ML19256A980
| ML19256A980 | |
| Person / Time | |
|---|---|
| Issue date: | 12/31/1978 |
| From: | NRC OFFICE OF MANAGEMENT AND PROGRAM ANALYSIS (MPA) |
| To: | |
| References | |
| NUREG-0090, NUREG-0090-V01-N03, NUREG-90, NUREG-90-V1-N3, NUDOCS 7901170358 | |
| Download: ML19256A980 (26) | |
Text
N U R EG-0090 Vol.1, No. 3 REPORT TO CONGRESS ON ABNORMAL OCCURRENCES July - September 1978 p" "' coq, f w,..,, 'g g
%, +.....f Office of Management and Program Analysis U.S. Nuclear Regulatory Commission 7901170 359
Available from National Technical Information Service Springfield, Virginia 22161 Price:
Printed Copy $4.50 ; Microfiche $3.00 The price of this document for requesters outside of the North American Continent can be obtained from the National Technical Information Service.
NUR EG-0090 Vol.1, No. 3 REPORT TO CONGRESS ON ABNORMAL OCCURRENCES July - September 1978 i
Status as of November 30,1978 Date Published: December 1978 Office of Management and Program Analysis U.S. Nuclear Regulatory Commission Washington, D.C. 20555
iii 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, the fourteenth in the series, covers the period from July 1 to September 30, 1978.
The following incidents or events, including those submitted by the Agreement States, in that time period were determined by the Commission to be significant and reportable:
1.
There was one abnormal occurrence at the 70 nuclear power plants licensed to operate.
The event involved a degraded primary coolant boundary in a boiling water reactor.
2.
There were no abnormal occurrences at fuel cycle facilities (other than nuclear power plants).
3.
There were no abnormal occurrences at other licensee facilities.
4.
There were two abnormal occurrences reported by the Agreement States.
One involved an overexposure of a radiographer's assistant.
The second involved the theft of two radiography devices.
This report also contains information updating previously reported abnormal occurrences.
iv TABLE OF CONTENTS PAGE ABSTRACT..................................................
iii PREFACE...................................................
v INTRODUCTION.........................................
v THE REGULATORY SYSTEM................................
vi REPORTABLE OCCURRENCES...............................
vii AGREEMENT STATES.....................................
vii REPORT TO CONGRESS ON ABNORMAL OCCURRENCES, J U LY-S E PT EMB E R 19 78.....................................
1 NUCLEAR POWER PLANTS.................................
1 78-4 Degraded Primary Coolant Boundary in a Boiling Water Reactor..............
1 FUEL CYCLE FACILITIES (OTHER THAN NUCLEAR POWER PLANTS)............................................
6 OTHER NRC LICENSEES (INDUSTRIAL RADIOGRAPHERS, MEDICAL INSTITUTIONS, INDUSTRIAL USERS, ETC.)......
6 AGREEMENT STATE LICENSEES............................
6 AS78-3 Overexposure of a Radiographer's Assistant.............................
6 AS78-4 Theft of Two Radiography Devices...............................
8 APPENDIX A - ABNORMAL OCCURRENCE CRITERIA.................
10 APPENDIX B - UPDATE OF PREVIOUSLY REPORTED ABNORMAL OCCURRENCES................................
13 NUCLEAR POWER PLANTS.................................
13
v PREFACE INTRODUCTION The Nuclear Regulatory Commission reports to the Congress each quarter under provisions of Section 208 of the Energy Reorganizatian Act of 1974 on any abnormal occurrances involving facilities and activities regulated by the NRC.
An abnorme occurrence is defined in Section 208 as an unscheduled incident or event which the Commission determines is signifi-cant 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 promulgated in an NRC policy statement which was published in the Federal Register (42 FR 10950) on February 24, 1977.
In order to provide wide dissemination of information to the public, a Federal Register notice is issued on each abnormal occurrence with copies distributed to the NRC Public Document Room and all local public document rooms.
At a minimum, each such notice contains the date and place of the occurrence and describes its nature and probable consequences.
The NRC has reviewed Licensee Event Reports, licensing and enforcement action (e.g., violations, infractions, deficiencies, 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 Agreement States, described in this report meet the criteria for abnormal occurrence reporting.
This report, the fourteenth in the series, covers the period between July 1 -
September 30, 1978.
Events which occurred during this quarter and are later determined to be abnormal occurrences will be included in the next quarterly report.
Some events require considerable time and effort to analyze due to the complexity of situations where actual consequences are not readily apparent and additional facts are required.
Information reported on each event includes:
date and place; nature and probable consequences; cause or causes; and actions taken to prevent recurrence.
vi THE REGULATORY SYSTEM The system of licensing and regulation by which NRC carries out its responsibilities 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 ttandards and rules; issuing licenses and permits; inspecting for compliance; enforcing license requirements; and carrying on continuing evaluations, studies and research projects to improve both the regulatory process and the protection of the public health and safety.
Public participation is an element of the regulatory process.
In the licensing and regulation of nuclear power plants, the NRC follows the philosophy that the health and safety of the public are best assured through the establishment of multiple levels of protection.
These multiple levels can be achieved and maintained through reguletions which specify requirements which will assure the safe use of nuclear materials.
The regulations include design and quality assurance criteria appropriate for the various activities licensed by NRC.
An inspection and enforcement program helps assure compliance with the regulations.
Stringent require-ments for reporting incidents or events exist which help identify deficien-cies early enough to prevent serious consequences and aid in assuring that prompt and effective corrective action is taken to prevent their recurrence.
Most NRC licensee employees who work with radioactive materials are required to utilize personnel monitoring devices such as film badges or TLD (thermoluminescent cosimeter) badges.
These badges are rrocessed periodically and the exposure results normally serve as the official and legal record of the extent of personnel exposure to radiation during the period the badge was worn.
If an individual's past exposure history is known and has been sufficiently low, NRC regulations permit an individual in a restricted area to receive up to three rems of whole body exposure in a calendar quarter.
Higher values are permitted to the extremities or skin of the whole body.
For unrestricted areas, permissable levels of radiation are considerably smaller.
Permissible doses for restricted areas and unrestricted areas are stated in 10 CFR Part 20.
In any case, the NRC's policy is to maintain radiation exposures to levels as low as reasonably achievable.
vii 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 overexposures, radioactive material releases above prescribed limits, and malfunctions of safety-related equipment.
Thus, a reportable occurrence is any incident or event occurring at a licensed facility or related to licensed activities which NRC licensees are required to report to the NRC.
The NRC evaluates each reportable occurrence to determine the safety implications involved.
Because of the broad scope of regulation and the conservative attitude toward safety, there are a large number of events reported to the NRC.
The information provided in these reports is used in the NRC and the industry in their continuing evaluation and improvement of nuclear safety.
Most of the reports received from licensed nuclear power facilities describe events that did not directly involve the nuclear reactor itself, but involved equipment and components which are peripheral aspects of the nuclear steam supply system, and are minor in nature with respect to impact on public health and safety.
The majority are discovered during routine inspection and surveillance testing and are corrected upon discovery.
Typically, they concern single malfunctions of components or parts of systems, with redundant operable components or systems continuing to be available to perform the design function.
Information concerning reportable occurrences at facilities licensed or otherwise regulated by the NRC is routinely disseminated by NRC to the nuclear industry, the public, and other interested groups as these events occur.
Dissemination includes deposit of incident reports in the NRC's public document rooms, special notifications to licensees and other affected or interested groups, and public announcements.
In addition, a biweekly computer printout containing information on reportable events received from NRC licensees is sent to the NRC's more than 120 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,
viii source and special nuclear materials (in quantities not capable of sustaining a chain reaction).
Comparable and compatible programs are the basis for agreements.
Presently, information on reportable occurrences in Agreement State licensed activities is publicly available at the State level.
Certain information is also provided to the NRC under exchange of information provisions in the agreements.
NRC prepares a semiannual summary of this and other information in a document entitled, " Licensing Statistics and Other Data," which is publicly available.
In early 1977 the Commission determined that abnormal occurrences happening at facilities of Agreement State licensees should be included in the quarterly report to Congress.
The abnormal occurrence criteria included in Appendix A is applied uniformly to events at NRC and Agreement State licensee facilities.
Procedures have been developed and implemented and any abnormal occurrences reported by the Agreement States to the NRC are included in these quarterly reports to Congress.
REPORT TO CONGRESS ON ABNORMAL OCCURRENCES JULY-SEPTEMBER 1978 NUCLEAR POWER PLANTS The NRC i reviewing events reported at the 70 nuclear power plants licensed to operate during the third quarter of 1978.
Through the end of September, the NRC had determined that the following event was an abnormal occurrence.
78-4 Degraded Primary Coolant Boundary in a Boiling Water Reactor Preliminary information pertaining to this incident was published in the Federal Register (43 FR 35134).
Although this incident involves pipe cracking in a boiling water reactor and could be considered a part of Abnormal Occurrence 75-5 (See Appendix B of this report), it is being reported as a separate abnormal occurrence because this cracking resulted in a major degradation of the primary coolant pressure boundary (see Appendix A, Example 2 of "For Commercial Nuclear Power Plants" of this report).
Date and Place - On June 17, 1978, Iowa Electric Light and Power Company reported to the NRC an event at the Duane Arnold Power Plant, a boiling water nuclear plant located in Linn County, Iowa.
Nature and Probable Consequences - The reactor had shut down on June 17, 1978 due to an unrelated problem experienced during a surveillance test.
Prior to this unplanned shutdown, a primary coolant system leak of approximately three gallons per minute (gpm) from an unidentified source had been detected by the plant's leakage monitoring equipment.
Although this leak rate was within the technical specification limit of five gpm, the licensee took advantage of the unplanned shutdown to perform an inspection to identify the source of the leakage.
The leaking water was collected in the reactor building drain system and pumped to the plant's radioactive waste treatment system for processing.
During the inspection of the reactor coolant system piping, a leaking through-wall crack was found in a nickel-steel alloy (Inconel) fitting, called a " safe end," which is a short transition piece (approximately l
8 inches long) joining a section of primary coolant recirculation line piping to the reactor vessel (see Figures 1 and 2).
This recirculation line carries primary cooling water to two jet pumps located inside the reactor vessel.
There are eight identical recirculation inlet safe ends, each leading to two pumps which circulate cooling water in the
- reactor, I This section of the recirculation system, called a recirculation riser line, consists of 10-inch diameter piping.
Most of the recircu-lation system consists of 22-inch (nominal) diameter piping.
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The entire recirculation system is part of the primary coolant pressure boundary--i.e., the essential portions of the reactor coolant system--
which is one of several barriers to prevent the release of radioactive materials.
Plant operations are not permitted if this pressure boundary is degraded.
However, the facility has been designed and constructed to prevent a significant radioactive release to the environment even in the event of a design basis Loss-of-Coolant Accident (LOCA) resulting from a postulated instantaneous double-ended rupture of the largest diameter primary coolant pipir.g used in the facility.
The plant's emergency core cooling systems (ECCS) are designed to handle such a postulated failure, even assuming an independent single failure in the ECCS.
The through-wall crack was about 8 inches long or about one quarter the distance around the outside circumference of the safe end.
Radiography and ultrasonic testing showed that additional cracking extended about three quarters of the way around the interior circumference of the safe end.
Nondestructive testing (radiographic and ultrasonic) of the other seven identical safe ends revealed that all had indications of cracks or weld irregularities; however, these flaws did not penetrate to the surface of the safe end.
Since the leak could not be isolated from the reactor vessel, such as by closing a valve, special repair procedures were necessary.
All fuel was removed from the reactor vessel and placed in the spent fuel storage pool.
The leakage did not pose a threat to public health or safety.
Any leaking water was collected and processed by the radioagtive waste treatment system.
Water lost from the primary system was replaced by plant systems designed for that purpose.
Cause or Causes - All eight safe ends have been removed.
The licensee has sent the leaking safe end to a metallurgical laboratory for analysis, and the NRC has sent a second cracked safe end to another laboratory for independent metallurgical analysis.
The preliminary findings of both analyses indicate that the cracks were caused by intergranular stress-assisted corrosion of the furnace sensitized Inconel safe end.
The through-wall crack originated at a tight crevice formed by the fit up of the safe end and an internal thermal sleeve (see Figure 2), and then propagated outward.
Such crevices as known to enhance the possibility of stress corrosion cracking in an adverse chemical environment (i.e., in stagnant oxygenated water).
Ou-ing fabrication, the outsides of the safe ends were incorrectly machined and then weld repaired.
Although these repairs were properly reviewed and met the appropriate welding and piping codes, this weld repair may have contributed to the propagation of the cracks.
Further analysis to determine the role of the weld repair is being conducted as part of the continuing metallurgical analyses of the safe ends.
Actions Taken to Prevent Recurrence Licensee / Vendor - The licensee for the Duane Arnold facility is installing new safe ends of a differeat design intended to minimize high stress pc.nts.
The licensee has developed an extensive training and mockup p"ogram to qualify the welding techniques, equipment and welders to be used in replacing the safe ends.
Replacement activities are underway.
Brunswick Unit No. 1 and Unit No. 2 have a safe end design similar to that of Duane Arnold.
These units are located near Southport, North Carolina and are operated by Carolina Power and Light Company.
Safe ends at these units were also examined using nondestructive techniques during September 1978.
An ultrasonic testing (UT) indication was found in one safe end of Brunswick Unit No. 1.
The response level for this UT indication was acceptable under ASME Code rules (i.e., the indication was not of sufficient magnitude to require action).
Brunswick Unit No. 1 was shutdown in late November for additional UT inspections of this and three other safe ends.
For the former, no growth in the indication was found; nothing of significance was found in the other three safe ends examined.
No significant UT indications were found in the safe ends of Brunswick Unit No. 2.
NRC - NRC personnel are closely monitoring the replacement of the safe ends, including the qualification of the welding techniques, equipment and personnel.
NRC representatives are also following the metallurgical examinations of the two safe ends sent to separate laboratories for analysis.
The NRC is also reviewing other boiling water reactors to determine if any similar problems exist.
In addition, this event will be included in the review made by the NRC Pipe Crack Study Group (reference Item 75-5, Appendix B, of this report).
Future reports will be made as appropriate.
FUEL CYCLE FACILITIES (Other Than Nuclear Power Plants)
The NRC is reviewing events reported by these licensees during the third quarter of 1978.
Through the end of September, the NRC had not determined that any events were abnormal occurrences.
OTHER NRC LlCENSEES (Industrial Radiographers, Medical Institutions, Industrial Users, etc.)
There are currently more than 8,000 NRC nuclear material licenses in effect in the United States, principally for use of radioisotopes in the medical, industrial and academic fields.
Incidents were reported in this category from licensees such as radiographers, medical institutions, and byproduct material users.
The NRC is reviewing events reported by these licensees during the third quarter of 1978.
Through the end of September, the NRC had not determined 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 third quarter of 1978, the following Agreement State licensee events were detarmined reportable as abnormal occurrences.
AS78-3 Overexposure of a Radiographer's Assistant Appendix A (Example 1 of "For All Licensees") of this report notes that an exposure of the feet, ankles, hands or forearms of any individual to 375 rems or more of radiation can be considered an abnormal occurrence.
Date and Place - On August 15, 1978, the Louisiana Nuclear Energy Division notified the NRC of overexposures of radiographers and radiographers' assistants at Monroe X-Ray Company.
The overexposures occurred between June 20 and July 8, 1978.
Nature and Probable Consequences - The overexposures occurred during pipeline radiography on a barge about 100 miles off-shore in the Gulf of Mexico.
One radiographer's assistant was hospitalized for serious exposures to his hands.
The exposures produced blistering of the thumb and index finger of both hands with an additional slightly blistered area on the middle finger of the right hand.
He had complained of a tingling sensation in his hands on July 10, 1978 and it is considered that he received the dose during the period of June 20-July 8, 1978. The individual did have his dosimeter go off-scale while working on the barge and, on at least one occasion, had worked with an inoperable survey meter.
The whole body dose, as reported oy the film badge supplier, was about 5.5 rems.
From clinical indications, it is estimated that the hand dose ranged from 3,000 to 10,000 rems.
The individual remains under a doctor's supervision and will continue to receive medical attention as needed.
Whole body film badge reports for the other 6 to 7 radiography personnel involved ranged up to 6.1 rems.
Medical consultants were furnished to the State by Oak Ridge National Laboratory.
Cause or Causes - This is an unusual incident in that neither the individual involved nor the other members of the crew could recall any unusual circumstances that might account for the excessive exposure.
The State presumes that, while making exposures with a crank-out type exposure device, the radiographer's assistant handled the source tube without retrieving the source to the shielded position.
Actions Taken to Prevent Recurrence The investigation is complete and a hearing was conducted by the Louisiana Department of Conservation.
Formal notification of violation in the form of an Order was issued to the licensee and included assessment of a civil penalty in the amount of $1000.
The Order included the following findings:
(1) allowing an individual to receive an excessive dose, (2) allowing an individual to work in a radiation area without wearing a film badge, (3) performing radiography for at least one shift without an operable survey meter, (4) not immediately processing an individual's film badge when his dosimeter was discharged beyond its range, and (5) inadequate availability of training records.
The licensee's actions which have been taken for preventing recurrence will be documented in the response to the Order.
This incident is closed for purposes of this report.
AS78-4 Theft of Two Radiography Devices Appendix A (Example 6 of "For All Licensees") of this report notes that a substantiated case of actual theft of licensed material can be considered an abnormal occurrence.
Date and Place - On August 28, 1978, the Louisiana Nuclear Energy Division notified the NRC of the theft of two radiography devices from the Pittsburgh Testing Laboratory storage vault in Morgan City, Louisiana.
The theft occurred sometime between Thursday night, August 24, 1978 and Friday night, August 25, 1978.
Nature and Probable Consequences - The licensee informed the Louisiana Nuclear Energy Division of the theft on August 26, 1978.
The report was delayed until the management of the company could contact all employees who might have had the devices without filling out a proper utilization log.
The devices are Gamma Industries Model Century crank-out radiography cameras.
They contain approximately 39 curies and 33 curies of iridium-192, respectively.
The licensee said that the devices were locked and properly labeled when stolen.
The licensee indicated that the devices were taken from a locked storage vault and that no other equipment was disturbed Therefore, it is assumed that the individuals knew what they were looking for and probably were familiar with the handling of radioactive material.
However, if the devices feli into the hands of individuals not knowledgeable in the handling of radioactive sources, the possibility of personnel exposure cannot be dismissed.
Cause or Causes - The State has determined that the licensee did not contribute, i.e., by neglect or lack of security, to + he thef t of the devices.
Therefore, a cause has not yet been determined.
The incident is still under police investigation.
Actions Taken to Prevent Recurrence Licensee - Management of the co,pany contacted the Morgan City police department shortly after making contact with the Louisiana Nuclear Energy Division.
Louisiana Nuclear Energy Division - The Division issued a press release to the AP, UPI, and Louisiana Network.
The press release contained a warning that the devices could be dangerous if unlocked and that if found, local authorities should be notified at once.
The Division provided the serial numbers of the devices to manufacturers in the area in the event an effort is made to sell them or have the sources replaced.
In addition, the Division contacted regulatory agencies in adjacent states to notify them of the stolen equipment.
Future reports will be made as appropriate.
APPENDIX A ABNORMAL OCCURRENCE CRITERIA The following criteria for this report's abnormal occurrence determina-tions were set forth in an NRC policy statement published in the Federal Register (42 FR 10950) on February 24, 1977.
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 hea~..h or safety and could include but need not be limited to:
1.
Moderate exposure to, or release of, radioactive material licensed by or otherwise regulated by the Commission; 2.
Major degradation of essential safety-related equipment; or 3.
Major deficiencies in design, construction, use of, or manage-ment controls for licensed facilities or material.
Examples of the types of events that are evaluated in detail using these criteria are:
For All Licensees 1.
Exposure of the whole body of any individual to 25 rems or more of 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 Part 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 Part 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 Part 20 (10 CFR Part 20.403(b)).
4.
Radiation or contamination levels in excess of design values en 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 cf a package containing the radioactive material, or (b) release of radioactive material from a package in amounts greater than the regulatory limit (10 CFR Part 71.36(a)).
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 substantiated inventory discrepancy which is judged to be significant relative to normally expected petformance 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 Part 70.52(a)).
w.
A major deficiency in design, construction or operation heving safety implications requiring immediate remedial action.
11.
Serious deficiency in management or procedural controls in major areas.
12.
Series of esents (where individual events are not of major importance), recurring incidents, and incidents with implica-tions for similar facilities (generic incidents), which create major safety concern.
For Commercial Nuclear Power Plants 1.
Exceeding a safety limit of license Technical Specifications (10 CFR Part 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 Part 100 guidelines could result from a postulated tran-sient 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 require immediate remedial action.
5.
Personnel error or procedural deficiencies which result in loss of plant capability to perform essential safety functions such that a potential release of radioactivity in excess of 10 CFR Part 100 guidelines could result from a postulated transient or accident (e.g., loss of emergency core cooling system, loss of control rod systems).
For Fuel Cycle Licensees 1.
A safety limit of license Technical Specifications is exceeded and a plant shutdown is required (10 CFR Part 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 confine-ment system to perform its designated function.
APPENDIX B UPDATE OF PREVIOUSLY REPORTED ABNORMAL OCCURRENCES During the July through September 1978 period, the NRC, NRC 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 occurrences.
The referenced Congressional abnormal occurrence reports below provide the initial and updating information on these abnormal occurrences.
Those occurrences not now considered closed will be discussed in subsequent ceports in the series.
NUCLEAR POWER PLANTS The following abnormal occurrence was originally reported in NUREG-75/090,
" Report to Congress on Abnormal Occurrences:
January-June 1975," and updated in subsequent reports in this series, i.e., NUREG-0090-1, 2, 3, and 9.
It is further updated as follows:
75-5 Cracks in Pipes at Boiling Water Reactors (BWRs)
Pipe cracking has occurred in the heat affected zones of welds in primary system piping in BWRs since the mid 1960s.
These cracks have occurred mainly in Type 304 stainless steel that is being used in most operating BWRs.
The major problem is recognized to be intergranular stress corrosion cracking (IGSCC) of austenitic stainless steel components that have been made susceptible to this failure mode by being " sensitized" either by post weld heat treatment or by sensitization of a narrow heat affected zone near welds.
" Safe ends" (short transition pieces between vessel nozzles and the piping) that have been highly sensitized by furnace heat treatment while attached to vessels during fabrication were found to be susceptible to IGSCC in the late 1960s.
Because they were susceptible to cracking, the Atomic Energy Commission took the position in 1969 that furnace sensitized safe ends should not be used in the future.
Most of the furnace sensitized safe ends in older plants have been removea or clad with a protective material, and there are only a few BWRs that still have furnace sensitized ends in use.
Most of these, however, are in smaller diameter lines and are subjected to inservice inspection (ISI).
Earlier reported cracks (prior to 1975) occurred primarily in 4" diameter recirculation loop-by pass lines and in 10" diameter core spray lines.
More recently cracks were discovered in recirculation riser piping (12" to 14") in foreign plants.
Cracking is most often detected during Inservice Inspection using ultrasonic testing techniques.
Some piping cracks have been discovered as a result of primarv coolant leaks.
Because of these occurrences of BWR primary system cracking, there has been a wide variety of actions undertaken by the NRC.
These actions included:
issuance of Regulatory Guide 1.44 on " Control of the Use of Sensitized Stainless Steel;"
issuance of Regulatory Guide 1.45 on " Reactor Coolant Boundary Leak Detection Systems;"
closely following the incidence of cracking in BWRs, including foreign experience; encouraging replacement of furnace sensitized safe ends; requiring augmented inservice inspection of lines having less corrosion resistant stainless steel, especially those that have a high potential for cracking (service sensitive lines);
requiring upgrading of leak detection systems.
Pipe cracking and furnace sensitized safe end cracking have been recently reported in larger (24" diameter) lines in a GE-designed BWR in Germany with over ten years of service.
Because the safe ends in that facility had been furnace sensitized during fabrication, IGSCC was suspected.
As a result of concerns regarding these furnace sensitized safe ends, a safe end was removed in order to perform destructive examination.
During laboratory examination of the removed safe end, including a small section of attached pipe, cracks were discovered at various locations in the safe end and in the weld heat affected zone of the pipe.
The cracks in the pipe weld area were very shallow with the maximum depth less than 5 mm (about 1/8") in a wall thickness of about 1.5 inches.
Cracking in the furnace sensitized safe end, also having a wall thickness of about 1.5 inches, was somewhat deeper.
The German experience was the first known occurrence of IG5CC in pipes as large as 24" in diameter.
In June 1978, a through-wall crack was discovered in an Inconel recircula-tion riser safe end (10" diameter) at the Duane Arnold facility (see Abnormal Occurrence 78 4 of this report).
The crack has been attributed to IGSCC, although the material in this instance is different fram the type 304 stainlec; steel that has been historically found to be susceptible to IGSCC.
Prior to safe end removal, ultrasonic examination showed several indications of possible cracks.
Following their removal, cracking was discovered in all eight safe ends.
The cracking appeared to have originated in a tight crevice between the inside wall of the safe end and an internal thermal sleeve.
Such crevices are known to enhance IGSCC.
Differences in materials, geometry, stress levels and crevices appear to make the problem at Duane Arnold unique to a particular type of recirculation riser safe end (Type I).
As a result of this event, ultrasonic examination of the other Type I safe ends in U.S. BWRs, i.e.,
at the Brunswick 1 and 2 facility, was conducted.
No significant indications were found in Unit 2, and one indication was identified at Unit 1.
Although this is relatively minor and is not " reportable" pursuant to the NRC Regulations, evaluation is continuing.
The ultrasonic indication which was found was reevaluated at another plant shutdown in November 1978 (see page 5 of this report).
In addition to discussions with General Electric (the reactor vendor) regarding recent pipe cracking experience, General Electric has been asked to provide an in-depth report on the significance of recent events regarding current inspection, repair, and replacement programs.
They were also asked to address any new safety concerns related to the occurrence of cracking in large main recirculation piping.
Based on information presented by General Electric to date, and extensive staff evaluation, it has been concluded that the recent occurrences do not constitute a basis for immediate concern about plant safety, nor require any new immediate actions by licensees.
The staff briefed the Commission on pipe cracking in BWRs on August 31, 1978, and re-established an NRC Pipe Crack Study Group on September 14, 1978.
The Study Group will specifically address the following issues:
the significance of the cracks discovered in large diameter pipes relative to the conclusions and recommendations set forth in the referenced report and in its implementation document, NUREG-0313; resolution of concerns raised over the ability of ultrasonic techniques to detect cracks in austenitic stainless steel',
the significance of the cracks found in large diameter sensitized safe ends, and any recommendations regarding the current NRC program for dealing with this matter; the potential for stress corrosion c. racking in PWRs; and the significance of the safe end cracking at Duane Arnold relative to similar material and design aspects at other facilities.
The study group is scheduled to complete its evaluation and report in January 1979.
In addition to the study grtup effort, the NRC has underway generic technical review efforts regarding flaw detection, aimed at improving piping inspection techniques and requirements.
Further reports will be made as appropriate.
The following abnormal occurrence was originally reported in NUREG-0090-3,
" Report to Congress on Abnormal Occurrences:
January-March 1976," and updated in subsequent reports in this series, i.e, NUREG-0090-4, 6, and Vol. 1, No. 1.
It is further updated as follows:
76-1 Deficiencies in the Mark I Containment Systems of Certain Boiling Water Reactors Actions Taken to Prevent Recurrence Licensee / Vendors - General Electric Company (GE) and the Mark I Owners Group are continuing to conduct the Mark I Containment Long Term Program (LTP).
The objective of the LTP are (1) to establish design basis (conservative) loads that are appropriate for the anticipated life (40 years) of each Mark I BWR facility, and (2) to restore the original intended design safety margins for each Mark I containment system.
The LTP consists of a series of major tasks and subtasks which are designed to provide a detailed basis for hydrodynamic load definition and the methodology and acceptance criteria for the structural assessments.
The generic aspects of the LTP will be described in a Plant Unique Analysis Applications Guide, scheduled to be completed in October 1978, and a Load Definition Report, scheduled to be completed in December of 1978. Subsequently, each utility will perform a plant-unique analysis using approved load definition and structural analysis techniques to demonstrate conformance with the LTP structural acceptance criteria.
These analyses are currently scheduled for completion in October 1979.
The scheduled completion date for the Mark I LTP, including the issuance of license amendments and the implementation of any plant modifications necessary to satisfy the LTP structural acceptance criteria, is December 1980.
To maintain this schedule, a number of utilities have undertaken plant modifications prior to the completion of their plant unique analysis.
This action has been considered necessary to minimize the potential for extended plant outages later in the program.
Similarly, modifications to coniponents external to the containment (e.g., support structures) have and are being conducted during normal plant operatien.
In the Vermont Yankee facility, the utility attempted to install gussets on the suppression chamber to support column attachments during normal plant opeiation.
Examination of the gussett revealed cracking in the new attachment welds which were subsequently determined to exceed the maximum allowable depth.
This led to a voluntary unscheduled shutdown of the plant by the utility to repair the cracked welds.
The planned modifications were removed and the containment was restored to its original condition to permit the plant to return to power.
Similar strengthening modifica-tions have been performed successfully during normal plant operation by other utilities.
Future reports will be made as appropriate.
NRC 7 oeu 33s U.s. NUCLE AF4 HEGUL ATORY COMMISSION fJUREG-0090 BIBLIOGRAPHIC DATA SHEET Vol. 1, No. 3 4 Tl TLE AN D SUUTI T L E (A dd Votume too, si appemenares
- 2. (Leave boeki Report to Congress on Abnormal Occurrences July-September 1978 3 RE CIPIEN T'S ACCESSION NO.
- 7. At lT HOHIS)
- 5. D ATE REPOH T COUPL E TE D l YEAR MONTH November 1978
- 9. PE HF OHMING ORGANIZATION N AME AND M AILING ADDHESS (include /,p Codel DATE REPORT ISSUED U.S. Nuclear Regulatory Comission Dece ber 1978 Office of Management & Program Analysis 6 " " "'" '#
Washington, D.C.
20555
\\ 8 (Leave blank)
- 12. SPONSOHING OHGANIZ ATION N AME AND MAILING ADDRE SS (/nclude le Code)
U.S. Nuclear Regulatory Commission Office of Management & Program Analysis
- 11. CONT R ACT NO.
Washington, D.C.
20555
Quarterly July-September 1978
- 15. SUPPLEMENTARY NOTE S
- 14. (Leave blank)
- 16. ABSTR ACT 200 words or less)
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, the fourteenth in the series, covers the period July 1 to September 30, 1978.
During this period, there were three abnormal occurrences. One occurred at one of the 70 nuclear power plants licensed to operate; it involved a degraded primary coolant boundary in a boiling water reactor. The other two occurred at Agreement State licensees
- one involved an overexposure of a radiographer's assistant and the other involved the theft of two radiography devices.
This report also contains information updating previously reported abnormal occurrences.
- 17. KE Y WORDS AND DOCUME NT AN ALYSIS 17a. DESCRIPTORS 17b. (DENTIFIE RS/OPEN EN DE D TERMS
- 19. SE CURITY, CLASS (Th,s reporr/
- 21. NO. OF P AGc :
Unclassified No restriction on availability.
- 3'n#lRiTY cf AS*d t rh,s p,s,1
- 22. P RICE c assi ie s
PeRC FORM 335 17 77)