ML20041C519
| ML20041C519 | |
| Person / Time | |
|---|---|
| Issue date: | 01/31/1982 |
| From: | NRC OFFICE FOR ANALYSIS & EVALUATION OF OPERATIONAL DATA (AEOD) |
| To: | |
| References | |
| NUREG-0090, NUREG-0090-V04-N03, NUREG-90, NUREG-90-V4-N3, NUDOCS 8203020158 | |
| Download: ML20041C519 (25) | |
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NUREG-0090 Vol. 4, No. 3 Report to Congress on Abnormal Occurrences July - September 1981 U.S. Nuclear Regulatory Commission Office for Analysis and Evaluation of Operational Data
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Available from GP0 Sales Program Division of Technical Information and Document Control U.S. fluclear Regulatory Commission Washington, DC 20555 Printed copy price:
$3.00 and National Technical Information Service Springfield, VA 22161
NUREG4090 Vol. 4, No. 3 I
. Report to Congress on Abnormal Occurrences July - September 1981
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Date Published: January 1982 Office for Analysis and Evaluation of Operational Data U.S. Nuclear Regulatory Commission Washington, D.C. 20555 7
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U. S. NUCLEAR REGULATORY COMMISSION Previous Reports in Series for the Report to Congress on Abnormal Occurrences NUREG 75/090, January-June 1975, NUREG-0090, Vol.1, No.3, July-Septembcr 1978, published October 1975 published December 1978 NUREG-0090-1, July-September 1975, NUREG-0090, Vol.1, No.4, October-December 1978, published March 1976 published March 1979 NUREG-0090-2, October-December 1975, NUREG-0090, Vol. 2, No.1, January-March 1979, published March 1976 published July 1979 NUREG-0090-3, January-March 1976, NUREG-0090, Vol. 2, No. 2, April-June 1979, published July :976 publish.a November 1979 NUREG-0090-4, April-June 1976, NUREG-0090, Vol. 2, No.3, July-September 1979, published October 1976 published February 1980 NUREG-0090-5, July-Septmber 1976, NUREG-0090, Vol. 2, No.4, October-December 1979, published March 1977 published April 1980 NUREG-0090-6, October-December 1976, NUREG-0090, Vol.3, No. h January-March 1980, published June 1977 published September 1980 NUREG-0090-7, January-March 1977, NUREG-0090, Vol. 3, No. 2, April-June 1980, published June 1977 published Novcmber 1980 NUREG-0090-8, April-June 1977, NUREG-0090, Vol.3, No.3, July-September 1980, published September 1977 published February 1981 NUREG-OO90-9, July-September 1977, NUREG-0090, Vol.3, No.4, October-December 1980, published November 1977 p blished May 1981 NUREG-0090-10, October-Deccmber 1977, NUREG-0090, Vol.4, No.1, January-March 1981, published March 1978 published July 1981 NUREG-0090, Vol.1, No.1, Jan-March 1978, NUREG-0090, Vol. 4, No. 2, April-June 1981, published June 1978 published October 1981 NUREG-0090, Vol.1, No. 2, April-June 1978, published September 1978
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l ABSTRACT Section 208 of the Energy Reorganization Act of 1974 identifies an abnormal occurrence as an unscheduled incident or event which the Nuclear Regulatory Commission determines to be significant from the standpoint of public health or safety and requires a quarterly report of such events to be made to Congress.
This report covers the period from July 1 to September 30, 1981.
The report states:
1.
There were two abnormal occurrences at the nuclear power plants licensed to operate.
One involved a misalignment of a high head safety injection isolation valve.
The other involved a failure of the high pressure safety injection system.
2.
There were two abnormal occurrences at other licensee facilities.
Both involved calculated radiation exposures in excess of 10 CFR 20 limits.
3.
There were two abnormal occurrences reported by the Agreement States.
The first involved excessive radiation doses to hospital patients.
The second involved overexposure of a radiographer and two barge crew members.
The report contains information updating some previously reported abnormal occurrences.
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CONTENTS PAGE ABSTRACT........................................................
iii PREFACE......................................................
vii INTRODUCTION vii THE REGULATORY SYSTEM.....................................
vii REPORTABLE OCCURRENCES.....................................
viii AGREEMENT STATES...........................................
ix REPORT TO CONGRESS ON ABNORMAL OCCURRENCES, JULY-SEPTEMBER 1981....
1 NUCLEAR POWER PLANTS........................................
1 81-3 Misalignment of High Head Safety Injection Isolation Valve 1
81-4 Failure of High Pressure Safety Injection System....
3 FUEL CYCLE FACILITIES (Other than Nuclear Power Plants).......
7 OTHER NRC LICENSEES (Industrial Radiographers, Medical Institutions, Industrial Users, Etc.)...............
7 81-5 Calculated Radiation Exposures Exceeding 10 CFR 20 Limits.....
7 81-6 Calculated Overexposure in an Unrestricted Area............................................
9 AGREEMENT STATE LICENSEES....................................
11 AS81-1 Excessive Radiation Doses to Hospital Patients......................................
11 AS81-2 Overexposure of a Radiographer and Two Barge Crew Members..................................
13 REFERENCES.........................................................
17 APPENDIX A - ABNORMAL OCCURRENCE CRITERIA..........................
19 APPENDIX B - UPDATE OF PREVIOUSLY REPORTED ABNORMAL OCCURRENCES 23 NUCLEAR POWER PLANTS..........................................
23 APPENDIX C - OTHER EVENTS OF INTEREST..............................
29 REFERENCES (FOR APPENDICES)........................................
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l PREFACE
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INTRODUCTION The Nuclear Regulatory Commission reports to the Congress each quarter under provisions of Section 208 of the Energy Reorganization Act of 1974 on any abnormal occurrences involving facilities and activities regulated by the NRC.
l 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.
af Events are currently identified as abnormal occurrences for this report by the NRC using the criteria delineated in Appendix A.
These criteria were promul-r; 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 and probable consequences.
The NRC has reviewed Licensee Event Reports, licensing and enforcement actions (e.g., notices of violations, civil penalties, license modifications, etc. ),
generic issues, significant inventory differences involving special nuclear material, and other categories of information available to the NRC.
The NRC has determined that only those events, including those submitted by the
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Agreement States, described in this report meet the criteria for abnormal occurrence reporting.
This report covers the period between July 1 to September 30, 1981.
7 Information reported on each event includes:
date and place; nature and JP 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 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, evalua-tion 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 estab-lishing standards and rules; issuing licenses and permits; inspecting for compliance; enforcing license requirements; and carrying on continuing evalua-tions, 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 s
viii 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 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 l
changes needed to improve safety.
Some deficiencies in design, operation and l
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 tele-phone lines to the NRC for rapid communication in the event of any incident.
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) bauges.
These badges are processed periodically and the exposure results normally serve as the official and legal record of the extent of personnel exposure to radiation during the period the badge was worn.
If an individual's past exposure history is known and has been sufficiently low, NRC regulations permit an individual in a restricted area to rereive up to three rems of whole body exposure in a calendar quarter.
Higher values are permitted to the extremities or skin of the whole body.
For unrestricted areas, permissible levels of radiation are considerably smaller.
Permissible doses for restricted areas and unrestricted areas are stated in 10 CFR Part 20.
In any case, the NRC's policy is to maintain radiation exposures 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 cver-exposures, 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.
ix 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 by the NRC and the industry in their continuing evaluation and improvement of nuclear safety.
Some of the reports describe events that have real or potential safety implications; however, most of the reports received from licensed nuclear power facilities describe events that did not directly involve the nuclear reactor itself, but involved 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.
Many are discovered during routine inspection and surveillance testing and are corrected upon discovery.
Typically, they concern single salfunctions of compo-nents or parts of systems, with redundant operable components or systems con-tinuing 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 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, source and special nuclear materials (in quantities not capable of sustaining a chain reaction).
Com-parable and compatible programs are the basis for agreements.
Presently, information on reportable occurrences in Agreement State licensed activities is publicly available at the State level.
Certain information is also provided to the NRC under exchange of information provisions In the agree-ments.
NRC prepares a semiannual summary of this and other information in a document entitled, " Licensing Statistics and Other Data," which is publicly available.
In early 1977, the Commission determined that abnormal occurrences happening at facilities of Agreement State licensees should be included in the quarterly report to Congress.
The abnormal occurrence criteria included in Appendix A is applied uniformly to events at NRC and Agreement State licensee facilities.
Procedures have been developed and impleraented and 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 1981 NUCLEAR POWER PLANTS The NRC is reviewing events reported at the nuclear power plants licensed to operate during the third calendar quarter of 1981.
As of the date of this report, the NRC had determined that the following were abnormal occurrences.
81-3 Misalignment of High Head Safety Injection Isolation Valve Preliminary information pertaining to this event was reported in the FEDERAL REGISTER (Ref. 1).
Appendix A (one of the general criteria) of this report notes that major degradation of essential safety-related equipment can be con-sidered an abnormal occurrence.
Date and Place - On June 6, 1981, the NRC was notified by Duquesne Light Company (the licensee) that a manually-operated high head safety injection isolation valve, which should normally be open during routine reactor power operation, had been found closed at Beaver Valley Power Station Unit 1.
The Beaver Valley Unit 1 plant utilizes a pressurized water reactor and is located in Beaver County, Pennsj!vania.
Nature and Probable Consequences - On June 6, 1981, with the reactor at 99%
power, an operator making a routine early morning inspection (to verify the correct position of specific Engineered Safety Features equipment) found that a manually-operated suction isolation valve (SI-26) in the emergency core cool-ing system was closed and the chain and padlock that were supposed to secure the valve in the open position were missing.
The valve, which is to be checked by operators once each shift, was immediately reopened, chained, and locked into position.
The licensee was also aware of a different but similar event that occurred on June 5, 1981.
The latter event concerned three auxiliary feed-water pumps' manual suction isolation valves being found unlocked.
The chains and locks, which are one element of the administrative controls to ensure the valves are correctly positioned, were missing and could not be found.
The difference between the two events is that the auxiliary feedwater valves were found to be in their proper position (open).
The safety implication due to the closure of valve SI-26 is associated with the loss of automatic high-head safety injection (HHSI) capability.
With the valve shut, emergency core cooling water from the refueling water storage tank would not have been available automatically to the three HHSI pumps for high pressure injection of water into the core under emergency conditions.
Manual action by an operator would have been required for the system to complete its intended function through this or an alternate flow path.
It should be noted that with the suction valve shut, the high head safety injection pumps could possibly have been damaged had they been operated.
Although the event did not result in any adverse effects on the health of the public or licensee personnel, it did represent a major degradation of essential safety-related equipment
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designed to mitigate the consequences of a n.ajor occurrence such as a loss-of-coolant accident.
This event, in combi 1ation with the June 5, 1981 event, also raised concern for the possibility of criminal acts or sabotage.
Cause or Causes - Careful consideration of available information has led the NRC to conclude that the mispositioned valve and missing chains and padlocks were possible acts of sabotage, rather than operator errors.
Actions Taken to Prevent Recurrence Licensee - When the SI-26 valve was found closed, it was immediately opened, chained, and locked.
The licensee initiated an investigation of the events including requesting FBI assistance.
Access controls to vital areas were strengthened by implementing interim emergency procedures, including restric-tions on personnel access and movement control within the plant.
The normal Engineered Safety Features (ESF) equipment position checks were supplemented by special tours during each shift to make special verification of ESF equip-ment.
The frequency of security tours of specific vital areas was increased.
Stronger chains were installed on all ESF equipment and better accountability and control procedures were instituted for all vital equipment padlocks.
The licensee maintained these enhanced safety and security measures in force for an extended period of time, after which the licensee returned to a program upgraded to strengthen the control of plant activities.
NRC - Investigations by both the NRC and the FBI were initiated.
The purposes of the NRC investigation were (1) to assure that there were no additional unde-tected incidents of tampering with safety-related equipment that could impact on continued safe operation of the reactor or endanger the health and safety of plant employees or the public and (2) to determine the details and sequence of events surrounding the incidents of June 5 and 6, 1381.
The FBI investiga-tion was to determine if an act of sabotage had been coamitted and, if so, who had committed the act.
Immediate Action Letter No. 81-25 was issued on June 9, 1981 (Ref. 2) confirming the actions taken by the licensee, as stated above, with the stipulation that these measures will remain in effect until notification to remove these controls is issued from the NRC.
An on-site assessment of the safety and security pro-gram was performed on August 17-21, 1981.
The purpose of this assessment was to confirm (1) that vital plant areas and equipment were being ptotected, and (2) that an acceptable combination of security and safety programs and pro-cedures was in effect to ensure that public safety was not compromised.
Based on this assessment, the NRC staff has approved relaxing, somewhat, the more stringent interim emergency procedures that were implemented shortly after the event occurred.
The NRC investigation was completed and documented in an investigative report dated December 10, 1981 (Ref. 3).
Based on this investigation, the NRC issued a notice of violation identifying four violations of the licensee's safety related commitments.
Additionally, the investigation identified two generic
3 procedural concerns which may have contributed to the June 5 and 6 events:
(1) procedures did not assure timely withdrawal of access authorizations of individuals being terminated under adverse circumstaces and, (2) the criteria for authorizing unescorted access to vital areas were not sufficiently selective.
This incident is closed for purposes of this report.
81-4 Failure of High Pressure Safety Injection System Preliminary information pertaining to this event was reported in the FEDERAL REGISTER (Ref. 4).
Appendix A (one of the general criteria) of this report rotes that major degradation of essential safety-related equipment can be con-sidered an abnormal occurrence.
Date and Place - On September 3, 1981, the NRC was notified by Southern California Edison Company (the licensee) of a condition involving failure of two safety injection valves to open upon receipt of a safety injection signal at the San Onofre Nuclear Generating Station Unit 1.
The San Onofre Unit 1 plant utilizes a pressurized water reactor and is located in San Diego County, California.
Nature and Probable Consequences - At about 3:30 AM on September 3, with the reactor at 88% power, one of the regulated power supplies serving one of the two redundant paths of the Reactor Protection System and a portion of the Con-trol and Indication System failed.
As a result, feedwater and steam flow and steam generator water level indications were lost for the steam generator served by this power supply and oscillations were observed in similar flow and level indications of the other two steam generators.
The operators placed the steam generator water level controls under manual control and then manually tripped the reactor.
During this period, high steam generator water levels resulted in increased cooling of the Reactor Coolant System (RCS), thereby reducing the system pressure.
At 1735 psig, a Safety Injection Actuation Signal (SIAS) was automatically initiated, in accordance with system design.
Subsequently, RCS pressure increased, SIAS was reset at 4:00 AM, and safety injection terminated.
At San Onofre Unit 1, the two steam generator feedwater pumps are also part of the safety injection system (see Figure 1).
During normal operation, water from the condensate pumps flows through valves HV854 A and B, the feedwater pumps, and valves HV852 A and B to the secondary side of the steam generators.
Another line, which connects to each feedwater pump discharge line is normally isolated by safety injection valves HV851 A or B; these lines join to a common header which then connects to the three cold leg return lines from the three steam generators to the reactor vessel.
Upon receipt of a SIAS, the feedwater pumps are automatically valved out of the steam generator feedwater system, by closing valves HV854 A and B and HV852 A and B (steam generator feedwater is then supplied by the Auxiliary Feedwater System), and become part of the safety
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FROM CONDENSATE f' TO STEAM PUMPS '
GENERATORS V
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HV 853A HV 851 A MOV 850C LOOP C FEEDWATER PUMP A COLD G
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Figure 1. Simplified Schematic of Safety injection System (San Onofre Unit 1)
5 injection system by opening the HV851 A and B and HV853 A and B valves.
The safety injection valves, HV851 A and B, downstream of the feedwater pumps, are designed to automatically open upon receipt of the SIAS.
Then, if RCS system pressure decreases sufficiently (such as in the case of a loss-of-coolant acci-dent), borated water (normally from the Refueling Water Storage Tank) would be pumped by the safety injection pumps, through the feedwater pumps and safety injection valves, into the RCS.
During the September 3, 1981 event, RCS pres-sure only decreased to about 1700 psi.
Therefore, since the feedwater pump discharge shutoff head pressure is about 1200 psi by design, no safety injec-tion flow would have occurred even if the safety injection valves had opened.
In addition, none was actually required since system pressure and inventory were adequate throughout the event.
However, while SIAS was present, plant operators in the Auxiliary Building noted that neither valve HV851 A nor HV851 B had opened.
Later, after the event was terminated, the valves did open when the feedwater pumps were tripped.
Since operability of the valves could not be demonstrated by subsequent tests, the licensee placed the plant in a cold shutdown condition and agreed not to restart the plant without NRC concurrence.
For this event, the failure of both these valves to open did not result in plant damage, radiological releases, or health effects.
The injection of borated water into the RCS was not required by the plant conditions present.
- However, had a loss-of-coolant or steamlin break accident occurred, one of the key safety systems designed to provide mitigation may not have performed as intended.
Cause or Causes - To date, the following valve design deficiencies have been identified:
(1) the contact stress between the valve seat and disc exceeded the threshold of galling (transfer of metal at contact surfaces resulting from local material being overstressed), (2) the coefficient of friction assumed for sizing the hydraulic valve activators was too small, and (3) the lack of any valve bonnet leak-off potentially allowed for an excessive differential pressure across both discs if the feedwater pumps were tripped.
As noted above, the valves did not open until the feedwater pumps were tripped and the differential pressure across the valve gates had significantly decayed.
Testing indicated that valve packing tightness was not a factor.
The valves were disassembled and the seating surfaces inspected.
Maintenance was performed on the valve seats to remove galling which was observed on the valve discs anc additional tests were performed.
Although valve operation was noted in some instances while the feedwater pumps were running, reliable operation within the acceptance criterion could not be demonstrated.
The valves had been regularly tested at each refueling outage, but the tests were performed with no significant pressure differential across the valves.
In 1976, the valve operators were changed from a motor operator to a hydraulic operator to decrease valve stroke time.
As indicated above, the hydraulic operators were selected based on a coefficient of friction that was too low.
Proper valve operations were demonstrated during the acceptance tests conducted with a design pressure difference of up to 1350 psi across the valves.
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6 restart in 1977 and September 3, 1981, no safety injection system challenges occurred--the only operation of the valves was during tests with little or no pressure differential.
No major valve maintenance was performed during this period.
Actions Taken to Prevent Recurrence Licensee - Meetings were held on September 15, 1981, September 30, 1981, and October 22, 1981, with the NRC to discuss possible resolution of the problem.
The licensee is continuing to evaluate the problem and possible corrective actions.
For the interim period, the licensee has altered the sequence for safety injection, installed valve bonnet leak-off lines, and has completed an extensive preoperational valve and system test program.
In addition, the licensee has agreed to institute a surveillance program which is more extensive, in terms of the frequency of testing and the detail of measurement, than required for any other_ plant.
The licensee has also committed to replace the valves in ques-tion and their actuators. The plans and schedule for the valve and actuator replacement are being provided to the NRC.
In addition, the licensee has commit-ted to perform a long-term study of the advisability of completely redesigning the safety injection system.
NRC - The NRC's resident inspector followed the licensee's inspection and corrective actions onsite.
The NRC's Offices of Inspection and Enforcement and Nuclear Reactor Regulation have reviewed and evaluated the licensee's pro-posed actions and the licensing submittal addressing the causes of the event and the proposed corrective actions including Emergency Core Cooling System modifications and testing.
The licensee has submitted a request to the NRC to amend the license to incor-porate surveillance tests of the valves, and the submittal has been reviewed and approved.
l The NRC is also evaluating the generic implications of the event, i.e., testing the valves under operating pressure differentials.
The problem remains under active review within the NRC.
The NRC issued Inspection and Enforcement Infor-mation Notice No. 81-31 (Ref. 5) on October 7, 1981 to licensees to inform them of this event.
The NRC agreed that the licensee's interim corrective actions were satisfactory and on November 1,1981 concurred with the licensee that the plant could be restarted.
The licensee initiated startup of the plant late on November 3, 1981.
Further reports will be made as appropriate.
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i FUEL CYCLE FACILITIES r
1 (Other than Nuclear Power Plants) i The NRC is reviewing events reported by these licensees during the third calendar quarter of 1981. As of the date of this report, the NRC has not determined that any were abnormal occurrences.
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OTHER NRC LICENSEES (Industrial Radiographers, Medical Institutions, l
Industrial Users, etc.)
There are currently more than 8,000 NRC nuclear material licenses in effect in i
the United States, principally for use of radioisotopes in the medical, indus-trial, and academic fields.
Incidents were reported in this category from j
licensees such as radiographers, medical institutions, and byproduct material users.
The NRC is reviewing events reported by these licensees during the third calendar quarter of 1981.
As of the date of this report, the NRC had determined that the following were abnormal occurrences.
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81-5 Calculated Radiation Exposures Exceeding 10 CFR 20 Limits Preliminary information pertaining to this event was reported in the FEDERAL REGISTER (Ref. 6).
Appendix A (Example 2 of "For All Licensees") of this report notes that 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)) can be considered an abnormal occurrence.
Date and Place - On May 5, 1981, the Eveleth Expansion Company of Eveleth, Minnesota reportea that between April 3 and 7, 1981, personnel working inside i
an iron ore pellet cooler had been exposed to radiation from a 10-curie cesium-137 sealed source contained in a level control gauge.
Nature and Probable Consequences - The Pellet Plant cooler had been shut down for repairs on March 30, 1981.
On that day, the shutter mechanism of the level control gauge, containing a nominal 10-curie cesium-137 sealed source was locked in the closed position.
Radiation surveys performed at that time appeared to indicate that the source was properly shielded.
After a cooldown period, work-men entered the cooler on April 3,1981, to replace refrautory material on the cooler walls.
On April 7, licensee personnel discovered that there were radia-tion levels in excess of 100 mr/hr within the cooler (later determined to be up to 2.2 rems /hr where the radiation beam entered the cooler).
It was deter-mined that several individuals had been exposed to a radiation beam from the source during the working days between April 3 and 7,1981.
The gauge source holder was removed from its mounting and licensee personnel l
found that the lead shielding in the shutter had melted and drained from the shield location.
This rendered the shielding integrity of the shutter useless.
8 Investigation showed that 17 licensre personnel and 14 contractor personnel had entered the cooler between Aprii 3 and 7, 1981.
Radiation levels in the work area were in excess of 100 mr/hr.
The calculated radiation exposures received ranged from 0.14 rem to 3 rem.
During the repairs, the pellet cooler area was considered an unrestricted area; of the 31 personnel, it is calcu-lated that 14 had received whole-body exposures in excess of 0.5 rems.
No health effects were observed nor would be expected from these exposures.
Cause or Causes - The gauge head was mounted external to the cooler on an upper level so that the source beam would traverse the cooler diagonally to activate the detector on the other side of the cooler, which controlled the speed of the circular cooler.
A hole had been cut in the side of the cooler to reduce shielding and allow more effective operation of the cesium-137 source in the gauge.
During recent efforts to increase production, the pressure of the air forced into the cooler had been increased as a means of accelerating the cooling of the pellets.
As a result, hot gases may have been forced out of the aperture in the cooler wall at the location of the source holder.
The temperature of the pellets entering the cooler is about 2400 F, considerably above the melting point of lead.
The heat reaching the gauge was sufficient to melt the aluminum alloy dust cover over the gauge shutter mechanism, and to melt the lead in the shutter, thereby allowing a radiation beam to escape the gauge.
The plant was shut down for periodic maintenance and overhaul of the cooler.
At the time of the shutdown, the level gauge was secured and the shutter placed in the " lock out" position.
The licensee performed a survey, but because of the relative inaccessibility of the gauge location, likely did not detect the unshielded beam.
A reading of about 10 mr/hr was reportedly obtained from this measurement.
This indicated to the licensee that the gauge was shut off with the shutter in the closed position.
(There is also some evi-dence that the survey meter was not fenctioning properly at the time and this may be another reason why the radiation beam was not discovered during the survey neither NRC regulations nor the license conditions required the licen-see to perform a survey when the source holder shutter is locked in a closed position.) The level gauge indicators in the control room could have been used to determine whether the shutter was shielding the source.
Actions Taken to Prevent Recurrence Licensee - The source holder was adequately shielded, placed in properly posted storage, and subsequently shipped to the supplier for repair or disposal.
In the meantime, a gauge manufactured by a different supplier, containing a cobalt-60 sealed source, was installed in place of the inoperable gauge.
A new lockout procedure has been established and is being followed by the licensee.
Indi-vidual workers have been notified in writing of their calculated exposures.
The removed source holder shutter assembly is to be repaired by the supplier so that the lead shielding would be contained in a sealed steel enclosure to prevent the lead from leaking out in the event of exposure to high tempera-tures.
The shutter compartment cover is to be changed to stainless steel to prevent its deterioration.
9 NRC - The incident was investigated by the NRC on May 19-21, and 26,1981 (Ref. 7).
Two items of noncompliance were identified:
10 CFR 20.101 - individuals received enposures in excess of regulatory limits; and 10 CFR 20.105 - radiation levels were created in an unrestricted area in excess of regulatory limits.
Possible enforcement action is under review; to date, the licensee has not been cited based on the section of the Interim Enforcement Policy which states, in part, taat licensees will not ordinarily be cited for violations resulting from matters not within their control that could not have been reasonably foreseen.
The NRC has since amended the license to make further events of this nature subject to a civil penalty.
The event is being considered for possible generic implications.
On December 15, 1981, the NRC issued Inspection and Enforcement Information Notice No. 81-37 (Ref. 8) to appropriate licensees to inform them of this event.
This incident is closed for purposes of this report.
81-6 Calculated Overexposure in an Unrestricted Area Preliminary information pertaining to this event was reported in the FEDERAL REGISTER (Ref. 9).
Appendix A (Example 2 of "For All Licensees") of this report notes that 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)) can be considered an abnormal occurrence.
Date and Place - Based on an investigation of a lost 1.5-curie cesium-137 source reported to the NRC on June 26, 1981 by a consultant of Mustang Services Company of Houston, Texas, it was found that a member of the general public may have received a radiation exposure in excess of NRC regulatory limits on June 23, 1981 in Norman, Oklahoma.
Nature and Probable Consequences - Mustang Services Company (the licensee) was l
closing its Oklahoma City facility and had sold a trailer containing a mounted i
gauge for determining pipe wall thickness; the gauge contained the radioactive Since the new owner of the trailer did not have a license to possess source.
the source, the licensee removed the gauge prior to the new owner taking posses-sion of the trailer in Oklahoma City.
A former employee of the licensee was contracted to remove the gauge and transport it to the licensee's facilities in Houston in a pickup truck owned by the licensee.
The contract employee began the gauge removal alone at about 11:00 PM on June 18, l
l 1981.
He, however, did not have any personnel dosimetry nor did he use any l
survey meters.
During the gauge removal, the small sealed radioactive source was inadvertently and unknowingly released from its shielded enclosure and apparently fell into a pan recessed below the floor of the trailer.
There were
10 several large wear holes, several inches in length, in the pan.
The employee transferred the gauge to the bed of the pickup truck and left the site, bound for Houston, at about midnight.
Th: traile" was locked upon his departure.
For the next five days, the locked trailer remais.ed in the locked compound, a restricted area.
No work was performed in the compound during this time; there-i fore, no known exposures are believed to have occurred during this time interval.
1 On June 23, 1981, a representative of the new owner began towing the trailer to the new owner's facilities in Houston, Texas.
A stop was made in Norman, Oklahoma for engine repairs and a second stop was made in Ardmore, Oklahoma for refueling.
During the first stop for engine repairs, the trailer was dis-connected from the truck and the representative waited near the trailer for about four hours.
On June 24, 1981, the licensee discovered that the source was not in the gauge.
The licensee had a consultant perform radiation surveys of the gauge, pickup truck, the trailer, and the facilities and grounds of both the licensee and the new owner of the trailer.
The consultant informed the NRC of the loss during the morning of June 26, 1981 and stated that the health departments of Texas and Oklahoma had been notified and were assisting the consultant in a highway search for the source.
Using radiation detection equipment, a Texas Department of Health Resources representative located the source on a bridge near Lewisville, Texas during the evening of June 26, 1981.
The source had fallen onto a structural member of the bridge several feet below the surface of the roadway.
The dose rate at the bridge surface was about 5 millirem / hour.
Thus, it is unlikely that anyone received any appreciable exposure from the source while it was lodged on the bridge support structure.
Apparently, the source fell through one of the holes i
in the trailer floor pan during transport.
Since the source was not within the licensee's immediate control for several days, there was a potential for serious radiation overexposures if someone came in close contact with it.
Based on the NRC investigation, however, it is believed that only two individuals received appreciable exposures.
One was the contract employee who dismantled the gauge and may have received a calcu-lated whole-body exposure of less than 600 millirems, which is less than the NRC regulatory limit for a worker in a restricted area in a calendar quarter.
The other was the representative of the trailer's new owner who may have received a conservatively calculated whole-body exposure of about 1.4 rem (an average of 350 millirems / hour) standing near the trailer while the truck was being repaired.
10 CFR 20.105(b)(1) requires that a licensee not use, possess, or transfer radioactive material such that there is created in an unrestricted area radiation levels which, if an individual were continuously present in the area, could result in his receiving a dose in excess of two millirems in any one hour.
The exposures possibly received by the licensee's contract employee and the truck dr',ver are not expected to result in any detectable medical effects.
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- a-
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s.
11 J
Cause or Causes - The principal causes of the incident were (1) failure of the licensee to use or supervise the use of licensed material by an individual as authorized by the license, and (2) failure of the contract employee to make proper radiation surveys during the gauge removal from the trailer.
I Actions Taken to Prevent Recurrence Licensee / Consultant - The immediate corrective actions were directed to finding the lost source, as described above.
l The licensee, having closed its Oklahoma City facility, is no longer performing
{
NRC licensed activities.
The licensee requested termination of its license, a
and the license was terminated by the NRC on September 14, 1981.
NRC - An investigation was conducted by an inspector of the NRC Region IV office l
on June 29, 1981; three violations of regulatory requirements were identified.
j A telephone enforcement conference was held with the President of the licensee on August 5, 1981.
On August 21, 1981, the NRC issued a Notice of Violation and Proposed Imposition of Civil Penalties, in the amount of $6,000 (Ref. 10).
The licensee responded by a letter dated September 18, 1981 (Ref. 11).
Based on an analysis of this response, the NRC reduced the penalties and issued an j
Order Imposing Civil Penalties in the amount of $4,000 on October 20, 1981 (Ref. 12).
The licensee paid the civil penalties in full on November 11, 1981.
{
On December 15, 1981, the NRC issued Inspection and Enforcement Information j
Notice No. 81-37 (Ref. 8) to appropriate licensees to inform them of this event.
1 The incident is closed for purposes of this report.
l l
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 l
report the events to the NRC for inclusion in this report.
During the third calendar quarter of 1981, the Agreement States reported the following abnormal occurrences to the NRC.
1 AS81-1 Excessive Radiation Doses to Hospital Patients l
Appendix A (The General Criteria) of this report notes that a major reduction in the degree of protection of the public health or safety can be considered f
an abnormal occurrence.
Date and Place - On Decemter 1, 1980, the Radiation Safety Officer of St. Joseph's Hospital of Albuquerque, New Mexico reported to the State by telephone that a I
number of patients with prostate cancer may have received excessive radiation doses over a time span of about 22 months.
i i
12 Nature and Probable Consequence - During June 1977, the hospital was licensed by the State of New Mexico to possess and implant iodine-125 seeds for the treatment of certain cancers.
In July 1977, the hospital installed a com-puterized treatment planning program which gave an output in terms of rads per hour.
A dose conversion factor was then to be used to convert rads per hour i
to total rads. When necessary, accelerator produced external radiation was also used to achieve the planned dose of radiation.
During the period from December 8, 1977 to October 15, 1979, eighteen prostate cancer and eighteen other cancer cases were treated at the hospital.
Severe adverse reactions were noted and questioned by urologists in the prostate cancer patients.
For seven of these patients, six permanent and one temporary colosto-l mies were necessary.
Examination of the non prostate cancer cases disclosed no adverse effects.
3 Cause or Causes - Investigation showed that use of an incorrect dose conver-sion factor misled two radiotherapists, who had little experience in use of iodine-125 seed implants, to conclude that an inadequate dose had been de-livered by the seeds.
The radiotherapists then provided unnecessary accelera-tor produced external radiation to some patients.
This resulted in an exces-sive dose and complications in several cases.
The investigation was not able to determine the source of the incorrect dose conversion factor.
Faulty implant techniques may have contributed to complications in two prostate cancer cases where external radiation was not used.
Inadequate health physics, treatment planning support, records and documentation were also factors contributing to the occurrence.
Actions Taken to Prevent Recurrence
'f Licensee - The hospital conducted an investigation into the causes of the inci-4
]
dents and cooperated with the State Agency in their investigations.
The two I
radiotherapists resigned from the medical staff of the hospital.
The hospital administrators appointed a fact-finding committee and retained independent con-l sultants to review the situation.
The hospital ordered a new therapy treat-ment planning system, hired a new director of radiotherapy, and upgraded their i
documentation and administrative procedures.
Iodine-125 seed implant therapy l
ceased pending approval of a license amendment requiring appropriate procedures.
State Agency (New Mexico Environmental Improvement Division, Radiation Protec-tion Bureau) - The State Agency conducted an investigation and reviewed the reports submitted by the licensee to ensure that the corrective actions were adequate.
In addition, the NRC was requested to provide assistance in an inde-pendent review of the licensee's reports.
The NRC arranged to make available a
a medical physicist consultant to the State Agency.
The consultant's findings generally agreed with those of the licensee and the State Agency concerning the incident.
The State Agency provided a copy of the report of their investi-gation to the NRC on November 13, 1981.
This incident is closed for purposes of this report.
13 i
I AS81-2 Overexposures of a Radiographer and Two Barge Crew Members Appendix A (Example 1, "For All Licensees") of this report notes that exposure of the feet, ankles, hands, or forearms of any individual to 375 rems or more of radiation can be considered an abnormal occurrence.
In addition, Example 2 of "For All Licensees" notes that an exposure to an individual in an unrestricted area such that the whole-body dose received exceeds 0.5 rem in one calendar year can be considered an abnormal occurrence.
Date and Place - The Louisiana Nuclear Energy Division (State Agency) reported that on July 14, 1981, a radiographer for a Louisiana licensee (Analytic Inspec-tion, Inc. of Lafayetts, Louisiana) received an estimated dose to the fingers of 3,000 to 5,000 rads while working on a barge at Mound Point A in the Gulf of Mexico near Intracoastal City, Louisiana.
The Captain of the barge and his helper (considered as members of the general public) received estimated whole-body doses of somewhat less than 10 rem and 2 rem, respectively, i
Nature and Probable Consequences - The radiographer received sufficient dose to produce blistering of three fingers on his left hand.
The radiographer had retrieved an 11-:urie cobalt-60 source on July 14, 1981.
Approximately nine days later he experienced pain in his left hand, and eventually, blisters devel-oped on three fingers of his left hand, including the thumb.
The Captain of the barge on which the radiographer was working was also near the unshielded 11-curie cobalt-60 source for approximately the same length of time as the radiographer and at approximately the same distance from the source.
The Captain also had a helper on the barge; however, this individual was not
],
as near the source as the Captain.
The reenactment of the incident did not provide sufficient information to estab-lish an accurate whole-body dose (the radiographer was not wearing a film badge since radiographic operations were not being performed at the time); however, taking into consideration the imprecise reenactment, it is presumed that the whole-body dose to the radiographer and Captain would not exceed 25 rem.
Cytogenetic studies were performed in an attempt to give a more precise estimate of dose to these individuals, but the results indicated no significant whole-body dose to either individual.
Therefore, the cytogenetic studies do support the conclusion that the whole-body dose to these individuals would not exceed 25 rem and was, in all likelihocd, less than 10 rem.
The dose to the helper was estimated to be less than 2 rem.
From the clinical indications, the dose to the fingers of the radiographer was estimated to be 3,000 to 5,000 rads.
He is receiving medical treatment.
(
Cause or Causes - The barge is designed such that after it is moved to a spe--
cific location, mooring stilts are put down to the ocean floor and the barge is jacked up (raised) to lift the barge out of the water; this considerably reduces the effects of the water waves on the barge during inspections, repair work, etc. On Tuesday morning, July 14, 1981, the Captain of the barge was jacking 1
--=
4 4
i J
i 14 4
i the barge down to move it from one location to another.
During this process the barge tilted, and the exposure device which had been secured with a half-
)
inch grass rope to one railing of the barge broke loose and rolled under a pump i
assembly on the opposite side of the barge.
The pump assembly was at the proper I
height for the exposure device to roll under it, allowing the lock box to hit i
the assembly. This action sheared the four bolts that attached the lock box to l
the exposure device, and it pulled the lock box and source assembly from the exposure device. After jacking the barge down, the Captain moved the barge to j
the next location and raised the barge into place.
Ttts took approximately 15 l
minutes, and the source was approximately 30 feet from the Captain during this time.
j After jacking the barge up, the Captain went to look at the exposure device.
j At about the same time the radiographer came from the crew quarters, which was j
also approximately 30 feet from the source and approached the exposure device.
The Captain stated that the exposure device had been broken and handed the lock box and source assembly to the radiographer.
The radiographer stood there for a few minutes, looking at the pigtail assembly and lock box.
During this time, i
each individual handled the pigtail assembly and lock box.
Then the radiographer 1
obtained a survey meter and attempted to perform a survey, however, the survey meter indicated a "0" mr/hr reading.
(It is assumed that this was because the i
survey meter had saturated, and the needle had returned to "0").
The radiographer assumed that the source was still in the exposure device and only the lock box and pigtail had been torn from the device, but this was an incorrect assumption.
I The radiographer then returned the pigtail assembly and lock box to the exposure device and taped the lock box in place in preparation for transportation to his office.
j 1
The helper was approximately 3 feet from the source for no more than 5 minutes, j
a very conservative estimate.
I During the handling of the lock box and pigtail assembly, there were several j
occasions when the radiographer may have actually touched the source capsule.
The unshielded source was not discovered until the barge returned to port and the broken exposure device was returned to the manufacturer for repair.
Actions Taken to Prevent Recurrence f
i Licensee - The licensee has not issued a formal reply to the State Agency's i
notice of violation; however, the licensee has informed the State Agency that all-of their radiographers have been cautioned concerning survey meter satura-i tion and have been reinstructed in emergency procedures.
State Agency (Louisiana Nuclear Energy Division) - Appropriate citations have I
been made for the excessive exposure to individuals and for the radiographer not following prescribed emergency procedures.
In addition, the State Agency
15 performed an independent study of the survey meter in question and established that the survey meter would saturate and the needle return to "0".
The licensee was made aware of this finding.
This incident is closed for purposes of this report.
17 i
REFERENCES 1.
U.S. Nuclear Regulatory Commission, " Abnormal Occurrence:
Misalignment of High Head Safety Injection Valve," FEDERAL REGISTER (Item is being published in the FEDERAL REGISTER).
2.
"Immediate Action Letter No. 81-25" from 8. H. Grier, Director, NRC Region I Office, to J. J. Carey, Vice President, Nuclear Division, Duquesne Light Company, Docket No. 50-334, June 9, 1981.*
3.
Letter from R. C. Haynes, Regional Administrator, NRC Region I, to J. J. Carey, Vice President, Nuclear Division, Duquesne Light Company,
" Investigation 50-334/81-16," Docket No. 50-334, December 10, 1981.*
4.
U.S. Nuclear Regulatory Commission, " Abnormal Occurrence:
Failure of High Pressure Safety Injection System," FEDERAL REGISTER Vol. 46, No. 239, December 14, 1981, 61020-61023.
5.
U.S. Nuclear Regulatory Commission, Inspection and Enforcement Information Notice No. 81-31, "?ailure of Safety Injection Valve to Operate Against Differential Presstre," October 7, ]981.*
6.
U.S. Nuclear Regulatory Commission, " Abnormal Occurrence:
Calculated Radiation Exposures Exceeding 10 CFR 20 Limits," FEDERAL REGISTER Vol. 16, No. 239, December 14, 1981, 61020.
7.
Inspection and Enforcement Investigation Report No. 30-12130/81-01, conducted on May 19-21 and 26, 1981, forwarded by letter from J. G.
Keppler, Director, NRC Region III, to D. O. Wilson, Plant Manager, Eveleth Expansion Company, License No. 22-11072-03, August 17, 1981.*
8.
U.S. Nuclear Regulatory Commission, Inspection and Enforcement Information Notice No. 81-37, " Unnecessary Radiation Exposures to the Public and Workers During Events Involving Thickness and Level Measuring Devices,"
December 15, 1981.*
9.
U.S. Nuclear Regulatory Commission, " Abnormal Occurrence:
Overexposure in an Unrestricted Area," FEDERAL REGISTER Vol. 46, No. 223, November 19, 1981, 56958.
10.
Letter from Victor Stello, Jr., Director, NRC Office of Inspection and Enforcement, to N. M. Howard, Vice President, Mustang Services Company, forwarding a Notice of Violation and Proposed Imposition of Civil Penalties (for $6,000), License No. 42-17552-01, August 21, 1981.*
A Available in NRC Public Document Room, 1717 H Street, NW., Washington, D.C.
20555, for inspection and copying for a fee.
l 18 11.
Letter from J. H. Coulter, President, Mustang Services Company, to Director (Victor Stello, Jr.), NRC Office of Inspection and Enforcement, License No. 42-17552-01, September 18, 1981.*
12.
Letter from Victor Stello, Jr., Director, NRC Office of Inspection and Enforcement, to Mustang Services Company, forwarding an Order Imposing Civil Penalties (for $4,000), License No. 42-17552-01, October 20, 1981.*
l
19 APPENDIX A ABNORMAL OCCURRENCE CRITERIA The following criteria for this report's abnormal occurrence determinations were set forth in an NRC policy statement published in the FEDERAL REGISTER on February 24, 1977 (Vol. 42, No. 37, pages 10950-10952).
Events involving a major reduction in the degree of protection of the public health or safety.
Such an event would involve a moderate or more severe impact on the public health or safety and could include but need not be limited to:
1.
Moderate exposure to, or release of, radioactive material licensed by or otherwise regulated by the Commission; 2.
Major degradation of essential safety-related equipment; or 3.
Major deficiencies in design, construction, use of, or management controls for licensed facilities or material.
Examples of the types of events that are evaluated in detail using these criteria are:
For All Licensees 1.
Exposure of the whole body of any individual to 25 rems or more of 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 rens or more of radiation (10 CFR Part 20.403(a)(3)), 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 Fsrt 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 on packages, or loss of confinement of radioactive material such as (a) a radiation dose rate of 1,000 mrem per hour three feet from the surface of a package containing the radioactive material, or (b) release of radioactive material from a package in amounts greater than the regulatory limit (10 CFR Part 71.36(a)).
20 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 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 Part 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.
12.
Series of events (where individual events are not of major importance),
re;urring incidents, and incidents with implications 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 fro.n 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 l
plant capability to perform essential safety functions such that a potential release of radioactivity in excess of 10 CFR Part 100 i
1
~
21 guidelines could result from a postulated transient or accident (e.g., loss of emergency core cooling system, loss of control rod system).
For Fuel Cycle Licensees 1.
A safety limit of license Technical Specifications is exceeded and a plant shutdown is required (10 CFR 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 confinement system to perform its designated function.
l l
l l
l t
23 APPENDIX B UPDATE OF PREVIOUSLY REPORTED ABNORMAL OCCURRENCES During the July through September 1981 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-The referenced Congressional abnormal occurrence reports below provide rences.
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 The following abnormal occurrence was originally reported in NUREG-0090, Vol. 2, No. 1, " Report to Congress on Abnormal Occurrences:
January-March 1979,"
and updated in subsequent reports in this series, i.e., NUREG-0090, Vol. 2, No. 2, Vol. 2, 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, and Vol. 4, No. 2.
It is further updated as follows:
79-3 Nuclear Accident at Three Mile Island Reactor Building Entries Five reactor building entries were made during the third calendar quarter of 1981.
A total of four individuals participated in the entry on July 1,1981; the reactor building polar crane was inspected and surveyed to assess the amount of damage resulting from the March 28, 1979 accident.
Tests on 14 crane motors indicated that six may not be operable.
Insulation on some of the crane power cables was damaged to the extent that bare conducting wires were visible.
The damage observed in the polar crane cab appeared to indicate the presence of flames, due to the hydrogen burn during the accident.
Crane mechanical parts appeared undamaged; however, additional mechanical repairs and refurbishments are being evaluated by the licensee.
This crane is essential for the future removal of the reactor vessel head and internals.
i All four of the individuals entering the reactor building on July 1,1981 exited with varying degrees of skin contamination.
The same type of protective clothing had been worn during the initial climb on the crane during a previous entry with no instances of skin contamination.
The July 1 climb was physically more demanding and all team members exited from the reactor building exhausted, with the inner layer of protective clothing completely soaked.
The licensee is evaluating available information to determine if a different combination of l
protective clothing is required, l
l
_ _ =
24 4
Four teams, a total of nine people, entered the Unit 2 reactor building during j
the entry on July 23, 1981.
The following tasks were successfully completed:
4 maintenance of the closed circuit TV system, removal of the Core Flood Tank "B" transducers.
photographs of the air cooler equipment and components of the reactor coolant makeup system, radiation surveys of the shallow end of the refueling pool, miscellaneous sample collection, installation of two area radiation monitors.
During the entry on August 27, 1981, five teams, a total of eleven people, entered the Unit 2 reactor building.
The following tasks were completed:
measurements using a portable gamma spectrometer, closed circuit television maintenance, air cooler inspection and survey, reactor building tool removal, cleanup and inspection.
On Thursday, September 3, 1981, a ten minute unscheduled entry was made into j
the Unit 2 reactor building by one individual to visually confirm the water level on the floor of the reactor building; this confirmation was needed s;nce the instrument that is normally used for level measurement began to give erratic indications. A visual verification, made through the open stairwell, confirmed the assumption that there had not been a noticeable change in the water level.
It was subsequently determined that air leaks into the instrument were causing the erroneous readings.
After repairs, the instrument indicated that the water surface was at the 290.96 elevation.
1 l
An entry was made on Septamber 24, 1981, by twelve individuals and was the
)
first to be made since the caset of sump water processing by the Submerged Demineralizer System.
Since the initial batch removed from the sump was small (15,000 gallons) no significant changes were found in the radiological condi-tions inside of the building.
The only detectable change was an increase of 1
i approximately 100 mr/hr on contact at a hose coupling on the sump water transfer hose.
A 100 ml sample from the reactor building sump was also taken.
The water was murky with contact radiation levels similar to water samples previously taken (1 R/hr).
Additional activities included the following:
inventory of defueling tools, photographs to support future operations, radiological surveys to characterize the reactor building, sump sample under the Reactor Coolant System Drain Tank rupture disk.
Advisory Panel On July 9,1981, the Advisory Panel for the Decontamination of Three Mile
{
Island, Unit 2 held a public aeeting in Harrisburg, Pennsylvania.
The panel m
4 i
25 l
received comments from the audience which included options on various solutions to the Unit 2 cleanup financial problems, questions on hydrogen generation, l
questions on Unit 1 reactor vessel embrittlement, emergency planning, notifi-l cation of meetings and requests for additional discussion of issues by experts not associated with the NRC, the licensee, or the licensee's parent corporation (GPU).
On Tuesday, September 1,1981, the Advisory Panel held another public meeting i
in Harrisburg.
The Chairman of the Advisory Panel, John Minnich, Daupnin l
County Commissioner, announced four new panel members recently appointed by the NRC.
They are General Dewitt Smith, Director of Pennsylvania Emergency Management Agency; Neil Wald, Professor at the University of Pittsburgh; Tom Smithgall, a local citizen; and Elizabeth Marshall, Mayor of York.
The meeting was then opened up to members of the public for discussion of proposals to resolve the financial impass that is delaying the cleanup of Unit 2.
During the meeting, Representative Allen Ertel introduced his Nuclear Power Plant Property Damage Insurance Act of 1981 which has been assigned to the House Committee on Energy and Commerce and Interior and Nuclear Affairs.
It proposes establishing a government insurance corporation with initial financ-ing by the U.S. Treasury and the payment of premiums to the insurance corpor-ation from the nuclear utilities of at least $150 million a year until a 4
reserve of $750 million has been established.
The corporation would be author-ized to provide up to $2 billion in insurance coverage for any single nuclear accident.
The insurance corporation would pay 75 percent of the uninsured costs incurred by GPU for TMI Unit 2 cleanup.
GPU would repay 50 percent of this amount over an extended period of time, and when the cleanup is completed, the government insurance corporation would be converted to a mutual insurance company with the insured utilities as the owners.
i Other proposals were also introduced by designated representatives which are j
described as follows:
Governor Thornburgh's proposal is a cost-sharing arrangement for the cleanup which provides that 25 percent of $750 million in estimated unfunded cleanup costs would be contributed by nuclear utilities, manufacturers and suppliers.
I This industry sharing would amount to an average of $31.7 million a year for six years.
Twenty-five percent, or another $31.7 million a year for six years, would be contributed by the federal government in the form of research and development grants.
Thirty-two percent of the unfunded cleanup costs or
$245 million would be contributed by GPU.
The State of Pennsylvania would contribute 4 percent or $5 million annually, and the State of New Jersey would contribute 2 percent or $2.5 million a year.
GPU would devote a further 12 percent sharing by committing the remaining $90 million in unexpended i
insurance coverage to the cleanup.
Senator Specter's Bill has been assigned to the Senate Ccmmittee on Governmental Affairs.
It is similar to the Ertel Bill except it provides for the establish-l ment of a national nuclear property insurance corporation without any federal l
financing.
l l
l l
26 Senator Heinz has announced that he intends to introduce a bill that would establish an insurance fund financed by nuclear utilities under the supervision of an existing federal agency and provide 75 percent of the funds for the uninsured portion of the Unit 2 cleanup.
Senator Heinz has also agreed to introduce any legislatiot, to Congress that may be necessary to implement the Thornburgh proposal.
After all the proposals were introduced, GPU Vice President for Planning, Bernard Cherry, stated that GPU is optimistic that the financial obstacles can be satisfactorily resolved later this year or early next year.
In addition to the above, a representative of the Pennsylvania Public Utilities Commission stated that the cleanup funding should not be the sole responsibility of the GPU ratepayers.
Submerged Demineralizer System (SDS)
Startup of the SDS began on June 30, 1981 with the processing of intermediate level waste water collected in auxiliary building tanks.
Approximately 120,000 gallons of this water was processed through the SDS.
Efficiency results on the radioactivity removal capability of the SDS indicated that greater than 99% of Cs-137 and Sr-90 were removed from the process stream.
On September 22, 1981, the licensee began transferring water from the Unit 2 Reactor Building Sump into the SDS Feed Tanks located in fuel pool "A" of the Unit 2 Fuel Handling Building.
The transfer of this water was in preparation for processing approximately 15,000 gallons as a trial batch to evaluate SDS performance before processing the remaining reactor building sump water.
Water transfer was completed on September 23, 1981.
The system which transfers the reactor building sump water through two underwater filters to the feed tanks operated as expected without problems.
Local radiation levels near system components and piping increased slightly (less than 1 mR/hr general area) as expected and returned to normal background following system flushes with demineralized water when the transfer was completed.
On September 23, 1981, the licensee began processing the reactor building sump water from the SDS Feed Tanks through the SDS zeolite ion exchangers.
Initial samples indicated a removal efficiency of greater than 99.9% for Cs-137 and approximately 99.6% for Sr-90.
Processing of the 15,000 gallon batch was completed on September 25, 1981.
Radiation levels during processing of the water increased as expected.
General area radiation levels in the fuel pool areas continue to be less than 1 mR/hr.
In isolated areas in contact with shielding some radiation levels rose to approximately 40 mR/hr, but since personnel access is not necessary in these areas the radiation levels are not considered excessive.
Radiation levels on the SDS ventilation system roughing filter increased to approximately 2 mR/hr as measured 18 inches from the filter.
The two high efficiency particulate air (HEPA) filters and charcoal filters downstream of the roughing filter did not indicate any increase in radiation levels.
There have been no increases
27 in effluent levels detected by the SDS ventilation monitor or the plant effluent monitor.
On September 26, 1981, the licensee began transferring 50,000 gallons of roactor building sump water to the SDS Feed Tanks.
This transfer was completed on September 27, 1981.
Processing of the 50,000 gallons commenced on the same day and is expected to be completed in approximately one week.
Disposition of Radioactive Waste On July 15, 1981, the Nuclear Regulatory Commission (NRC) and the U.S. Department of Energy (DOE) signed a memorandum of understanding which specifies inter-agency procedures for removing and disposing of the radioactive wastes resulting from the Three Mile Island Unit 2 accident.
Under the agreement, the NRC will be responsible for regulating all licensee activities, including waste management, at the Three Mile Island site to assure that the activities comply with applicable rules and regulations and the licensee's operating license as modified.
Also, the NRC staff will continue to keep public state and local officials informed of its activities and involve DOE officials in information exchanges.
Close cooperation between the two agencies is expected to ensure the following disposition of TMI-2 accident generated solid radioactive wastes which currently enist or are planned to be generated:
EPICOR-II system wastes - forty-nine liners are onsite with loadings up to 1500 curies.
These liners, in their present form, are not considered suitable for commercial disposal.
DOE has agreed to take these 49 liners for research and development purposes.
Submerged Demineralizer System wastes - DOE will take possession and retain, for research and development purposes, zeolite liners generated during the operation of the system.
Reactor Fuel - DOE will analyze appropriate fuel assemblies ar.d samples, with the remainder being stored in containers in the TMI-2 spent fuel storage pool.
Disposition of the balance of the damaged fuel will await resolution of the spent fuel storage issue.
Transuranic contaminated waste materials - waste materials accumulated at TMI with transuranic levels above the levels authorized for acceptance at commercial land burial facilities will be considen.d on a case-by-case basis.
Alternatives include archiving, research and development, temporary onsite storage or DOE processing / disposal with reimbursement by the licensee.
Makeup and purification system resins and filters - due to the high levels of contamination deposited on the filters during the accident and the generic value of these filters, DOE will take possession and retain these filters for reseach and development activities.
DOE's
i 28 I
activities regarding the purification system resins will either take the form of an R&D program of generic value, or DOE will take possession of these resins for storage or disposal on a reimbursable basis.
Other solid radioactive wastes - the low-level wastes associated with decontamination (e.g., some ion exchange media, booties, gloves, trash) will be disposed of by the licensee in licensed commercial low-level burial facilities.
Further reports will be made as appropriate.
A A
A A
A A
A The following abnormal occurrence was originally reported in NUREG-0090, Vol. 3, No. 1, " Report to Congress on Abnormal Occurrences:
January-March 1980," and updated in a subsequent report in this series, i.e., NUREG-0090, Vol. 3, No. 2.
It is further updated as follows:
80-1 Occupational Overexposures to Skin and Extremities The NRC Office of Inspection and Enforcement has taken enforcement action on this matter and is currently inspecting the licensee's implementation of corrective actions and actions to prevent recurrence.
During the special health appraisal inspections of operating reactor licensees, NRC inspectors reviewed survey equipment inventories to make certain proper types of equipment were available and that instructions and calibration curves were available and were being used.
These special health physics appraisals were completed in early 1981.
The subject of monitoring for beta skin dose was also addressed in Draft NUREG-0761, " Radiation Protection Plans for Nuclear Power Reactor Licensees," that was published for comment in March 1981 (Ref. B-1).
The notice of availability of that report was published in the FEDERAL REGISTER on April 9, 1981 (Ref. B-2).
This incident is closed for purposes of this report.
I
29 APPENDIX C OTHER EVENTS OF INTEREST l
The following events art described belos because they may possibly be perceived by the public to be of public health significance.
The events did not involve a major reduction in the level of protection provided for public health or safety; therefore, tney are not reportable as abnormal occurrences.
1.
Waste Gas Decay Tank Failure On July 17, 1981, while the plant was in cold shutdown for repairs to a diesel generator, an hydrogen ignition occurred in one waste gas decay tank at San Onofre Unit 1 resulting in tank damage and a small release of noble gases.
The San Onofre Unit 1 plant, which is operated by Southern California Edison Company (the licensee), utilizes a pressurized water reactor and is located in San Diego County, California.
The event occurred while a venting process was underway on a tank located in the Reactor Auxiliary Building (Tank C-6A).
Just prior to this, a similar tank had been processed without incident.
As processing began on tank C-6A, difficulty was experienced in adjusting the flow rate through the system.
Mhile attempting to alleviate the problem, an operator heard " popping" noises in the piping.
Shortly thereafter, the gas mixture in the tank ignited.
Fire alarms were observed in the control room and operating personnel within the gas treatment system area were notified.
These personnel reported a loud noise, apparently from the gas decay tank area.
Investigation revealed the presence of smoke in the building and zero local pressure indication on the tank.
Later, the area was found clear of smoke and no fire was present.
The tank manway cover bolts were found loose and minor damage around the tank manway was noted.
The cause of the hydrogen ignition was air contamination of the inert nitrogen system which is used to control the hydrogen-oxygen concentrations in the tank.
The source of air was identified as instrument air leaking through the check valves at the cross connections between instrument air and nitrogen lines.
Under normal operating conditions, the pressure in the instrument air system is higher than that of the nitrogen system.
These cross connections had been installed in response to TMI Action Plan requirement item II.E.1.2 of NUREG-0737 (Ref. C-1).
The nitrogen system provided a backup gas supply to the air-operated steam supply valve for the steam-driven auxiliary feedwater pump.
Other cross connections, which apparently did not leak air into the nitrogen system, had been previously installed in response to another TMI Action Plan requirement, II.G.1 of NUREG-0737 (Ref. C-2), to provide a redundant gas supply to the air operated pressurizer relief valves and the associated block valves.
30 Following the occurrence, the licensee sampled all potentially affected tanks and determined that most of the tanks had oxygen levels above 10 to 15 percent.
4 Generally, the gas in pressurized-water reactor (PWR) waste gas systems is hydrogen rich and the oxygen concentration is controlled to prevent flammable gas mixtures.
Flammable concentration of gas mixtures can be prevented by limiting either the hydrogen or the oxygen concentration to less than 3 percent.
To eliminate the possibility of recurrence, the licensee has now completely separated those portions of the nitrogen system that are a backup supply to the air system from the balance of the nitrogen system that supplies cover l
gas.
Bottles of compressed nitrogen are now used to provide the backup to the air system.
The affected decay tank was repaired as necessary.
l The NRC monitored the corrective action taken by the licensee and issued Inspection and Enforcement Information Notice No. 81-27 on September 3, 1981 to power reactor licensees to inform them of this event (Ref. C-3).
The amount of noble gases released as a result of this event was about 8.8 curies.
The calculated plant boundary activity concentration was determined to be a factor of ten below the maximum permissible concentration given in 10 CFR 20.
In addition, onsite personnel exposure associated with this event was well within limits.
License conditions limit the amount of radioactivity that can normally be stored in the<.e tanks.
Therefore, the event is not considered to be an abnormal occurrence.
A contractor (Idaho Nuclear Engineering Laboratory) for the NRC is studying the problem of explosions in waste gas systems at PWRs in order to evaluate and recommend appropriate control measures.
2.
Excessive Surface Contamination on a Spent Fuel Shipping Cask On July 22, 1981, the NP,C issued an Order to Show Cause (Immediately Effective) prohibiting use of NFS-4 spent fuel shipping cask Serial No. NAC-1D outside of licensed facilities, because on a number of occasions following offsite trans-portation, the cask had displayed impermissibly high levels of surface contamination (Ref. C-4).
This matter was the subject of Congressional and local media interest at the time.
These casks are submerged in fuel pools during the process of loading 'or
]
unloading the spent fuel being transported. As a result of this submersion in the pools, the cask surfaces become contaminated.
Upon removal from the pool, i
j the casks are required to undergo decontamination before transport.
- However, a small amount of contamination frequently becomes " fixed" in the surface finish of the cask.
00T regulations establish permissible limits for such surface contamination during transport.
On at least seven occasions between August 1980 and July 1981, after transport following decontamination, Cask Serial No. NAC-10 arrived with surface contami-nation that substantially exceeded permissible limits. The increase in surface contaminaton following transport suggested that contamination which originally
_==
31 was fixed had been released in transit.
The reason for this excessive contami-nation, which may be related to the surface finish of the cask, was not fully understood.
The Order to Show Cause issued on July 22, 1981 (Ref. C-4) suspended use of the cask until it could be shown that there was reasonable assurance that surface contamination levels would not exceed the limits'of DOT regulations during future shipments.
Testing of the cask was conducted by Nuclear Assurance Corporation (NAC), the own0r of the cask.
On the basis of its evaluation, NAC defined the potential cause of the excessive surface contamination as inadequate decontamination procedures by the shippers, which had allowed a buildup of surface contamination to accumulate over a period of time.
NAC has shown that by following a pre-scribed decontamination procedure, a reduction in surface contamination to acceptable levels can be achieved.
As a result of the testing, inspection, and d: contamination program carried out to date, the NRC determined that there was reasonable assurance that the cask can be safely permitted limited usage, i.e., shipment over relatively short distances and for a limited number of shipments, to demonstrate that preparation of the cask for shipment will not result in excessive levels of contamination in transport.
Consequently, on October 9,1981, NRC issued a partial rescission of the Order to Show Cause, permitting limited use of the cask for further evaluation (Ref. C-5).
The contamination which occurred during the shipments in question was limited to the cask surface and the truck bed.
The cask is completely enclosed by a rigid metal cage which rendered the cask and truck bed inaccessible during transport.
No personnel were contaminated.
Because the contamination was limited to an inaccessible area, the situation did not present a hazard to the public health and safety, and is therefore not reportable as an abnormal occurrence.
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33 REFERENCES (FOR APPENDICES)
B-1 U.S. Nuclear Regulatory Commission, " Radiation Protection Plans for Nuclear Power Reactor Licensees," USNRC Draft Report NUREG-0761, March 1981.**
B-2 U.S. Nuclear Regulatory Commission, " Availability of Draft NUREG-0761, Radiation Protection Plans for Nuclear Power Reactor Licensees," FEDERAL REGISTER, Vol. 46, No. 68, April 9, 1981, 21285.
C-1 U.S. Nuclear Regulatory Commission, " Clarification of TMI Action Plan Requirements," Item II.E.1.2 (Auxiliary Feedwater System Automatic Initiation and Flow Indication), USNRC Report NUREG-0737, November 1980.t C-2 U.S. Nuclear Regulatory Commission, " Clarification of TMI Action Plan Requirements," Item II.G.1 (Emergency Power for Pressurizer Equipment),
USNRC Report NUREG-0737, November 1980.t C-3 U.S. Nuclear Regulatory Commission, Inspection and Enforcement Information Notice No. 81-27, " Flammable Gas Mixtures in the Waste Gas Decay Tanks in PWR Plants," September 3, 1981.*
C-4 Letter from John G. Davis, Director, NRC Office of Nuclear Material Safety and Safeguards, to various licensees, forwarding an Order to Show Cause (Immediately Effective), reference Docket No. 71-6698, July 22, 1981.*
C-5 Letter from John G. Davis, Director, NRC Office of Nuclear Material Safety and Safeguards, to various licensees, forwarding a Partial Rescission of Order, reference Docket Nc. 71-6698, October 9, 1981.*
A Available in NRC Public Document Room, 1717 H Street, NW., Washington, DC 20555, for inspection and copying for a fee.
QQSingle copies available free upon written request to.the NRC/GP0 Sales Program, Division of Tachnical Information and Document Control, U.S. Nuclear Regulator Commission, Washington, DC 20555.
iAvailable for purchase from NRC/GP0 Sales Program, Division of Technical Information and Document Control, U.S. Nuclear Regulatory Commission, Washington, DC 20555, and National Technical Information Service, Springfield, VA 22161.
$R roRM 335 U S. NUCLEAR REGUL ATORY COMMISSION BIBLIOGRAPHIC DATA SHEET NUREG-0090, Vol. 4, No. 3 4 TITLE AND SUBTITLE (Aad Volume No. d apprannate) 2 (Leave t>lanki Report to Congress on Abnormal Occurrences July - September 1981 3 HECIPIENT'S ACCESSION NO 1 AUTHORis)
- 5. DATE REPORT COMPLE TED l YE AR M ON TH Januarv 1982
- 9. PE RF ORMING ORGANIZATION N AME AND M AILING ADDRESS (include les Code)
DATE REPORT ISSUED U.S. Nuclear Regulatory Commission MONTs l YEAR Office for Analysis and Evaluation of Operational Data January 1982 Hashington, D.C.
20555 6 (tea,e manii 8 (Leave Naohl
- 12. SPONSORING ORGANIZATION N AME AND M Alt ING ADD RE SS (include lep Codel 10 PROJE CT T ASKiWORK UNIT NO.
U.S. Nuclear Regulatory Commission Office for Analysis and Evaluation of Operational Data M CONTRACT NO Washington, D.C.
2G555 13 T YPE CF REPORI PE RiGO COV E RE D //oclus ve dates /
Quarterly July - September 1981, 15 SUPPL E ME N T ARY NOTE S 14 (L eave n/ao k /
- 16. ABS TR ACT (20d words or less/
Section 208 ef 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 July 1 to September 30, 1981.
During the report period, there were two abnormal occurrences at the nuclear power plants licensed to operate. One involved a misalignment of a high head safety injection isolation valve. The other involved a failure of the high pressure safety injection system. There were two abnormal occurrences at other licensee facilities. Both involvec calculated radiation exposures in excess of 10 CFR 20 limits. There were two abnorma' occurrences reported by the Agreement States. One involved excessive radiation doses to hospital patients. The second involved overexposures of a radiographer and two barge crew members.
The report also contains information updating some previously reported abnormal occurrences.
I 7 A E Y WORDS AND DOCUME N T AN ALY S'S I 7a DE SC hip T O HS 1 In IDE N TIF IF HS OPE N EN DE D TE RYis 18 AV AIL ABILITY ST ATEME NT 19 SE CUMITY CL ASS ITF i &pmf /
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- O OF P AGE S Unclassified Unlimited 20 sECuRgTYgtgsSgog; n Rice NRC F ORM 335 67 771
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