was also known since a similar problem had occurred earlier at the Trojan Nuclear Plant, which was reported as an Appendix C item in an earlier report of this series, i.e., NUREG-0900, Vol. 5, No. 2, " Report to Congress on Abnormal Occurrences:
April-June 1982."
Following the Trojan event, the NRC issued Inspection and Enforcement Informa-tion Notice No. 82-27 (Ref. C-1) to licensees on August 5, 198? to inform them of the event.
The cause of the fuel cladding damage was due to vibrations induced by water jetting through joints in the core baffle, a bolted steel assembly which sur-rounds the reactor core.
The baffle jetting is the result of a pressure differential across the baffle plate joints which exists due to reactor coolant flowing down the outer surface and flowing up on the core side'of the baffle plates.
The vibration of the fuel rods, all in peripheral locations in the core, con inued until the cladding was damaged.
39
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During the Cycle 3 outage of Farley Unit 1, the core baffle plate corner joints had been peened in an attempt to close the corner joint gaps; however, the I
peening apparently only made the. problem worse.
{
Visual inspection showed 11 fuel assemblies with damaged fuel pins.
Four other
(
fuel assemb!!es were determined to be leaking on the basis of fuel sipping, a technique which checks for the release of fission products from the fuel assem-bly.
The 15 damaged fuel assemblies were replaced with new fuel assemblies.
Debris from the damaged fuel, consisting of one 3-foot piece of fuel rod plus several other smaller pieces as well as 30 to 50 fuel pellets, were recovered from the reactor vessel.
During Cycle 4, the licensee and Westinghouse, the fuel vendor, completed plans for a design change to eliminate the water jetting.
This change, which con-verted the coolant flow from down flow to up flow on the baffle exterior (thereby eliminating the pressure differential), was completed during the refueling outage.
After refueling, during preparations to start up the plant, noise monitors detected an unusual noise in the hotleg to steam generator (SG) IC.
Investi-gation revealed a piece of a control rod guide tube support pir (1 3/4" long by 3/4" diameter) in the primary side of SG 1C, However, no SG tube damage resulted since the noise monitor detected the loose part early.
The failure of the support pin was attributed to the same cause as similar failures at North Anna Unit 1, which was reported as an Appendix C item in the previously dis-l cussed earlier report of this series, i.e., NUREG-0900, Vol. 5, No. 2.
l Licensees were informed of the North Anna Unit 1 event by NRC Inspection and Enforcement Information Notice No. 82-29, issued July 23, 1982 (Ref. C-2).
Both Westinghouse and the licensee performed a safety evaluation prior to the I
return to power for completion of the power test programs.
This item is not considered reportable as an abnormal occurrence since the effects on public health and safety were minimal.
The fuel failures did not result in r1dioactivity levels or effluent releases in excess of those allowed by the plant's license.
In addition, the foreign object in SG 1C was detected and removed before SG tube damage occurred.
4.
Plant Construction Deficiencies at Midland Nuclear Power Station l
On February 8, 1983, the NRC's Region III Office (Chicago) proposed a $120,000 l
fine against Consumers Power Company (the licensee) for alleged violations in the Quality Assurance Program at the Midland Nuclear Power Station (Ref. C-3).
The station, which is under construction in Midland County, Michigan, will use two Babcock & Wilcox designed pressurized water reactors.
These violations were first identified by the NRC staff in October and November 1982 during inspections of the Midland Diesel Generator Building.
As a result of the NRC's findings, coupled with its own investigation of quality assurance problems at Midland, the licensee announced on December 2, 1982, that it was halting most safety-related construction work at the plant and was laying off 40
approximately 1,000 of its 5,000 member workforce.
(Not affected by the work 3
stoppage were installation of the nuclear steam supply system by Babcock and
/
Wilcox, Inc., installation of the heating, ventilating and air conditioning system by the Zack Company, remedial soils work, ongoing inspection and main-tenance activities, and Bechtel Corporation engineering work).
The $120,000 proposed fine pertained to two violations, each carrying a $60,000 penalty. The first violation was for a breakdown in the licensee's implemen-tation of the Quality Assurance program at Midland.
This was evidenced by multiple examples of plant personnel failing to follow procedures, drawings, and specificationc in the installation of safety-related equipment.
The second violation was the result of the NRC's determination that quality control super-visors instructed quality-control inspectors to suspend inspectiors when exces-sive numbers of deficiencies were observed.
The construction being inspected was then turned back to the construction staff for rework.
The intent of this practice was to improve construction quality prior to completing the quality control (QC) inspections.
In some cases, however, the follow-up QC inspections focused only on the previously identified deficiencies, instead of conducting a full " reinspection." This practice, as a result, provided no assurance that undocumented deficiencies were later identified cr repaired (all areas where this practice occurred are to be reinspected by the licensee, as a result of the NRC inspection which identified the violation).
Mcanwhile, the licensee requested that the $120,000 fine be mitigated.
The request, dated March 28, 1983, was in response to the NRC's " Notice of Viola-tion and Proposed Imposition of Civil Penalties" dated February 8, 1983.
The licensee's response is currently being reviewed by the NRC staff.
The licensee also has announced a proposed plan for a " Construction Completion Program."
This program, designed to identify and correct past deficiencies and assure the orderly and efficient completion of future construction, was outlined in a January 10, 1983, letter to the Region III Administrator (Ref. C-4). The pro-gram includes:
integrating the Bechtel quality control function into the licensee's; retraining and recertification of QC personnel; reinspecting safety-related systems; and the integrating of engineering and construction workers into teams with each team responsible for the completion of one or more safety systems.
In addition, the licensee has committed to using independent, third parties to (1) overview the entire Construction Completion Program, and (2) to conduct an independent design and construction verification program.
The Construction Completion Program is still under review by the NRC staff and has not yet been approved.
Since the deficiencies were found while the plant was still under construction, and there was no reactor fuel on site, there was no impact on public health or safety; therefore, the item is not reportable as an abnormal occurrence.
ERRATA In the previous report of this series (i.e., NUREG-0090, Vol. 5, No. 1), the first Appendix C item described an event involving control rod drive failure and reactor trip.
In the first paragraph, a line was inadvertently dropped 41
in the final processing of the report.
The complete paragraph should have read as follows.
1.
Control Rod Drive Failure and Reactor Trip On September 30, 1982, Commonwealth Edison Company (the licensee) experienced l
a control rod insertion problem at their Zion Unit 1 plant while the plant was operating at full power.
The control rods would not move into the reactor core in the normal operational mode, position-by position.
The capability to fully insert all rods immediately (scram) remained available, both manually and auto-matically.
Zion Unit 1 utilizes a Westinghouse designed pressurized water reactor and is located in Lake County, Illinois.
l l
l 42
~
l l
REFERENCES (FOR APPENDICES)
B-1 U.S. Nuclear Regulatory Commission, Inspection and Enforcement Bulletin No. 82-03, " Steam Corrosion Cracking in Thick-Wall, Large-Diameter, Stainless Steel, Recirculation System Piping at BWR Plants," October 14, 1982.*
B-2 U.S; Nuclear Regulatory Commission, Inspection and Enforcement Bulletin No. 82-03, Revision 1, " Stress Corrosion Cracking in Thick-Wall, Large-Diameter, Stainless Steel, Recirculation System Piping at BWR Plants," October 28, 1982.*
B-3 U.S. Nuclear Regulatory Commission, Inspection and Enforcement Bulletin No. 83-02, " Stress Corrosion Cracking in Large-Diameter Stainless Steel Recirculation System Piping at BWR Plants," March 4, 1983.*
B-4 U.S. Nuclear Regulatory Commission, " Environmental Qualification of Electric Equipment Important to Safety for Nuclear Power Plants,"
Federal Register Vol. 48, No. 15, January 21, 1983, 2729-2734.
B-5 U.S. Nuclear Regulatory Commission, Inspection and Enforcement Informa-tion Notice No. 83-32, " Rupture of Americium-241 Source (s) Contained in a Well Logging Device," May 26, 1983.*
C-1 U.S. Nuclear Regulatory Commission, Inspection and Enforcement Information Notice No. 82-27, " Fuel Rnd Degradation Resulting from Baffle Water-Jet Impingement," August 5, 1982.*
C-2 U.S. Nuclear Regulatory Commission, Inspection and Enforcement Information Notice No. 82-29, " Control Rod Drive (CRD) Guide Tube Support Pipe Failures at Westinghouse PWRs," July 23, 1982.*
C-3 Letter from James G. Keppler, Regional Administrator, NRC Region III, to John D. Selby, President, Consumers Power Company, transmitting a Notice of Violation and Proposed Imposition of Civil Penalties, Docket Nos. 50-329 and 50-330, February 8, 1983.*
C-4 Letter from James W. Cook, Vice-President, Consumers Power Company, to J. G. Keppler, Regional Administrator, NRC Region III, Docket Nos. 50-329 and 50-330, January 10, 1983.*
- Available in NRC Document Room, 1717 H Street, NW, Washington, DC 20555, for inspection and copying (for a fee).
43
NRC r onu 335 U.S. NUCLE AR REGUL ATORY COMMGSION er in BIBLIOGRAPHIC DATA SHEET flVREG-0090, Vol. 6, flo.1 4 TlT L L AN D SUBTlT LL (A dd Volurne IVo, of appervraatel
- 2. (Leave blank)
Report to Congress on Abnormal Occurrences January-March 1983 3 HECIPIE N T'S ACCESSION NO
- 1. AU T HOHISI
- 5. D ATE HEPOHT COMPLE TE D l Yf AR MON TH September 1983 9 PE Hf OHMiNG ORGANIZATION N AME AND M AILING ADDHESS (include Ip Co*/
DATE HE POHT ISSUE D l YEAH MONTH Office for Analysis and Evaluation of Operational Data September 1983 U.S. fluclear Regulatory Comission 6 (te=,o<anas Washington, DC 20555 8 (Leave blank) 12 SPONSOHING ORGANIZ ATION N AME AND M AILING ADDRE SS I/nclu* lip Co*J 10 PHOJE CT T ASK.WOHK UNIT NO Office for Analysis and Evaluation of Operational Data
"'"^N U.S. fluclear Regulatory Comission Washington, DC 20555 13 T Y PE OF HE PO H T F's alon cov t HE D (loctus.ve ustest Quarterly January-March 1983 15 SUPPLEMENTARY NOTES 14 IL " *" "/ * * '
16 AUS T H AC T (200
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Section 208 of the Energy Reorganization Act of 1974 identifies an abnormal occurrence as an unscheduled incident or event which the fluclear Regulatory Commission determines to be sigi.ificant 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 January 1 to March 31, 1983.
During the report period, there were three abnormal occurrences at the nuclear power plants licensed by the flRC to operate. The first involved a niain feedwater line break due to water hammer. The second involved management and procedural control deficiencies. The third involved failure of the automatic reactor trip system. There were no abnormal occurrences for the other flRC licensees. There were six abnormal occurrences at Agreement State licensees. One involved an individual who inrjested and was contaminated by radio-active material. Four involved lost or stoleri radioactive sources. One involved radioactive contamination of a metals production facility.
The report also contains information updating some previously reported abnormal occurrences.
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