ML20207J398
| ML20207J398 | |
| Person / Time | |
|---|---|
| Issue date: | 07/31/1988 |
| From: | NRC OFFICE FOR ANALYSIS & EVALUATION OF OPERATIONAL DATA (AEOD) |
| To: | |
| References | |
| NUREG-0090, NUREG-0090-V11-N01, NUREG-90, NUREG-90-V11-N1, NUDOCS 8809060206 | |
| Download: ML20207J398 (56) | |
Text
- . .- - -
4 t
i NUREG-0090 Vol.11, No.1 l Report to Congress on ;
, Abnormal Occurrences '
l January - March 1988 I
a i
l U.S. Nuclear Regulatory Commission j Office for Analysis and Evaluation of Operational Data l
\ u j [, . .. e ,%,
' . . . . . o#
d !
BOO 9060206 000731
o , e O
Available from Superintendent of Documents U.S. Government Printing Office Post Office Box 37082 Washington, D.C. 20013-7082 A year's subscription consists of 4 issues for this publication.
Single copies of this publication are available from National Technical Information Service, Springfield, VA 22161
- Vol.11, No.1 '
Report to Congress on Abnormal Occurrences i
January - March 1988 Data Published: July 1988 Office for Analysis and Evaluation of Operational Data U.S. Nuclear Regulatory Comm!ssion Wcshington, DC 20565
,s.. ..m, i
Previous R: ports in Series MUREG 75/090. January June 1975 NUREG-0090. Vol.4. No.4. October. December 1981, published October 1975 published May 1982 NUREG 0090-1. July-September 1975. MUREG.0090. Vol.5. No.l. .tanuary-March 1982, published March 1976 published August 1982 NUREG 0090 2. October-Deceeber 1975. NUREG.0090. Vol.5. No.2. April June 1987, published March 1976 published December 1987 NUREG 0090 3 January-March 1976. NUREG-0090. Vol.5. No.3 July.5eptember 1982, published July 1976 published January 1983 NUREG 0090 4. April June 1976. NUREG.0090. Vol.5. No.4. October. December 19P2 published March 1977 published May 1983 NUREG 00?0 5. July September 1976. NUREG.0090. Vol.6. No.l. January March 19A3,-
published March 1977 published September 1983 NUREG.0090 6 October-December 1976 NUREG 0090. Vol.6. No.2. April-June 1983, published June 1977 published November 1983 NUREG-0090 7. January March 1977 NUREG 0090. Vol.6. No.3 July-September 1983, published June 1977 published April 1984 NUREG 0090 8. April. June 1977 NUREG 0090. Vol.6. No.4. October-December 198.1, published September '977 published May 1984 NUREG 0090 9. July-September 1977 NUREG 0090. Vol.7. No.l. January-March 1984, published November 1977 published July 1984 ,
NUREG-0090 10 October-December 1977 NUREG 0090. Vol.7. No.2. April June 1984 published Parch 1978 published October 1984
! NUREG.0090. Vol.1 No.l. Januarv. March 1978 NUREG 0090. Vol.7. No.3 July September 1984, published June 1978 published April 1985 NUREG 0090. Vol.1. No.2. April-June 1978 NUREG 0090. Vol.7. No.4. October December 1984
- published September 1978 published May 1985 NURER 0090. Vol.1. No.3. July.5eptember 1978 NUREG 0090. Vol.3. No.l. January March 19P5 published Decenter 1978 published Au9ust 19A5 NURER-0090. Vol.1 No.4. October December 1978. NUREG.0090. Vol.B. No.2. April-June 1985, pubitsbed Merch 1979 published November 1985 NUREG-0090. Vol.2. No.l. January March 1979 NUREG 0D90. Vol.8. No.3. July.5eptember 1985, published July 1979 published February 1986 I NUREG-0090. Vol.2. No.2, April-June 1979. NUREG 0090. Vol.8. No.4. October December 1985 ;
published November 1979 published May 1986
]
NUREG 0090. Vol.2, No.3. July 5eptember 1979 NUREG.0090. Vol.9. No.l. January March 1986, published February 1980 published September 1986 NUREG-0090. Vol.2. No.4. October-December 1979. NUREG 0090. Vol.9. No.2. April June 1986 published April 1980 published January 1987 NUREG-0090. Vol.3. No.l. January-March 1980. NUREG-0090. Vol.9. No.3. July.5eptember 1986, published September 1980 published April 1987 NUREG 0090. Vol.3, 4o.2. April-June 1980 NUREG.0090. Vol.9. No.4. October. December 19P6 publist.ed Nove*ber 1960 published July 1987 ,
t NUREG 0090. Vol.3. No.3. July.5eptember 1980 NUREA-0090. Vol.10. No.l. January March 19A7 published February ?981 published October 1987 NUREG-0090. Vol.3. ko,4. October. December 1980. MURE4 0090. Vol.10. No.2. April. June 1987 publis%d Fay 1981 published November 19A7 NUREG-3090. 101.4 ho l. January. March 1981. NUREG 0090. Vol.'0. No.3. July Seotember 1987, puO hhed July 1981 published Parck !988 NURES 0090. V?l.4. No.2. April June 1961, NUREG 0090. Vol.10. No.4. October. December 1987, pub 1'shec OctWe MA1 published M W h 1988 NURIC4tYM Vol publ4hejdanuar.11982he.3.
y July.5eptember 1981.
ABSTRACT Section 208 of the Energy Reorganization Act of 1974 identifies an abnormal occurrence as an unscheduled incident or event which the Nuclear Regulatory Commission determines to be significant from the standpoint of public health or safety and requires a quarterly report of such events to be made to Congress.
This report covers the period from January 1 to March 31, 1988.
For this reporting period, there were three abnormal occurrences at nuclear power plants licensed to operate: a potential for common mode failure of safety-related components due to a degraded instrument air system at Fort Calhoun; common mode failures of main steam isolation valves at Perry Unit 1; and a cracked pipe weld in a safety injection system at Farley Unit 2. There were six abnormal occurrences at other NRC licensees: a diagnostic medical Disadministration; a breakdown in management controls at the Georgia Institute of Technology reactor facility; release of polonium-210 from static elimina-tion devices manufactured by the 3M Company; two therapeutic medical misad-ainistrations; and a significant widespread breakdown in the radiation safety program at Case Western Reserve University research laboratories. There was one abnormal occurrence reported by an Agreement State (Taxas) involving radia-tion injury to two radiographers.
The report also contains information updating some previously reported abnormal occurrences, i
l l
2 I
I l
l I'
iii 1
l
r i
CONTENTS i
r i
ABSTRACT.............................................................. iii vii PREFACE...............................................................
4 INTRODUCTION..................................................... vii THE REGULATORY SYS1EM............................................ vii 1
REPORTABLE OCCURRENCES........................................... vii AGREEMENT STATES................................................. viii FOREIGN INFORMATION.............................................. ix REPORT TO CONGRESS ON ABNORMAL OCCURRENCES, JANUARY-hARCH 1988........ 1 NUCLEAR POWER PLANTS............................................. 1 ;
88-1 Potential for Common Mode Failure of Safety-Related -
Components Due to a Degraded Instrument Air System at Fort Calhoun........................................ 1
) 88-2 Common Mode Failures of Main Steam Isolation Valves I at Perry Unit 1........................................ 4 88-3 Cracked Pipe Weld in Safety Injection System ,
at Farley Unit 2....................................... 6 FUEL CYCLE FACILITIES (Other than Nuclear Power Plants).......... 10 i
- OTHER NRC LICENSEES (Industrial Radiographers, Medical Institutions, Industrial Users, etc.)............................ 10 88-4 Diagnostic Medical Misadministration................... 10 l 4 i
! 88-5 Breakdown in Management Controls at Georgia Institute I l
of Technology Research Reactor Facility................ 11 l 88-6 Release of Polonium-210 from Static Elimination Devices
- Manufactured by 3M Company............................. 13 88 7 Therapeutic Medical Misadministration.................. 16 I l
68-8 Therapeutic Medical Misadministration.................. 17 1 88-9 Significant Widespread Breakdown in Radiation Safety i i Program at Case Western Reserve University Research j Laboratories........................................... 18 j
i l
l i v 4
l 1
i CONTENTS (Continued)
P,a21 AGREEMENT 3 TATE LICENSEES........................................ 21 ASB8-1 Radiation Injury to Two Radiographers.................. 21 REFERENCES............................................................ 23 APPENDIX A - ABNORMAL OCCURRENCE CRITERIA............................. 27 A. . ENDIX B - UPDATE OF PREVIOUSLY REPORTED ABNORMAL OCCURRENCES. . . . . . . 29-L NUCLEAR POWER PLANTS............................................. 29 79-3 Nuclear Accident at Three Mile Island.................. 29 81-7 Blockage of Coolant Flow to Safety-Related Systems and Components......................................... 31 85-14 Management Deficiences at Tennessee Valley-Authority.............................................. 33 FUEL CYCLE FACILITIES ........................................... 35 1
86-3 Rupture of a Uranium Hexafluoride Cylinder and Release of Gases................................... 35 APPENDIX C - OTHER EVENTS OF INTEREST................................. 37 ,
REFERENCES (FOR APPENDICES)........................................... 43 i
I l l '
I l
l 1
vi l
PREFACE l
INTRODilCTION The Nuclear Regulatory Commission reports to the Congress each quarter under provisions of Section 208 of the Energy Reorganization Act of 1974 on any abnormal occurrences involving facilities and activities regulated by the NRC. An abnormal occurrence is defined in Section 208 as an unscheduled incident or event which the Commission determines is significant from the ,
standpoint of public health or safety. ;
Events are currently identified as abnormal occurrences for this report by the NRC using the criteria listed in Appendix A. These criteria were promulgated ,
in an NRC policy statement which was published in the Federal Reaister on Feb- '
ruary 24, 1977 (Vol. 42, No. 37, pages 10950-10952). In order to provide wide ,
dissemination of information to the public, a Federal Recister notice is issued !
on each abnormal occurrence. Copies of the notice are cistributed to the NRC Public Document Room and all Local Public Document Rooms. At a minimum, each notice must contain the date and, place of the occurrence and describe its ,
nature and probable consequencol n l
The NRC has determined that only those events, including those submitted by the Agreement States, described in this report meet the criteria for abnormal occur- ,
- rence reporting. This report covers the period from January 1 to March 31, 1988.
Information reported on each event includes date and place, nature and prob- .
able consequences, cause or causes, and actions taken to prevent recurrence, i THE REGULATORY SYSTEM i
The system of licensing and regulation by which NRC carries out its responsibil-ities is implemented through rules and regulations in Title 10 of the Code of Federal Regulations. This includes public participation as an element. To ac- i complish its objectives, NRC regularly conducts licensing proceedings, inspec- ,
tion and enforcement activities, evaluation of operating experience, and con-firmatory research, while maintaining programs for establishing standards and issuing technical reviews and studies. '
! In licensing and regulating nuclear power plants, the NRC follows the philosophy that the health and safety of the public are best assured through the establish-ment of multiple levels of protection. These multiple levels can be achieved and maintained through regulations specifying requirements that will assure the safe use of nuclear materials. The regulations include design and quality I assurance criteria appropriate for the various activities licensed by NRC. An inspection and enforcement program helps assure compliance with the regulations.
REPORTABLE OCCURRENCES Actual operating experience is an essential input to the regulatory process for assuring that licensed activities are conducted safely. Licensees are re-
- quired to report certain incidents or events to the NRC. This reporting helps i
- to identify deficiencies early and to assure that corrective actions are taken j to prevent recurrence.
I ,
vii ;
1
t For nuclear power plants, dedicated groups have been formed both by the NRC- i and by the nuclear power industry for the detailed review of operating experi- ;
ence to help identify safety concerns early, to improve dissemination of such 4 information, and to feed back the experience into licensing, regulations, and
, operations. In addition, the NRC and the nuclear power industry have ongoing efforts to improve the operational data systems which include not only the type, l and quality, of reports required to be submitted, but also the methods used to ,
analyze the data. In order to more effectively collect, collate, store, re-
. trieve, and evaluate operational data, the information is maintained in computer-based data files. (
j Two primary sources of operational data are Licensee Event Reports (LERs) and I 4
immediate notifications made pursuant to 10 CFR l-50.72.
] Except for records exempt from public disclosure by statute and/or regulation,
- information concerning reportable occurrences at facilities licensed or other-wise regulated by the NRC is routinely disseminated by the NRC to the nuclear industry, the public, and other interested groups as these events occur.
l Dissemination includes special notifications to licensees and other affected or interested groups, and public announcements. In addition, information on ;
reportable events is routinely sent to the NRC's more thaa 100 local public document rooms throughout the United States and to the NRC Public Document- ;
Room in Washington, D.C. The Congress is routinely kept informed of reportable events occurring in licensed facilities. ;
- Another primary source of operational data is reports of reliability data sub- ;
sitted by licensees under the Nuclear Plant Reliability Data System (NPRDS). !
! The NPRDS is a voluntary, industry-supported system operated by the Institute j
- . of Nuclear Power Operations (INPO), a nuclear utility organization. Both engi- -;
a neering and failure data are submitted by nuclear power plant licensees for a specified plant components and systems. The Commission considers the NPRDS '
to be a vital adjunct to the LER system for the collection, review, and feed- .
back of operational experience; therefore, the Commission periodically monitors i the NPRDS reporting activities. ,
1 :
! AGREEMENT STATES !
i l Section 274 of the Atomic Energy Act, as amended, authorizes the Commission to !
3 enter into agreements with States whereby the Commission relinquishes and the ,
- States assume regulatory authority over byproduct, source and special nuclear ;
3 caterials (in quantities not capable of sustaining a chain reaction). Agree- l ment State programs must be comparable to and compatible with the Commission's
( program for such material, i 1
Presently, information on reportable occurrences in Agreement State licensed t activities is publicly available at the State level. Certain information is !
also provided to the NRC under exchange of information provisions in the ;
j agreements. j 1
l In early 1977, the Commission determined that abnormal occurrences happening l 3 at facilities of Agreement State licensees should be included in the quarterly reports to Congress. The abnormal occurrence criteria included in Appendix A a
! viii l
t
. . _ - . - _ ~ _, ,._ _ _ - - _ _ . _ _ _ _ - - . .__,_,i
l l
are applied uniformly to events at NRC and Agreement State licensee facilities, r Procedures have been developed and implemented and abnormal occurrences reported by the Agreement States to the NRC are included in these quarterly reports to Congress.
FOREIGN INFORMATION The NRC participates in an exchange of information with various foreign govern-ments which have nuclear facilities. This foreign information is reviewed and considered in the NRC's assessment of operating experience and in its research and regulatory activities. Reference to foreign information may occasionally be made in these quarterly abnormal occurrence reports to Congress; however, only domestic abnormal occurrences are reported.
l i
j i
i i ;
l 1 l 1
1
! l 1
ix l l
REPORT TO CONGRESS ON ABNORMAL OCCURRENCES JANUARY-MARCH 1988 NUCLEAR POWER PLANTS !
The NRC is reviewing events reported at the nuclear power plants licensed to operate. For this report, the NRC has determined that the following events
! were abnormal occurrences:
- 88-1 Potential for Common Mode Failure of Safety-Related Components Due to a Degraded Instrument Air System at Fort Calhoun The following information pertaining to this event is also being reported concurrently in the Federal Register. Appendix A (see the third general criterion) of this report notes that major deficiencies in design, construction, i use of, or management controls for licensed facilities or material can be con-
- sidered an abnormal occurrence. In addition, Example 10 of "For All Licensees" of Appendix A notes that a major deficiency in design, construction, or opera-tion having safety implications requiring immediate remedial action can be considered an abnormal occurrence.
Date and Place - September 23, 1987; Fort Calhoun, a Combustion Engineering-designed pressurized water reactor, operated by Omaha Public Power District, and located in Washington County, Nebraska, i
Nature and Probable Consequences - While the licensee was performing a surveil-
~
lance test of emergency diesel generator (EDG) 2, the EDG tripped off due to i high temperatures in the engine cooling water system. Investigation revealed an instrument air problem (water in the system) which could have resulted in a ;
potential for common mode failure of redundant EDG 1 and other safety-related l components at the plant.
1 Many U.S. light water reactors rely on air systems to activate or control many safety-related, as well as many non-safety related, components. The instrument '
air system activates pneumatic controls, valves, dampers, and similar devices on these components. Through the years, the there have been numerous problems caused by either failures in the air system components, or failures in air-controlled components which have been degraded by contaminated air. Many prob-lems have been caused by air system design, operation, or maintenance deficien-i cies. The circumstances associated with the Fort Calhoun event are as follows: l During May 1985, the licensee modified its fire protection sprinkler system in the diesel generator rooms, where extremely cold weather had caused water to freeze and crack the pipes. Rather than keep this piping filled with water all ,
the way to the sprinkler head, the licensee devised a "dry pipe" arrangement 1
by which pressurized air would fill the pipe for several feet upstream from ,
the sprinkler head, keeping water in check that would be released in the event I i
of a fire. To accomplish this, the instrument air system was connected to the {
sprinkler system, and check valves were installed to keep water from flowing i 5
back into the air pipes. However, the licensee failed to establish a test i 1
l l
1 ,
I program that would assure that the check valves would perform satisfactorily ,
in service. !
On July 6, 1987, these check valves failed to work, and water entered the air system after the licensee had tested its "dry pipe" sprinkler. After this ;
incident, the licensee cleaned and inspected the check valves to see that they i operated, then reconnected the two systems. However, the licensee failed to j assure that the cause was properly identified and that appropriate corrective actions had been taken. For example, no dew point measurements were taken to verify that the air system complied with the design bases for the dewpoint maximum limit; and no review was performed to determine whether or not other instrument air / pressurized water interfaces existed. In addition, even though a formal program was established to perform blowdowns of the air system, the air accumulators for the EDG air motors were not included in either the piping drawings or procedures; therefore the accumulators were not checked for the presence of water.
On August 25, 1987, the licensee again found water in the instrument air system, discovering this time that the situation resulted from an interconnection with an air-actuated valve in the plant's potable water system.
On September 23, 1987, after EDG 2 tripped on high temperature during a surveil-lance test, the licensee determined that the most likely cause of the event was the failure of the exhaust damper on the diesel radiator to open. Without the damper open, no radiator cooling air flow was available and the engine cooling water temperatures increased.
The licensee disassembled the air motor used for opening and closing the radiator exhaust damper and found an accumulation of water in the motor air accumulator, apparently from the July 6, 1987 water intrusion event. The licensee found that the pilot valve used to direct air into the motor was coated with a gummy substance. The radiator exhaust damper air motor uses the plant instrument air system as a prime mover. The licensee cleaned up and reassembled the air motor on EDG 2. Subsequent testing on the EDG 2 damper indicated the damper operated satisfactorily.
Subsequently, the air motor on EDG 1 was disassembled and the same problem was
- found as in EDG 2. It was also cleaned, reassembled, and tested satisfactorily.
Cause or causes - The root cause of this incidemt was due to a breakdown in the ability of management to control activities that affect quality at the Fort Calhoun Station. A plant system had been modified without adequate evaluation of the safety implications and improper testing following the modifications.
This breakdown resulted in the plant being operated in an unanalyzed condition where a potential for common mode failure existed.
Actions Taken to Prevent Recurrence Licensee - An extensive corrective action program has been undertaken to ensu.e that all water is removed from the air system and to ensure that all safety- I related components function normally. The interface between the instrument air I
l 2
1
!l 7 i
system and the diesel generator fire protective system has been removed. A !
walk down of the entire plant instrument air system was made to assure that !
any other IntFconnections to the fire protection system were either isolated ;
or removed.
i Mc N freqtent inservice testing inspections (ISI) will be performed on various ;
component:, operated by the instrument air system until there is assurance that they have not been adversely affected. For components that cannot be tested j during plant operation, justification is to be provided to continue operation until the next scheduled or forced cold shutdown in excers of 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.
NRC - An initisi inspection was performed b W Region IV during the period of' :
i Tejitember 23 through October 2,1987. The inspecdon findings were formally i
sent to the licensee on October 23, 1987 (Ref. 1). On October 29, 1987, an.
enforcement conference was held with the licensee at the NRC Region IV office to discuss the issues related to the event.
l
- - A special review team from NRC Region IV performed a follow-up inspection during j the period of November 2-6, 1987. The inspection findings were formally sent to the licensee on December 10, 1987 (Ref. 2).
I During the inspections, several violations of NRC requirements were found.
Consequently, on February 22, 1988, the NRC issued to the licensee a Notice of Violation and Proposed Imposition of Civil Penalty in the amount of $175,000
, (Ref. 3).
The licensee was cited for modifying a plant system without adequately evaluat-ing the safety implications, including the possibility that water in air lines could simultaneously disable several safety systems. The licensee was further cited for failing to: (1) test the check valves designed to prevent water from i backflow'-~ into the air lines; (2) provide app /opriate instructions or proce-dures tr m asonnel involved with the modified systems; (3) recognize tho
, inoper d E ty of the disabled diesel generator or test the second diesel for
! operabit!t/; (4) declare an "unusual event" when the diesel penerator was l inoperabl u (5) notify the NRC promptly of these conditions; and (6) correct the
- degraded safety conditions promptly and adequately. !
The licensee has paid the civil penalty. The NRC will assure that the correc- i tive actions proposed are complete and satisfactorily implemented. t Instrument air system problems are being addressed generically by NRC in several '
ways. A staff task group is continuing to examine the need for further improve- )
i ments in instrument air systems under Generic Issue 43, "Reliability of Air Sys-J temr. . " Operating experience is being updu.ed and ted back to the industry, such :
! as the updated case study NUREG-1275, Vol. 2, "Operating Experience Feedback i Report - Air System Problems," published in December, 1987 (Ref 4) and NRC an- 1 formation Notice No. 87-28, Supplement 1, "Air System Problems at U.S. i.ight i Water Reattors" (Ref. 5). NRC activities focused on improving air systems ara !
continuing. 1 l
1 This item is considered closed for the purposes of this report,
- a******
f 3 ,
9
--- e, + - - - - , n . - . - ----..- , - ~ , ~.- - -.w-- ev - ,
88-2 Common Mode Failures of Main Steam Isolation Valves at Perry Unit 1 The following information pertaining to this event is also being reported concurrently in the Federal Register. Appendix A (see the second general criterion) of this report notes that major degradation of essential safety-related equipment can be considered an abnormal occurrence.
In addition, Example 2 of "For Commercial Nuclear Power Plants" notes that major degradation of fuel integrity, primary coolant pressure boundary, or primary containment boundary can he considered an abnormal occurrence. Also, ,
Exampie 12 of "For All Licensees" notes that recurring incidents which create I major _ safety concern can be considered an abnormal occurrence.
l Date and Place - October 29 and November 3, 1987, Perry Unit 1, a General l Electric-designed boiling water reactor, operated by Cleveland Electric Illum- 1 ir;ating Company, and lccated in Lake County, Ohio. '
Nature and Probable Consequences - On both of the above dates, Ur.it 1 experi- I enced a common mode failure during testing of the "D" steam line main steam l isolation valves (MSIVs). Both the inboard and outboard "0" MSIVs failed to close within the required time li; nit.
The MSIVs are part of the primary system and are provided to limit the release of radioactive materials to the environment or tn limit reactor vessel inven-tory loss during a design basis accident. Therefore, the failures of both the inboard and outboard "0" MSIVs to fast close were major degradations of the primary containment boundary and constitute a major degradation of safety-related equipment. The severity of a radioactivity release outside containment depends on the total time that radic;ctive steara escapes to the environment and the activity level in the reuctor coolant system during the event. The circum-stances associated with the events are as follows:
Perry Unit 1 has four main steam lines which carry steam from the reactor to l the plant's turbine generator. Each line has two automatically operated MSIV l valves -- one inside the reactor containment (inboard MSIV) and one outside I (outboard MSIV) -- which are designed to close quickly in the event of a steam !
line rupture. One valve closing would be sufficient to stop the flow of steam, which contains radioactivity and which could be released to the environment through a ruptured pipe, if the other valve did not close.
l The valves are periodically tested to assure that they meet the required test closure time of five seconds in what is called the "fast closure test." The valves are operated by pneumatic pressure and by compressed springs. Closure of the MSIVs is controlled by Automatic Switch Company (ASCO) dual solenoid valves.
On October 29, 1987, as part of the startup test program, the licensee tested l each of the eight MSIV valves (two on each steam line). The two valves on the l "0" steam line and one on a second line failed to close in the five second test period. Subsequent testing, however, showed the valves to meet the test criteria, and the unit remained in operation.
l 1
1 4
l l
1
On November 3, 1987, the valves were retested at the request of NRC Region III, prior tc the licensee perfnrming a full reactor isclation test (which involves simultaneous closure of all MSIVs) in its startup test program. The two "D" MSIVs again failed to close within the time limit - one closed in 18 seconds and the second did not close until the valve r, itch was cycled again in the control room after about 160 seconds. The U nsee commenced an orderly shut-down of the reactor to develop a disassembly and troubleshooting program.
The licensee found that the elastomer discs and 0-rings in the ASCO solenoid valves which controlled the failed MSIVs showed significant deterioration and I
degradation. It was concluded that the most probable cause was prolonged exposure of the'ASCO solenoid valves to a high temperature environment, result-ing from steam leaks in the vicinity of the valves. These valves were not in-tended nor designed for use in such conditions. The licensee replaced or re-built all of the ASCO solenoid valves associated with operation of the MSIVs.
The plant resumed operation on November 13, 1987, and successfully completed the full reactor isolation startup test.
On November 29, 1987, the licensee was performing a different test on the MSIVs
- e fast closure operability check - when it was determined that the inboard valve on the "B" steam line would not close and stay closed as it should during the test. Again, the licensee commenced an orderly shutdown of the reactor to determine the cause of failure.
Disassembly and inspection of the failed ASCO valve revealed the presence of a sliver, and two smaller particles of foreign material in the solenoid housing assembly. The foreign material was deteriorated body gaskets that had remained inside the valve when it was rebuilt in early November. The licensee concluded that the sliver of material caused mechanical binding of the solenoid valve.
The licensee replaced all of the dual solenoid valves (including the ones pre-viously replaced earlier in November). The plant was restarted on December 8, 1987.
The licensee performed an analysis of the radioactivity expected to be released to the environment if a steam line was not isolated for 18 seconds and a steam line break occurred outside reactor containment. Their analysis assumed that other mitigating systems (ECCS) performed as designed. The maximum primary !
coolant activity permitted by technical specification operating limits was as- !
sumed plus any additional activity which may be released as a result of reactor scram and ves;e1 depressurization, but no additional fuel failures as a result ;
of the postulated accident. Calculation results using the mass release used in j the FSAR and data used for a GE computer code were about 192 rem and 80 rem, i respectively, for thyroid dose at the exclusion boundary. Their calculation I predicted that Part 100 limits (300 rem) would have been exceeded if the line I remained unisolated for 79 seconds or greater. )
NRC staff, using conservative design basis loss-of-coolant accident source terms, determined that 10 CFR Part 100 limits would be exceeded for such an event with two redundant MSIVs failing to fast close within 18 seconds. This is the limiting case and would exceed the releases for other design basis events such as a steam line break outside containment. ,
1 5
In this case, the root cause of the events (deterioration'due to high tempera-ture) may have, over time, caused further deterioration of the valves that failed, as well as deterioration of additional valves. Therefore, there was serious concern regarding the reliability of the MSIVs to perform their safety function to mitigate the consequences of design basis accidents.
The NRC sent fact-finding Augumented Inspection Teams (AITs) to the site after both the November 3 and November 2c 1987 events in order to determine the I cause(s), conditions, and circumsta.ces of the MSIV failures.
Cause or Causes - For the October 29 and Ncvember 3, 1987 events, the AIT deter-mined that the most probable cause of the testing failures was a malfunction of the ASCO solenoid valves that operate the MSIVs due to deteriorating parts inside the solenoid valves. The deterioration occurred as a result of high temperatures caused by steam leaks in the vicinity of the valves. The deteri-orated parts impeded the operation of the solenoid valves.
For the November 29, 1987 event, the AIT concluded that deteriorated materials had remained inside the solenoid valve when it was rebuilt in early November.
These materials impeded the valve operation.
Actions Taken to Prevent Recurrence Licensee - The licensee increased its testing frequency for MSIVs with the new solenoid valve actuators and modified its maintenance procedures to require replacement or complete rebuilding rather than repair of any malfunctioning solenoid valves.
The licensee repaired the steam leaks and installed temperature monitors to de-tect any future steam leaks which could degrade the ASCO solenoid valves.
NRC - As previously mentioned, NRC AITs were sent twice to Perry Unit 1. While primarily fact finding missions, the AITs also identified a number of issues (which were given to the licensee for its consideration) which may be examined ,
for possible enforcement in subsequent inspections. The AIT report covering '
the October 29 and November 3, 1987 events was sent to the licensee an Jan- I uary 22, 1988 (Ref. 6). The AIT report covering the November 29, 19E7 event I was sent to the licensee on February 10, 1988 (Ref. 7). NRC issued Iaformation Notice No. 88-43, "Solenoid Valve Problems," on June 23, 1988, addressing the ;
technical details and concerns identified by this event and several otoers (Ref. 8).
Unless new, significant information becomes available, this item is considered )
closed for the purposes of this report. ;
88-3 C l acked Pipe Weld in Safety Injection System at Farley Unit 2 The following information pertaining to this event is also being reported con-currently in the Federal Register. Appendix A (see Example 2 of "For Commercial i
I 6
l Nuclear Power Plants") of this report notes that a major degradation of fuel l integrity, primary coolant pressure boundary, or primary containment boundary 1 l' can be considered an abnormal occurrence. The specific reportability consid-eration.of this pipe weld failure is due to the possible generic implications ,
l from a thermal cycle mechanism not previously experienced in the industry.
i In addition, the event raised possible generic implications; Example 12 of "For i All Licensees" of Appendix A notes that incidents with implications for similar facilities (generic incidents), which create major safety concern, can be considered an abnormal. occurrence.
Date and Place - December 9, 1987; Farley Unit 2, a Westinghouse-designed, 3-loop pressurized water reactor, operated by the Alabama Power Company (the licensee) and located in Houston County, Alabama.
Nature and Probable Consequences - An unisolable leak was discovered in a safety injection system (SIS) pipe while Unit 2 was being restarted after a refueling outage. The significance of the event is that a generic safety ques-tion may exist in that more than one unisolable emergency core cooling system (ECCS) pipe failure may occur. Subjecting the flawed piping to excessive stress, induced by a seismic event, water hammer or other cause, conceivably could result in simultaneous double-ended failure of more than one ECCS pipe.
The circumstances associated with the event are as follows:
The licensee had noted increased moisture and radioactivity (1000 to 2000 counts per second) within containment. The unidentified leak rate for the reactor coolant system (RCS) was determined to be 0.7 gpm. The technical specifica-tions permit leakage up to 1.0 gpm. After entering containment to identify the location of the leak, licensee personnel determined that the leak could not be isolated. The reactor was taken to cold shutdown to facilitate repair.
1 After failing to detect crack indications by liquid penetrant testing (PT) l methods, the licensee located an indication of a through-wall crack using ultra- l sonic testing (UT) in the 6-inch ECCS piping connected to the cold leg of RCS j loop B. The indication was located at a weld connecting an elbow and a hori-zontal pipe section between the isolation check valve (V051B) and the reactor coolant piping. The indication was on the bottom of the pipe and extended cir-cumferentially 60 degrees in both directions trom the bottom inside of the pipe.
The crack, which was confirmed by radiography, extended through the wall for approximately one inch at the center of the indication and extended about six )
inches on the inside piping surface.
The licensee initiated a progressive UT exami ation plan for those welds adja-cent to the cracked weld and for all similar system welds on all three 1) ops in both Unit 1 and Unit 2. The examination included three welds in Unit 2 loop B cold leg of the SIS, upstream of valve V0518. The examination did not i reveal any relevant UT indications.
A complete system walkdown of all three SIS loops in both Units 1 and 2, per-formed to determine if any pipe restrictions, leakage, or other problems were evident, identified no additional significant problems.
The faulted piping assembly (containing the vertical pipe, the elbow and the horizontal pipe) was shipped to the Westinghouse Research and Development 7
l Laboratory in Pittsburgh,-Pennsylvania for evaluation. The metallurgical exami-nations were performed on a three inch wide ring section containing the weld joint exhibiting the circumferential cracking. The evaluations included sur-face examinations, metallographic examinations, fractographic examinations and chemistry evaluations.
Based on the results of these evaluations, it was concluded that the observed cracking in the SIS pipe to cibow joint was caused by a high cycle fatigue mechanism. Stress concentration at the knee of the standard counterbore region and normal surface machining grooves contributed to the crack 1nitiation. ;
Neither weld defects nor corrosion attack contributed in any significant way to
- the joint failure.
l l The visual and metallographic examinations showed that the weld had failed as I a result of fatigue after roughly one million stress cycles. The licensee examined the operating records and determined that the number of stress cycles l imposed, by starting up and shutting down the plant and by SIS initiation, was significantly less than the relevant design criteria.
l On the basis of this information, the licensee postulated that the weld stress loads were either: (1) thermal and created by either valve leakage or convective l
flow cells, o.- (2) mechanical and created by flow-induced vibrations. To test these postulations, the licensee replaced the failed piping and installed sensors for temperature and acceleration near the location of the failed weld and on the other side of the check valve at a 1ccation about two feet upstream of the failed weld. The licensee also installed sensors at similar locations on the ECCS pipe connected to RCS loop C.
Following repair, the reactor was restarted and taken to steady state full power operation. Data obtained demonstrated that there was an adverse temperature distribution in the loop B ECCS piping. The maximum circumferential temperature difference at the location of the failed weld was about 215 F. Further, the temperature at the bottom of the pipe fluctuated by as much as 30 F in 30 seconds.
This difference in temperature distribution was caused by failure of a valve in the bypass pipe around the boron injection tank to seat properly. This leaking valve is believed to have set up a thermal cycling leading to failure of the seld. Leakage through the valve apparently caused the check valves in the loop B ECCS pipe to partially open, or chatter, admitting relatively cold coolant to )
j the unisolable portion of the pipe between the nozzle and the first check valve. j Thus, cold water on the bottom of the pipe created thermal stress which was '
cyclic in nature causing excessive stresses and a resultant pipe crack.
I Data from the temperature sensors for loop C ECCS piping indicated that the I check valves in that pipe were not chattering ano that the temperature distri- )
bution was normal. Further, none of the accelerometers indicated adverse l mechanical stress cycling. I The event may have generic safety implications for other plants which have dual l purpose pumps used for charging the RCS with coolant during normal operation l and for injecting emergency core coolant at high pressure following an accident.
During normal operation, with one of the pumps providing charging flow to the RCS via the normal charging piping and with a leaky valve allowing coolant to flow to the ECCS manifold, pressure in the manifold will exceed RCS pressure. I 8
I l
This would allow check valves in the ECCS piping to open admitting relatively cold coolant to the RCS. The flow rate via this additional path or paths would be determined by the amount of leakage through tne leaking valve. If the check valves in more than one ECCS pipe open, then more than one unisolable ECCS sys-tem leak may occur. Subjecting the weakened piping to excessive stresses in-duced by a seismic event, water hammer, or some other cause conceivably could result in simultaneous double-ended failure of more than one ECCS pipe.
Cause or Causes - As discussed in more detail above, the cause of the pipe cracking was attributed to valve leakage which resulted in thermal cycling of the pipe.
Actions Taken to Prevent Recurrence Licensee - The initial actions were to replace the piping associated with the failed weld. The 90 elbow and straight runs of pipe at each end of the cracked elbow to pipe weld were removed from the SIS cold leg on loop B.
The initial corrective actions taken to prevent recurrence were to provide a means to assure that the pressure upstream of the check valve does not exceed RCS pressure, thereby reducing the possibility of its partially opening or chattering. Possible changes for long-term corrective actions are under review.
NRC - The licensee's examinations and initial corrective actions were reviewed by NRC Region II personnel during an inspection on December 12-16, 1987. The inspection identified no violations of NRC requirements. The inspection report was forwarded to the licensee on January 29, 1988 (Ref. 9). On January 15, 1988, the licensee met with the NRC staff to review their corrective actions and to b e ter define the generic aspects of the issue. As a side issue, the licensee's ECC analysis required licensee and Westinghouse reevaluation of a postulated double-ended SIS pipe break. The licensee s corrective actions and the ECCS reanalysis remain under review by the NRC staff.
On January 27, 1988, the NRC issued Information Notice No. 88-01, "Safety injection Pipe Failure," to all holders of operating licenses or construction permits for nuclear power reactors to inform them not only of the Farley event, but also of the potential generic problem concerning the reliability of piping in safety-related systems due to valve leakage which rr.ay result in thermal cycling of the piping (Ref. 10). On June 22, 1988, the NRC issued Bulletin No. 88-08, "Thermal Stresses in Piping Connected to Reactor Coolant Systems,"
to request that licensees: (1) review their reactor coolant systems (RCSs) to identify any connected, unisolable piping that could be ' subjected to temperature distributions which would result in unacceptable thermal stresses; and (2) take action, where such piping is identified, to ensure that the piping will not be subjected to unacceptable thermal stress (Ref.11).
Future reports wiH be made as appropriate.
A A A AA A A A 9
l FUEL cycle FACILITIES (Other Than Nuclear Power Plants)
- The NRC is reviewing events reported by these licensees. For this report, the NRC has not determined that any events were abnormal occurrences.
OTHER NRC LICENSEES (Industrial Radiographers, Medical Institutions, Industrial Users, etc.)
There are currently about 9,000 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. For this report, the l NRC has determined that the following events were abnormal occurrences.
l 88-4 Diagnostic Medical Misadministration The following information pertaining to this event is also being reported con-currently in the Federal Register. Appendix A (see the general. criterion) of this report notes that an event involving a moderate or more severe impact on public health or safety can be considered an abnormal occurrence.
Date and Place - November 23, 1987; Veteran's Administration Medical Center, Albuquerque, New Mexico.
Nature and Probable Consequences - A patient was administered 50 millicuries of technetium-99m (as sodium pertechnetate) instead of 3 millicuries of thallium-201 prescribed by the physician.
The purpose of the administration was for a Myocardial Perfusion Stress Test.
The licensee reported that there were no deleterious effects to the patient.
The licensee calculated that the patient incurred the following doses: thyroid
- 6.1 to 10.2 rads; stomach - 5.1 to 15.3 rads; colon - 5.1 to 15.3 rads; gonads - 0.5 to 2.0 rads; and whole body - 0.5 rad.
Cause or Causes - The misadministration was caused by a student technologist selecting the wrong syringe from the dosage cart.
Actions Taken to Prevent Recurrence Licensee - The student technologist was reprimanded, new procedures for radio-pharmaceutical labeling and handling will be implemen'.ed, personnel will be retrained, and the supervision of personnel will be ilmroved.
10
NRC - NRC Region IV telephoned the radiation safety officer reporting this' mis- l
- administration for additional details'on the incident. Those details were sub- i sequently provided by a February-1,'.1988 memorandum from the licensee. :The-incident will be reviewed during a special NRC inspection at the center, i
Unless new, significant information becomes available, this item is considered closed for the purposes of.this report.
88-5 Breakdown in Management Controls:at Georgia Institute of Technology Research Reactor Facility The-following information pertaining to this event is also being reported con-currently in the Federal Register. Appendix A (see the third general criterion) of this report notes that major deficiencies in design, construction ~, use of, or management controls for licensed facilities or material can be considered an abnormal occurrence.
Date and Place - This occurrence addresses licensee performance over a period of time until January 20, 1988, when the NRC issued an Order Modifying License (effective immediately) to the Georgia Institute of Technology (Georgia Tech) regarding their research reactor (GTRR). The GTRR is a 5 megawatt (thermal)-
facility, located in Atlanta, Georgia, and utilized for teaching and research including the performance of irradiation experiments.
Nature and Probable Consequences - The NRC Order required the licensee to cease using the reactor facility for irradiation experiments until certain conditions are met and the NRC approves the resumption of irradiation experiments (Ref.12).
NRC concerns with regard to the licensee's management control has been the sub-ject of enforcement actions in the past. An inspection conducted February 9-23, 1987, which included a review of the licensee's operations program, identified numerous failures to comply with NRC regulatory requirements. The areas of ,
non-compliance included inadequate procedures, failures to follow procedures, dnd problems in keeping adequate records documenting compliance with NRC re-quirements. Based on these inspection findings, the NRC raised concerns about programmatic weaknesses in the licensee's implementation of Technical Specifi-cation requirements.
Inspections conducted February 17-23 and April 7-10, 1987, which included a review of the licensee's operations and radiation protection programs, also identified s'gnificant failures to comply with NRC regulatory requirements in the same areas described above. The findings of these inspections clearly in-dicated the licensee's need for improved management control to ensure adherence 4
to NRC requirements and safe performance of licensed activities. On May 4, 1987, an enforcement conference was held at the NRC Region II office in which the licensee outlined steps to be implemented to improve management controls over operations and health physics at the facility to assure safe operation.
These actions included a change in the research facility's organizational structure, i
11
,,,.-..v. w-.,, - - , . -.-w-,,m-,-- -,- ,m - -- . - - - , - - - - --*,,-..--y- -- ,- . . - _ - , . - - - -
The events leading to issuance of the NRC Order were identified during recent inspections which showed that the licensee's actions have not been fully suc-cessful and indicated that management control problems continue. On Decem-ber 16, 1987, while reviewing management reorganization concerns for the GTRR program, an NRC inspector learned of a contamination event which occurred at the reactor facility during the week of August 17, 1987. The event involved the improper opening of an irradiated material container which resulted in the release of radioactive contamination within the reactor containment building.
At the time of the December 1987 NRC inspection, a detailed description and evaluation of the event had not been prepared by licensee or management staff.
This inspection was continued on January 4-5, 1988, and a team inspection was conducted from January 14-EL 1988. I l
The inspection findings (Ref. 13) revealed that the experiment conditions and l manipulation of the experiment materials resulted in unexpected elevated radi-ation levels from the experiment container and also the unmonitored release of i l
cadmium-115 in the reactor building. The dose rate at one foot from the I experiment material was approximately 3 rem per hour on August 18,1987, ar;d
, qualitative measurements of radioactive contamination indicated levels on 1 Masolin wipes of approximately 20 millirem per hour on August 19, 1987.
f The following violations were identified from the inspection findings: failure )
to have adequate procedures and failure to follow procedures for handling and !
I manipulating experiment material and for surveying and evaluating potential '
radiological hazards; failure to conduct adequate radiation surveys of the reactor building and GTRR personnel and their personal property for evaluation of exposure to radioactive material; failure to .onduct adequate air sampling and bioassay analyses for evaluation of personne, exposure to airborne radio-active material during experiment and decontamination activities; and failure to document and maintain records of radioactive material contamination surveys.
At the time of the inspection the licensee had failed to complete a thorough review of the August 1987 contamination event regarding its cause or causes, nor had any corrective measures been implemented as of January 5, 1988 to pre- )
vent recurrence during future experiments.
The issuance of the NRC Order was a direct result of NRC concerns over the {
licensee's past performance, their unsatisfactory slow rate of improvement, and, i most importantly, the licensee's lack of management control needed to assure 1 that continued irradiation experiments would not result in more significant safety problems.
Cause or Causes - The root cause was a lack of regard for and adherence to procedures, and a lack of management control over licensed activities.
Actions Taken to Prevent Recurrence Licensee - The licensee voluntarily shut down the GTRR on February 15, 1988.
The enforcement history and recent inspection findings were discussed with the licensee at an enforcement conference held at the NRC Region II office on February 23, 1988. The licensee addressed the violations identified and pre- l l
sented an action plan directed towards upgrading their operations and health l 12 1
physics programs. Also, the licensee committed not to restart the reactor without NRC concurrence.
NRC - The January 20, 1988 NRC Order (Ref. 12) required the licensee to immed-iately suspend certain activities under its NRC license until requirements of the Order are satisfied which includes: an assessment of management controls; review whether any other events similar to the August 1987 incident have oc-curred; assessment of personnel exposures for the August 1987 event, and any
, other similar events, and associated cleanup activities; review of health l physics and operating procedures; identification of corrective actions and schedule for implementation; and development and implementation of necessary training programs.
On March 17, 1988, the NRC issued to the licensee a Confirmatory Order Modifying License (effective immediately) confirming the licensee's commitments made at the February 23, 1988 enforcement conference (Ref. 14). The Order did not modify the conditions of the January 20, 1988 Order.
The inspection report forwarded to the licensee on February 10, 1988 (Ref. 13) identified several apparent violations of NRC requirements. However, no formal Notice of Violation was issued at the time since the apparent violations are under consideration for escalated enforcement action.
Future reports will be made as appropriate. ;
1 88-6 Release of Polonium-210 from Static Elimination Devices Manufactured by 3M Company The following information pertaining to this event is also being reported concurrently in the Federal Register. Appendix A (see Example 12 of "For All Licensees") of this report notes that a series of events (where individual events are not of major importance), and incidents with implications for simi-lar facilities (generic implications), which create a major safety concern, !
can be considered an abnormal occurrence. In addition, the first general crite- l rion notes that a moderate release of radioactive material licensed by or other- l wise regulated by the Commission can be considered an abnormal occurrence.
Date and Place - January 21, 1988; Ashland Chemical Company (Ashland) plant in Easton, Pennsylvania, and various other locations. l l
Nature and Probable Consequences - On January 22, 1988, the radiation safety consultant for Ashland reported to NRC Region I that radinctive contamination had been discovered at their plant on the previous day. Suosequent determina-tion of the cause of the contamination led to extensive investigations which showed that similar contamination problems existed at many plants in many states. l The cause of the contamination involved the failure of static elimination devices manufactured by the Minnesota Mining and Manufacturing Company (3M Company),
St. Paul, Minnesota, contailing polonium-210 (Po-210). Po-210 decays by emis-sion of a 5.3 MeV alpha particle and has a half life of 138 days. The devices 13
in 3M Company's Model 900 serias are used to ionize a stream of air in order to remove static electricity. Even-numbered models in the 900 series (Models 902, 902F, 906, and 908) are used in conjunction with compressed air; odd-numbered models (Models 905, 907, and 909) are used with blown air. (The devices used by Ashland were of the even-numbered models and were connected to a source of compressed air.) The 3M Company also manufactures bar-type static elimination devices.
l The Po-210 in these devices is contained in microspheres of zirconium pyro-phosphate that have been plated with nickel and are held in place with an epoxy adhesive. The microspheres are about 0.001 inch diameter and are hard, dense, and insoluble. The 3M Company manufactures the devices under NRC License No. 22-00057-06 and distributes then to general licensees (such as Ashland) under NRC License No. 22-00057-32G as permitted by the general license provi-sions of 10 CFR S31.5. Because general licensees lack the expertise tu leak-test the devices, the 3M Company requires that the devices be leased for a one-year period, returned to the 3M Company, and tested for leakage of radioactivity.
The discovery at Ashland resulted from an investigation occasioned by a com-plaint from one of Ashland's customers that its product was radioactively con-taminated. Subsequent to discovery of contamination at its Easton plant, Ashland conducted surveys at its other plants. On January 23, 1988, its plant at Dallas, Texas, was also found to be contaminated; its other plants were free of contamination. Surveys were begun at other plants using the 3M devices, and contamination was discovered at a KTI Chemical Corporation plant in Carroll-tor, Texas, oa January 28, 1988. On February 1, 1988, two beverage plants in Dallas, Texas, were found to bo contaminated. At this time, the U. S. Food and Drug Administration (FDA) began investigation for possible product contamina-tion at plants using the 3M Models 902, 902F, 906, and 908 devices where simi-lar device failure had been found in preceding years. No contaminated product was found.
Inspectors from the NRC examined 3M Company records of quality assurance on '
returned devices for the years 1986 and 1987 and discovered a large number of devices that were leaking upon return to the 3M Company. Most of these fail-ures had not been reported to the NRC because the 3M Company believed that the failure was caused by some kind of damage to the device. Inspectors from the NRC and individual states have now (as of late April 1988) surveyed scores of plants using 3M Company static elimination devices and have identified a large number of plants that show contamination levels exceeding 0.005 micro-curies. It is expected that the total number of such plants may be several hundred. For each plant in which contamination was found, the contamination was not widespread; rather, it appears to result from discrete Po-210 particles.
Po-210 emits alpha radiation which will not penetrate the outer layers of the skin. The size and density of the Po-210 microspheres indicate that they are not respirabie and, if ingested, they are expected to pass through the digestive tract in a short time without significant release to the bloodstream because they are very insoluble. Bioassay of workers at Ashland's plants at Easton, Pennsylvania and Dallas, Texas, demonstrated no uptake of polonium. No adverse health effects are expected because of the defective static elimination devices, and none have been found.
14
Cause or causes - No cause for failure of the static elimination devices has been ascertained. A postulated cause is moisture or solvents in the environment that affect the epoxy adhesive, which holds the radioactive material in the i device. J Actions Taken to Prevent Recurrence l Licensee (3M Company) - The licensee's investigation of the cause of the fail- I ures and possible corrective actions continues. The licensee is carrying out the requirements of the below described NRC Orders.
General Licensees - Plants where contamination has been found have been, or are being, cleaned up and returned to production. All 3M Company devices are being returned to the manufacturer (except as permitted by the February 18, 1988 NRC Order described below). As of mid-April 1988, about half of all static elimi-nation devices have been returned to the 3M Company (this includes 86% of the devices used in food, beverage, pharmaceutical, and cosmetic applications).
Of the devices returned,1.84% had leakages greater than 0.005 microcuries.
NRC - On January 25, 1988, the NRC ordered the 3M Company to suspend distribution of Models 902, 902F, 906, and 908 devices; to inform users of these devices of the problem discovered by Ashland; to survey a suitable sample of users to as-certain the extent of the problem; and to determine the cause of the failure of the devices (Ref. 15). On February 5, 1988, the NRC issued a confirmatory orde;, confirming the 3M Company's commitments to remove all devices with the above model numbers from applications related to the packaging of food, bever-ages, cosmetics, and pharmaceuticals (Ref. 16). On February 12, 1988, the NRC ordered the 3M Company to remove all static elimination devices (not just the 900 series) from all applications relating to the production and packaging of food, beverages, cosmetics, and pharmaceuticals (Ref. 17). These actions were coordinated with the FDA.
On February 18, 1988, the NRC ordered the 3M Company to suspend transfer of all static elimination devices using Po-210; to instruct users of the devices to return them to the 3M Company; to test returned devices for leakage and report any leakage to the NRC (or Agreement State) and the user; and to report the status of these activities to the NRC every 30 days (Ref. 18). The February 18, 1988 NRC letter also ordered all general licensees using the 3M Company static elimination devices to suspend use of the devices and to return them to the 3M Company as soon as feasible but no more than 90 days from the date of the Order. An exception was made for continued use of the devices, under certain conditions, for applications where use of the device is essential for work place safety (e.g., where static electricity may pose a significant fire, explosion or other hazard). On April 13, 1988, the NRC Order of February 18, 1988 was modified to allow the 3M Company to respond to the show cause order by July 18, 1988.
The NRC actions were coordinated with the Agreement States. As of March 25, !
1988 (the latest date, as of April 30, 1988, for which an estimate of total Agreement State efforts are available) the Agreement States had applied 8,224 professional staff-hours (equivalent to about 4.5 full-time staff persons) to conduct on-site surveys of facilities identified as possessing these sources. ,
The Agreement States took appropriate enforcement actions when contamination i 15
was found, and their survey data were incorporated into the NRC data base which served as a basis for NRC enforcement decisions.
Many of the non-Agreement States assisted NRC by surveying NRC generally licensed users of these devices at NRC's request and their survey data were also used by NRC in assessing the scope of the problem.
The International Atomic Energy Agency (IAEA) and regulatory and safety authori-ties in 44 countries were advised through the Department of State's cable sys-tem and directly via airmail of NRC concerns and subsequent actions related to the 3M static elimination devices. The IAEA and the safety contacts in all 44 countries received copies of NRC orders and background material on the defective devices in two separate mailings, dated February 12 and 19, 1988. Subsequent mailings also were sent to update the IAEA and foreign safety contacts on devel-opments in this area. On March 31, 1988, to prevent further exports of the de-fective devices, NRC issued an order confirming that 3M would not be permitted to export any polonium-210 static elimination devices under the general license for export in 10 CFR Part 110.
Future reports will be made as appropriate.
A A A A A A A A 88-7 Therapeutic Medical Misadministration The following information pertaining to this event is also being reported concurrently in the Federal Register. Appendix A (see the general criterion) of this report notes that an event involving a moderate or more severe impact on public health or safety can be considered an abnormal occurrence.
Date and Place - February 4, 1988; Medical X-Ray Center, Sioux Falls, South Dakota.
Nature and Probable Consequences - A patient was administered 7.5 millicuries of phosphorus-32 (as sodium phosphate) instead of 4.0 millicuries of the same radiopharmaceutical prescribed by the physician.
The purpose of administration was to treat polycythemia vera (excess blood red platelets). As a result of the misadministration, the patient received a dose of about 270 rads and 75 rads to the bone marrow and whole body, respectively, instead of about 145 rads and 40 rads, respectively, had the prescribed amount of pharmaceutical been administered. There were no apparent effects to the patient. The licensee reported that blood counts will be followed for several weeks post-therapy and that the last report on February 16, 1988, showed normal blood elements.
Cause or Causes - The misadministration was caused by a miscalculation of the dose by the technician.
Actions Taken to Prevent Recurrence Licensee - The technician administering the dose was reinstructed in the proper technique for calculating therapy doses and for reviewing the written physician orders prior to administering the doses.
16
NRC - NRC Region IV-telephoned the radiation safety officer reporting this mis-administration for additional information and assurance that corrective action had been taken. The incident will be reviewed during the next NRC inspection at the medical center.
Unless new, significant information becomes available, this item is considered closed for the purposes of this report.
88-8 Therapeutic Medical Misadministration The following information pertaining to this event is siso being reported con-l currently in the Federal Register. Appendix A (see the general criterion) of this report notes that an event involving a moderate or more severe impact on l public health or safety can be considered a') abnormal occurrence.
i l Date and Place - Discovered on February 15, 1988; St. Joseph's Hospital, Mil-waukee, Wisconsin.
Nature and Probable Consequences - On February 23, 1988, NRC Region III was notified by the licensee that an 86 year-old patient with a 10-yur history of-bladder cancer received a cobalt-60 therapeutic radiation dose of 2000 rads to the wrong side of his pelvis.
On January 19, 1988, the patient was admitted to the hospital with a severe right rib pain. A CAT scan of his abdomen (January 20), a bone scan (Jan-uary 25) and mid-spine and pelvit scan (January 28) confirmed the patient had a metastatic cancer. The Radiation Oncologist determined that two local areas should be treated, the spine and the left pelvis. Beginning February 3, 1988, i the licensee commenced treating the patient with cebalt-60 with a prescribed i dose of 5000 rads to the spine (20 treatmeats of 250 rads each) and 4000 rads to the pelvis (20 treatments of 200 rads each).
On February 15, after 10 treatments totalling 2000 rads, the Dosimatrist became suspicious that an error had been made and that the wron-; . cide of the patient's pelvis (the right side) had been treated. This was confirmed on February 16 b'; j the Radiation Oncologist. The patient and referring physician were notified, i and treatment on the left side of the pelvis was begun the following day.
In evaluating the event, the licensee said the patient had "documented bone destruction of the dorsal spine and left pelvis, and therefore, it is most-probable there is disease throughout all the pelvic a~reas. The patient also had reported right side pain prior to the therapeutic treatment. Therefore, the dose given to the palliative right pelvis, rather than having caused him harm, could be considered prophylactic treatment."
In a report to NRC Region III dated March 9, the licensee saia it was unclear whether the right-side treatment was "inadvertent or a conscious decision due ta a misread of the bone scan." According to the referring physician, the patient exhibited no adverse afteraffects as a result of the misadministration.
17 l
l i
Cause or Causes - The event is attributed to personnel errors and inadequate procedures. The radiation therapist had prescribed treatment to the dorsal spine and left pelvis. However, a therapy technologist set the patient up and marked the right pelvis. Neither the physicist, who performed the dose calcula-tions, nor the chief technologist, who performed the treatment, noted the dis-crepancy between the treatment plan and the prescription. In addition, the dosimetrist, while performing a weekly chart check, failed to notice the error.
About 10 days later, the dosimetrist again performed a chart check and noticed the discrepancy. She brought this to the attention of the physicist, who then discussed it with the radiation therapist. Trect=:r.t to the right pelvis was terminated at 2000 rads.
Actions Taken to Prevent Recurrence Licensee - The licensee agreed to develop and implement procedures which require its staff to thoroughly review all aspects of therapy prescriptions and treat-ment parameters when the following events occur: (1) during the initial dose calculations, (2) just prior to initial treatment, and (3) during weekly chart checks.
NRC - A region-based inspector went to the hospital to review the it.;ident on March 3 and 4, 1988. The NRC also retained an NRC medical consultant to re-view the misadministration. In the meantime, Region III conferred with the licensee on corrective action, and the licensee agreed to the above procedural changes. In a letter confirming the licensee's course of action dated March 10, 198P., Region III also requested that the procedural changes be formalized as a license amendment (Ref. 19).
On April 14, 1988, a Notice of Violation was issued to the licensee; the therapy misadministration had not been reported to the NRC Regional Office until seven days after discovery, contrary to 10 CFR S 35.33(a) which requires telephone notification within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
This item is considered closed for the purposes of this report.
- A A A A 88-9 3Si n_ificant Widespread Breakdown in Radiation Safety Program at Case Western Reserve University Research Laboratories The following information pertaining to this event is also being reported con-currently in the Federal Register. Appendix A (see Example 11. of "For All Licensees") of this report notes that serious deficiency in management or pro-cedural controls in major areas can be considered an abnormal occurrence. In addition, the first general criterion of Appendix A notes that a moderate release of radioactive material licensed by or otherwise regulated by the Commission can be considered an abnormal occurrence.
Date and Place - This occurrence addresses licensee performance over a period of time until February 26, 1988, when the NRC proposed imposing a $10,000 fine on Case Western Reserve University, Cleveland, Ohio.
4 18
Nature and Probable Consequences - The proposed fine was for numerous violations of NRC rcquirements in the licensee's radiation safety program for its research laboratories, indicating a significant breakdown in its management control pro- !
gram (Ref. 20). The violations were in the licensee's research programs, not i medical care and treatment of patients. The circumstances associated with the l enforcement action are as follows: j On November 8, 1987, the NRC Region III office received a news media inquiry concerning the radioactive contamination of a research laboratory at Rainbow Babies and Children's Hospital in Cleveland, Ohio. The laboratory, although located at the hospital, was under the NRC license of Case Western Reserve University. Telephone discussions on November 9, 1987 with the licensee de-termined that a licensee consultant had identified tritium and carbon-14 con-tamination in the laboratory (Diabetes Laboratory) and that it was being I decontaminated. 1 It was learned later, through subsequent telephone conversations with the li- -
censee, that the contamination in the laboratory was more widespread than ini-l tially found. On November 17, 1987, the NRC began an inspection to review the j circumstances of the contamination and to determine if the problems associated ,
with the laboratory were indicative of additional problems at other licensee l laboratories.
1 The initial and subsequent NRC inspections during November and December 1987 identified about 20 violations of NRC requirements, involving the training of laboratory personnel, radiation safety practices, and control and oversight of the laboratories using radioactive materials. Several of the violations (i.e., l failure to calibrate radiation survey instruments, failure to perform a contami- '
nation survey, failure to perform leak tests of sealed radiation sources, and ;
evidence of food and beverage consumption in laboratories) were similar to i violations identified during a May 1986 NRC inspection at the licensee's facil-ities. Therefore, the licensee's corrective actions, taken following the 1986 inspection, were insufficient to prevent a recurrence of the violations.
l Based on the inspection findings, it appeared that the University was unable l to keep track of the number of laboratories engaged in licensed activities, was not controlling the required training of workers handling radioactive ma-terials, and was unable, through its radiation safety committee, to assure com-pliance with NRC requirements and license commitments. The latter became evi-dent when the University found it necessary to contract with an outside consul-tant to perform required radiation surveys and audits which the University ra-diation staff could not complete in a timely manner. (It was a survey performed by the consultant which initially identified the contamination of the Diabetes l Laboratory.) l In regard to the Diabetes Laboratory, the inspection indicated that no single incident appeared to have contributed to the contamination; rather the wide-spread, low-level contamination in the laboratory was caused by inadequate han-dling procedures (the technicians had not been adequately trained) and a lack of contamination surveys. There was no evidence that any workers or members of the public received a significant radiation exposure as a result of the contam-ination incident or of the violations found in the licensee's radiation safety 19
program. Bioassay tests on the two Diabetes Laboratory technicians showed no detectable indication of ingestion or inhalation of radioactive material.
Cause or Causes - The failure to adequately correct past violations identified in a May 1986 inspection, as well as the numerous violations identified in the November - December 1987 inspections, demonstrated a serious, widespread break-down in the management of the licensee's radiation safety program.
Actions Taken to Prevent Recurrence l
Licensee - The licensee conformed to the various NRC actions described below. 1 Following suspension of all NRC-licensed work (which affected about 350 laboratories), the licensee retained an interim Radiation Safety Officer, pro-vided training to laboratory workers, and expanded the work of its consultant to review all laboratories for compliance with University and NRC requirements.
Extensive programmatic changes were made to the licensee's radiation safety program. Based on these changes, on December 8, 1987 the NRC authorized the gradual lifting of the suspension as each laborat'ory was checked and found to comply with NRC requirements. By mid-February 1988, work had been permitted to resume in all laboratories, except for the Diabetes Laboratory. The latter laboratory required final decontamination work before its suspension could be lifted.
During March 1988, the licensee hired a new Radiation Safety Officer to oversee NRC-licensed activities.
NRC - When the initial inspection revealed violations of NRC requirements, NRC Region III issued a Confirmatory Action Letter on November 20, 1987, document-ing the University's agreement to accelerate its radiation survey program and to direct each laboratory supervisor to assure that the requirements were being followed (Ref. 21).
Based on further inspection findings, a second Confirmatory Action Letter was issued on November 25, 1987, confirming the suspension of NRC-licensed work (Ref. 22).
At the time work was authorized to resume on December 8, 1987, the NRC issued a license amendment to include the modifications and improvements to the radia-tion safety program adopted by the licensee.
On February 26, 1988, the NRC issued to the licensee a Notice of Violation and Proposed Imposition of Civil Penalty in the amount of $10,000 for the numer-ous violations identified during the inspections (Ref. 17). The inspection reports were also enclosed. The violations were categorized as Severity Level II (on a scale in which the most significant and least significant are cate-gorized as Severity Levels I and V, respectively). The base value of a civil penalty for a Severity Level II violation is $4,000. This was increased to
$10,000 because of the licensee's poor prior performance in their radiation safety program and the failure to take adequate corrective actions subsequent to the identification of violations during the most recent events. The licensee has paid the civil penalty.
20
b The NRC will continue to monitor the licensee's performance through periodic inspections. ?
This item is considered closed for the purposes of this report. ,
- A A A A A A A AGREEMENT STATE LICENSEES Procedures have been developed for the Agreement States to screen unschedJled incidents or events using the same criteria as the NRC (see Appendix A) and ,
. report the events to the NRC for inclusion in this report. During the first l calendar quarter of 1988, an Agreement State (Texas) reported the following l abnormal occurrence to the NRC. 4 AS88-1 Radiation Injury To Two Radiographers Appendix A (see Example 1 of "For All Licensees") of this report notes thet an exposure of the feet, ankles, hands, or forearms of any individual to 375 rem or more of radiation can be considered an abnormal occurrence.
Date and Place - November 16, 1987; North Shore X-Ray and Testing Company, Houston, Texas.
Nature and Probable Consecuences - On December 8,1987, the licensee reported to tne Texas Bureau of Raciation Control (the Agency) film badge overexposures (whole body) of 5.0 rem and 5.2 rem to a radiographer and a radiography helper, respectively. Based on subsequent follow-up by the Agency and referral of the men to a physician experienced in radiation exposures, it was determined that each man had received a radiation burn to the skin of one ankle. The exposures associated with the burns are estimated to be in the range of 860-1940 rem. A 1 summary of the Agency's investigation is as follows. j On December 14, 1987, a physician called the Agency to report that the radiog-rapher had a radiation burn on his left ankle. On December 15, 1987, the Agency began the investigation to determine when and how the radiation injury occurred.
Ager investigators interviewed both the radiographer and the radiography helc s.' Both men had been working as radiographers for a number of years.
Neiti,er of the individuals could recall any circumstance that would have resulted in a radiation burn. During the interview, the helper indicated he had observed a reddened area on his right ankle. He later was examined by a !
physician (the same physician who examined the radiographer) and was also '
diagnosed as having a radiation burn. Both men indicated noticing evidence of injury around the first of December, although neither could recall an exact date.
Based on extensive interviews, review of records, and other investigations, the Agency concluded that the exposures had most likely occurred on November 16, 1987. On February 9, 1988, the Agency sent an investigator to attempt to re-enact the bench work performed on November 16, 1987. The reenactment was speculative since the radicgrapher and the helper could not recall details of 21
the work. Set-up times for the work could vary from a few seconds to several minutes. Assuming just a brief contact with the source tube connector, the exposures to each ankle was estimated to be in the range of 860-1940 rem.
Cause or Causes - Evidence indicated that the radiographers were working on the same job at the time of injuries and, based on the company records, they were working with a 125 curie iridium-192 source. Statements made by the radiographers indicate the survey meter was not in an ideal location to observe a significant radiation field. It is also possible that it was not observed after the source was returned to its shielded position each time. In any case, radiation surveys were not properly conducted after each exposure.
The only explanation for the exposures considering their location, on the right ankle of the helper and the left ankle of the radiographer, is that the source somehow came to an exposed position, probably just inside the outlet nipple, and this was occurring after each exposure or occurred once and the radiographers came in contact with the outlet nipple resulting in the radia-tion burns, apparently as the men stood on either side of the source tube.
Actions Taken to Prevent Recurrence Licensee - The licensee discussed the incident with the individuals and with other employees performing radiography, and stressed the importance of using the survey meter each time the source is cranked out and back in. Management did not feel there was much they could do to prevent these incidents except to stress the use of detection equipment.
A enc - The licensee was cited for allowing the overexposure to occur and for a1 ure to use survey instruments. One of the radiographers passed the radiog-raphy qualification exam after the incident occurred and was issued an identi-fication card prior to completion of the investigation. The radiographer has been notified that he may be required to show cause why the identification card should not be revoked or suspended, if he violates the Agency's regulations for control of radiation in the future. The identification card is a requirement for qualification as a radiographer in Texas.
This item is considered closed for the purposes of this report.
22
REFERENCES
- 1. Letter from L. J. Callan, Director, Division of Reactor Projects, NRC Region IV, to R. L. Andrews, Division Manager, Nuclear Production, Omaha Public Power District, forwarding NRC Inspection Report No. 50-285/87-27 and a meeting agenda for an October 29, 1987 enforcement conference, Docket No. 50-285, October 23, 1987.*
- 2. Letter from L. J. Callan, Director, Division of Reactor Projects, NRC Region IV, to R. L. Andrews, Division Manager, Nuclear Production, Omaha Public Power District, forwarding NRC Inspection Report No. 50-285/87-30, Docket No. 50-285, December 10, 1987.*
- 3. Letter from Robert D. Martin, Regional Administrator, NRC Region IV, to l
R. L. Andrews, Division Manager, Nuclear Production, Omaha Public Power i District, forwarding a Notice of Violation and Proposed Imposition of Civil Penalty, Docket No. 50-285, February 22, 1988.*
- 4. U.S. Nuclear Regulatory Commission, NUREG-1275, Vol. 2, "Operating Experi-ence Feedback-Report-Air System Problems," December 1987.*
- 5. U.S. Nuclear Regulatory Commission, NRC Information Notice No. 87-28, Supplement 1, "Air System Problems at U.S. Light Water Reactors," Decem-ber 28, 1987.*
- 6. Letter from Hubert J. Miller, Director, Division of Reactor Safety, NRC Region III, to Alvin Kaplan, Vice President Nuclear Group, The Cleveland Electric Illuminating Company, enclosing Augmented Inspection Teau Report No. 50 440/87-24, Docket No. 50-440, January 22, 1988.*
- 7. Letter from Hubert J. Miller, Director, Division of Reactor Safety, NRC Region III, to Alvin Kaplan, Vice President Nuclear Group, The Clevaland Electric Illuminating Company, enclosing Augmented Inspection Team Report No. 50-440/87-27, Docket No. 50-440, February 10, 1988.*
- 8. U.S. Nuclear Regulatory Commission, NRC Information Notice No. 88-43, "Solenoid Valve Problems," June 23, 1988.* j
- 9. Letter from Alan R. Herdt, Chief, Engineering Branch, Division of Reactor J Safety, NRC Region II, to R. P. Mcdonald, Senior Vice President, Alabama )
Power Company, forwarding NRC Inspection Report Nos. 50-348/87-36 and 50-364/87-36, Docket Nos. 50-348 and 50-364, January 29, 1988.* !
I
- 10. U. S. Nuclear Regulatory Commission, NRC Information Notice No. 88-01, "Safety Injection Pipe Failure," January 27, 1988.*
- 11. U.S. Nuclear Regulatory Commission, NRC Bulletin No. 88-08, "Thermal l Stresses in Piping Connected to Reactor Coolant Systems," June 22, 1988.* !
- Available in NRC Public Document Room, 1717 H Street, NW, Washington, D.C.
20555, for public inspection and/or copying.
l 23 l
- 12. Letter from James M. Taylor, NRC Deputy Executive Director for Regional Operations, to Georgia Institute of Technology, forwarding an Order Modi-fying License, Effective Immediately, Docket No. 50-160, January 20, 1988.*
- 13. Letter from J. Nelson Grace, Regional Administrator, NRC Region II, to Dr. T. E. Stelson, Vice President for Research, Georgia Institute of Tehenology, forwarding NRC Inspection Report No. 50-160/87-08, Docket No.
50-160, February 10, 1988.*
- 14. Letter from James M. .aylor, NRC Deputy Executive Director for Regional Operations, to Dr. J. P. Crecine, President, Georgia Institute of Technology, forwarding a Confirmatory Order Modifying License, Effective Immediately, Docket No. 50-160, March 17, 1988.*
- 15. Letter from Hugh L. Thompson, Jr. , Director, NRC Office of Nuclear Material Safety and Safeguards, to Robert G. Wissink, Chairman, Isotope Committee, Minnesota Mining and Manufacturing Company, forwarding an Order Modifying License, Effective Immediately, Docket Nos. 030-04951 and 030-04971, January 25, 1988.*
- 16. Letter from Hugh L. Thompson, Jr. , Director, NRC Office of Nuclear Material Safety and Safeguards, to Robert G. Wissink, Chairman, Isotope Committee, Minnesota Mining and Manufacturing Company, forwarding a Confirmatory Order Modifying License, Effective Immediately, Docket Nos. l 030-04951 and 030-94971, February 5, 1988.*
- 17. Letter from Robert M. Bernero, Deputy Director, NRC Office of Nuclear Material Safety and Safeguards, to Robert G. Wissink, Chairman, Isotope Committee, Minnesota Mining and Manufacturing Company, forwarding an Order Modifying License, Effective Immediately, Docket Nos. 030-04951 and 030-04971, February 12, 1988.*
- 18. Letter from Robert M. Bernero, Deputy Director, NRC Office of Nuclear Material Safety and Safeguards, to Robert G. Wissink, Chairman, Isotope Committee, Minnesota Mining and Manufacturing Company, forwarding (1) an Order Modifying License, Effective Immediately, and Order to Show Cause, and (2) an Order Modifying General License in 10 CFR S 31.5, Docket Nos.
030-04951 and 030-04971, February 18, 1988.*
- 19. Letter from Bruce S. Mallett, Chie , Nuclear Materials Safety and Safeguards Branch, NRC Region III, to Bruce Clement, Vice President of Professional and Support Services, St. Joseph's Hospital, Milwaukee, Wisconsin, License No. 48-00537-04, dated March 10, 1988.*
- 20. Letter from A. Bert Davis, Regional Administrator, NRC Region III, to I Agnar Pytte, President, Case Western Reserve University, enclosing Notice I of Violation and Proposed Imposition of Civil Penalty and Inspection '
Report Nos. 030-00902/87-1, 030-29480/87-01, 030-07040/87-01, and 070-00145/87-01; Docket Nos. 030-00902, 030-07040, 030-29480, and 070-00145; February 26, 1988.*
- Available in NRC Public Document Room, 1717 H Street, NW, Washington, D.C. )
20555, for public inspection and/or copying. ;
24
- 21. Confirmatory Action Letter No. RIII-88-020 from A. Bert Davis, Regional Administrator, NRC Region III, to Arthur P. Leary, Administrative Vice President, Case Western Reserve University, License No. 34-00738-04, November 20, 1987.* i
- 22. Confirmatory Action Letter No. RIII-88-023 from A. Bert Davis, Regional Administrator, NRC Region III,-to Arthur P. Leary, Administrative Vice President, Case Western Reserve University, License No. 34-00738-04, November 25, 1987.* !
I I
l l
l
- Available in NRC Public Document Room, 1717 H Street, NW, Washington, D.C.
20555, for public inspection and/or copying.
25
APPENDIX A ABNORMAL OCCURRENCE CRITERIA The following criteria for this report's abnormal occurrence determinations I were set forth in an NRC policy staterr.ent published in the Federal Register on February 24, 1977 (Vol. 42, No. 37, pages 10950-10952).
An event will be considered an abnormal occurrence if it involves a major reduction in the degree of protection of the public health or safety. Such an event would involve a moderate or more severe impact on the public health or safety and could include but need not be limited to:
- 1. Moderate exposure to, or release of, radioactive material licensed by or otherwise regulated by the Commission; l 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 cri-teria are:
For All Licensees
- 1. Exposure of the whole body of any individual to 25 rems or more of radia-tion; exposure of the skin of the whole body of any individual to 150 rems or more of radiation; or exposure of the feet, ankles, hands or forearms of any individual to 375 rems or more of radiation (10 CFR 920.403(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 620.105(a)).
- 3. The release of radioactive material to an unrestricted area in concentra-tions which, if averaged over a period of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, exceed 500 times the regulatory limit of Appendix B, Table II, 10 CFR Part 20 (10 CFR 620.403(b)).
- 4. Radiation or contamination levels in excess of design values on packages, or loss of confinement of radioactive material such as (a) a radiation dose rate of 1,000 mrem per hour three feet from the surface of a package containing the radioactive material, or (b) release of radioactive material from a package in amounts greater than the regulatory limit.
- 5. Any loss of licensed material in such quantities and under such circum-stances that substantial hazard may result to persons in unrestricted areas.
- 6. A substantiated case of actual or attempted theft or diversion of licensed material or sabotage of a facility.
27
- 7. Any substantiated loss of special nuclear material or any substantiated inventory discrepancy which is judged to be significant relative _ to normally expected performance and which is judged to be caused by theft or diversion or by substantial breakdown of the accountability system.
- 8. Any substantial breakdown of physical security or material control (i.e.,
access control, containment, or accountability systems) that significantly-weakened the protection against theft, diversion, or sabotage.
- 9. An accidental criticality (10 CFR 670.52(a)).
- 10. A major deficiency in design, construction, or operation having safety implications requiring immediate remedial ~ action.
- 11. Serious deficiency in management or procedural controls in major areas.
- 12. Series of events (where individual events are not of major importance),
recurring incidents and incidents with implications for similar facilities (generic incidents), which create major safety concern.
For Commercial ~ Nuclear Power Plants
- 1. Exceeding a safety limit of license technical specifications (10 CFR 650.36(c)).
- 2. Major degradation of fuel integrity, primary coolant pressure boundary, or primary containment boundary.
- 3. Loss of plant capability to perform essential safety functions such that a potential release of radioactivity in excess of 10 CFR Part 100 guidelines could result from a postulated transient or accident (e.g., loss of emer-gency core cooling system, loss of control rod system).
- 4. Discovery of a major condition not specifically considered in the safety analysis report (SAR) or technical specifications that requires immediate remedial action.
- 5. Personnel error or procedural deficiencies which result in loss of plant capability to perform essential safety functions such that a potential release of radioactivity in excess of 10 CFR Part 100 guidelines could result from a postulated transient or accident (e.g., loss of emergency core cooling system, loss of control rod system).
For Fuel Cycle Licensees
- 1. A safety limit of license technical specifications is exceed 2d and a plant shutdown is required (10 CFR 950.36(c)).
- 2. A major condition not specifically considered in the safety analysis re-port or technical specifications that requires immediate remedial action.
- 3. An event which seriously compromised the ability of a confinement system to perform its designated function.
28
APPENDIX B UPDATE OF PREVIOUSLY REPORTED ABNORMAL OCCURRENCES During the January through March 1988 period, the NRC, NRC licensees, Agreement States, Agreement State Licensees, and other involved parties, such as reactor vendors and architects and engineers, continued with the implementation of actions necessary to prevent recurrence of previously reported abnormal occur-rences. The referenced Congressional abnormal occurrence reports below provide the initial and any updating information on the abnormal occurrences discussed.
The updating provided generally covers events which took place during the report period, thus some information is not current. Some updating, however, l
is more current as indicated by the associated event dates. Open items will be i discussed in subsequent reports in the series.
l NUCLEAR POWER PLANTS 79-3 Nuclear Accident at Three Mile Island This abnormal occurrence was originally reported in NUREG-0090, Vol 2, No. 1, "Report to Congress on Abnormal Occurrences: January-March 1979," and updated in each subsequent report in this series, i.e., NUREG-0090, Vol. 2, No. 2 through Vol. 10, No. 4. It is planned to continue these updates until defueling activities at the site are completed. The update of activities for the period of December 1, 1987 through February 29, 1988 (except wSere otherwise noted) is as follows:
Reactor Building Activities During the December 1987 through February 1988 period, 85 c,tries were made into the TMI-2 reactor building, bringing the total number of entries since the March 1979 accident to 1599. Reactor building activities de:lng this period centered on the continuing defueling operation, including data acquisition, video inspection, bulk defueling, debris vacuuming and 'emoval of standing partial, fuel assemblies. Hydrolazer scarifying (removal of a shallow surface layer of concrete with a high pressure water jet) of contaminated surfaces in the reactor building basement has continued. The technique has been successful in reducing dose rates from 300 rem /hr to 30 rem /hr on some concrete block surfaces.
Reactor Vessel and Ex-Vessel Defueling Operations During the December 1987 through February 1988 period, about 1,564 pounds of debris were removed from the reactor vessel, generally by pick and place proc-esses and by the use of the air lift tool.
At the close of the reporting period, 175 of 177 fuel assemblies had been re-moved and loaded into defueling canisters. The total mass loaded into canis-ters was approximately 193,000 lbs (64 percent) out of a total of approximately 300,000 lbs of core debris and other materials. The total mass to be removed includes the mass of the core, structural and absorber materials, mass added by oxidation of core and structural material, and portions of the baffle plates, 1
l 29 l l
i i
formers, and other components that will become comingled with core debris dur-ing cutting operations. Core region defueling has been completed and the i licensee is preparing to begin efforts to remove fuel which was relocated in i the lower core support assembly and the lower reactor vessel head. The video examination of the accessible areas of the lower core support structure indi-cates that there is less fuel debris in the core former region of the reactor vessel than was previously estimated. Source range neutron monitors indicate I that about 28 metric tons of core debris are in the lower reactor vessel head, an increase from previous estimates which is believed attributable to core i debris relocation during core region defueling. !
Ex-vessel defueling was conducted in the "A" and "B" reactor coolant system hot ;
legs and the decay heat drop line. During the vacuum defueling it was determined !
that the decay heat drop line was plugged by a relatively large mass of material. !
Both the "B" hot leg and decay heat drop line, as well as the pressurizer are i expected to be defueled further. j i
Cask and Liner Shipments 1 Shipments of THI-2 core debris to INEL have continued. Six shipping cask loads ,
of seven defueling canisters each were transferred by rail. As of February 29, l 1988, 175 canisters have been sent to INEL. Approximately 185,700 pounds of debris have been shipped (62 percent of the total expected). Three EFICOR II liners, six high integrity containers of dry activated waste, and an SDS liner filter were also shipped offsite.
EPICOR II/ Submerged Demineralizer System (SDS) Processing Through the end of the reporting period, a total of 4,162,181 gallons of water had been processed through the EPICOR II system, and a total of 4,505,022 gal-lons had been processed through the SDS. During the reporting period, approxi-mately 256,400 gallons were processed by the EPICOR II system. SDS remained shut down throughout the period.
N Auxiliary and Fuel Handling Building (AFHB) Activities Attempts to move resins from the Makeup and Purification Demineralizers have been partially successful as indicated by decreasing radiation levels in the "A" AFHB cubicle and increasing levels in the spent resin storage tank. Work is expected to continue to remove all the resins and package them for shipment off the TMI site. To date 118 of the 143 AFHB cubicles hav0 been decontaminated.
Post-Defueling Monitored Storage The NRC continues to evaluate the licensee's plans for Post Defueling Monitored Storage and expects to issue a draft environmental statement during the next reporting period.
Proposal to Dispose of Accident-Generated Water The Atomic Safety and Licensing Board Panel, estabibbed to preside over the proceedings on disposition of the accident generated water, held a pre-hearing 30
conference in Harrisburg, Pennsylvania on December 8, 1987. At the pre-hearing conference, which was open to the public, the parties to the proceedings [i.e.,
the licensee, Susquehanna Valley Alliance (SVA), Three Mile Island Alert (TMIA),
the Commonwealth of Pennsylvania, and the NRC staff] discussed proposed con-tentions and conduct of the hearing. On January 6, 1988, the Licensing Board ruled on t5e contentions and established a discovery period ending February 22, 1988. Subsejuntly the Board, at the request of SVA, extended the discovery period to 'M o? March 29, 1988.
TMI-2 Advisory Panel Meeting
- The Advisory Panel for the Decontamination of Three Mile Island Unit 2 met on l January 13, 1988 in Harrisburg, Pennsylvania. The Panel heard additional de-tails of the planned dis-establishment of the NRC TMI-2 Cleanup Project Direc-torate (which became effective on January 31, 1988). The Panel also heard an update on the status of the cleanup by the licensee. Additionally, the Panel was informed of the status of the transfer of radiological monitoring functions to the Pennsylvania State University Hershey Medical Center.
Future reports will be made as appropriate.
A A A A A A A A 81-7 Blockage of Coolant Flow to Safety-Related Systems and Componerts This abnormal occurrence was originally reported in NUREG-0090, Vol. 4, No. 4, "Report to Congress on Abnormal Occurrences: October-December 1981." It was updated, and closed out, in NUREG-0090, Vol. 7, No.1, "Raport to Congress on Abnormal Occurrences: January-March 1984" and in NUREG-0090, Vol. 10, No. 2, "Report to Congress on Abnormal Occurrences: April-June 1987." It is being reopened to report a similar, significant event which occurred at Catawba Unit 2 on March 9, 1988. Catawba Unit 2 is a Westinghouse-designed pressur-ized water reactor, operated by Duke Power Company (the licensee), and located in York County, South Carolina.
O reported previously, the Service Water Systems (SWSs) of nuclear power plants are typically open cycle systems. An "open-cycle SWS" pumps water directly from a river, cooling pond, or ocean body, into the service water intake structure.
In addition to water, me2, silt, sand, algae, bacteria, fungi and aquatic l organisms may also be pamped into the intake structure. Although gratings, screens, and' filters block out many of the impurities, fouling of SWSs still occurs.
Safety-related SWSs, which already have separate and redundant piping systems, l share the same intake structure and ultimate heat sink. Thus, they share a potential for common mode failure due to service water impurities. To deal with tnis concern, the NRC staff developed Generic Issue 51, ' Improving Relk l bility of Open Cycle Service Water Systems." In addition, the NRC has issued l Informa 1 Notice Nos. 81-21 (Ref. B-1) and 86-96 (Ref. B-2) and Bulletin No.
81-03 (hu . B-3), all of which address fouling of heat exchangers or piping.
31
On March 9, 1988, Catawba Unit 2 scrammed from 20% power while in the process of starting up from the unit's first refueling. Steam generator (S/G) level fluctuations occurred due to a main feedwater regulating valve failing open while being placed in automatic which caused a feedwater pump trip and a feedwater isolation. Upon loss of the main feedw aer pumps, both motor driven auxiliary feedwater (AFW) pumps auto-started as designed; however, a momentary "A" train AFW pump low suction pressure resulted in the transfer of pump suc-tion from its normal source of condensate grada water to the Raw Service Water System [also known at Catawba as the Nuclear Service Water (NSW) System] wnich serves as the assured source to AFW. Approximately 13 minutes into the event, the operator noted that the service water valve to AFW had opened. As a result, NSW had been pumped into the A and B S/Gs.
Actions were initiated to determine the cause(s) of the event. Upon disassem-bly of the A and B S/G AFW flow control valves, it was found that the valves were clogged with Asiatic c',am shells. It was concluded that continued safe operation of Unit 1, which was at full power, could not be assured due to the possibility of clam fouling; therefore, the licensee placed Unit 1 in hot shut-down. On March 10, 1988, the NRC sent an Augmented Inspection Team (AIT) to Catawba to review the circumstances associated with the event. A Confirmation of Action Letter was issued on March 11, 1988 (Ref. B-4).
The AFW flow control valves at Cat _ s have an anti-cavitation trim which could not pass all the particles of clam shells and caused the flow degradation.
The ' licensee has in place several programs and practices desigi.ed to verify NSW flow to verious systems and components. This includes: periodic flaw balancing of NSW; testing to verify heat transfer capability of essential heat exchangers servfced by NSW; and periodic cleaking of heat exchangers based on differential pressure indications and examination of these heat exchangers and related piping during cleaning for the presence of clams and unusual fouling conditions. The inspection cf the NSW system dead leg piping for clams had consisted of spot radiographic (RT) inspection. None of the RT inspections had revealei any clams. The valves in the NSW/AFW suction swapover lines had b;:<n periodically stroke tested, but ne fl~ through the lines m established.
As a result of the Unit 2 AFW swapove, fre .ondensate to NSW and the introdue" tion of raw NSW and clams into the AFR system, the licensee initiated a program of flushes and inspection of relatively stagnant portions of piping between the NSW system and various safety-related components. The NSW system flusEn and inspections were performed on: NSW backup to AFW system piping, NSW backup to ~ 1se injection A ,'ent cooling system Flushing piping, was also and NSW planned backup on NSW backuptotopenetration vs spent fuel pool cooling pip m .
The itcensee confirmed Unit l's operability and was given NRC concurrence for restart on March 11, 1988. Unit 2 required more extensive corrective actions and testing. The test results were reviewed by the NRC, and concurrence to restart was given on March 18, 1988.
On April 20, 1988, the AIT findings were issued in Ir gection Report Nos.
50-413/88-14 W '0-414/88-14 (9ef. B-5).
32 l
l l On June 14, 1988, NRC Information Notice No;88-037, "Flow Blockage of Cooling L
Water to Safety System Components" was issued to alert licensees to a potentially generic problem involving flow blockage in safety-related piping interconnec-tions due to biofouling (Ref. B-6). ,
Progress on resolution of Generic Issue 51 is reported biannually in NUREG-0933,
",A Prioritization of Generic Safety Issues" (Ref. B-7).
This item is considereo closed for the purpose of this report.
- A A A A 85-14 Managuent Deficencies at Tennessee Valley Authority This abnormal occurrence was originally reported in NUREG-0090, Vol. 8, No. 3,
Report to Congress on Abnormal Occurrences: July - September 1985," and updated in Vol. 9. No. 1; Vol. 9, No. 2; Vol. 9, No. 3; Vol. 10, No. 2; and Vol. 10, No. 4. It is further updated for this report periou as follows:
Overview of Sequoyah Issues Since August, 1985, Tennessee Valley Authority (TVA) and the NRC have worked to resolve the issues that must be addressed before the restart of Sequoyah
. Unit 2. The NRC staff has reviewed and approved TVA's program to resolve the management issues required for restart of Unit 2. These issues are addressed in the NRC's NUREG-1232, Volume 1, "Safety Evaluation Report (SER) on Tennessee Valley Authority Revised Corporate Nuclear Performence Plan,"
published July 1987 (Ref. B-8), and in draft NUREG-1232, Volume 2, "Safety Evaluation Report (SER) on Tennessee Valley Authority Sequoyah Nuclear Performance Plan" (Ref. B-9). In the latter document, the staff discGsses the resolution of the technical concerns that were identified in Volume 2 of TVA's Nuclear Performance Plan (Sequoyah).
Other significant issues that TVA ns required to address included the Inte-grated Design Inspection (IDI), and electrical issues including the adequacy of silicone rubber-insulated cables and the adequacy of emergency diesel generators. These issues are discussed below.
Sequoyah Uait 2 Restart TVA stai'ted non-nuclear heatup of Unit 2 on February 6, 1988. The heatup to normal operating temperature was scheduled for a five to six week period. The puroose of the heatup period was to allow TVA to conduct the necessary equipment tests, tn c'emonstrate plant operability, and to provide the Sequoyeh operators ex.norience in close-to-normal plant operations. TVA then requested NRC approval for plant restart, subject to closure of all outstanding safety conc'rns.e On March 22, 1988, the Commission authorized the staff to permit TVA to restart Unit 2. Subsequently, on Har;h 30, 1988, the NRC staff authorized TVA to re-start Unit 2.
Uri: ' startup was interruoted on April 6, 1988 when TVA identified indication of 'eam generator (S/G) t'ube leakage. Unit 2 was cooled down and helium leEage and eddy current testing identified a crack in the U-bend of a tube in 33
the innermost row as the cause of the leakage. The cause of the cracking was the residual stress remaining from the cold forming of the U-bends in the tight radius inner tubes. TVA has elected to plug all Row I tubes in all four S/Gs, When this work and correction of problems with the primary safety valve line are completed, Unit 2 will resume its startup. This resumption is presently scheduled for the second quarter CY 1988.
The NRC staff has closely monitored the non-nuclear hestup phase of the Sequoyah restart program through an augmented inspection effort E a restart test plan that includes an onsite manager and 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> coverage of piant operations. This restart test plan includes special inspections of plant operations, evaluations of TVA's analysis of events, and evaluation of plant hardware that will lead to NRC decisions regarding whether or not to allow continued plant operation past five pre-estab1f shed hold points.
Integrated Design Inspection A description of the purpose and scope of the Integrated Design Inspection (IDI) was provided in a previous report in this series (i.e., NUREG No. 0090, Vol.10, No. 2, "Report to Congress on Abnormal Occurrences: April-June 1987").
The NRC issued a letter t3 TVA on October 9, 1987 (Ref. B-10) that defined the IDI findings that had to be addressed before restart. The full IDI report was issued November 6,1987, and identified addition? t ^ start items (Ref. B-11). l In total, the staff required that 64 of the IDI inspection findings be resolved l prior to startup of Sequoyah Unit 2, and 21 additional items were required to be addressed by TVA after restert.
OveraII, the IDI inspection team concluded that despite some identified defi-ciencies of the TVA design process at Sequoyah, the IDI did not uncover any systematic or re:urring failures that would be indicative of unresolved pro-grammatic design problems.
The NRC has conducted additional followup inspections to review additional information provided by TVA and to verify correction of identified deficiencies.
Correction of all restart required items has been completed and verified by the NRC staff.
Silicone Rubber-Insulated Cable The NRC staff investigated the adequacy of installed elee.trical cable t'ased on employee concerns regarding allegedly improper installation practices at Sequoyah. Following extensive testing of installed cable, the staff concluded that cable was not damaged due to improper installation. The test program revealed unrelated problems with silicone rubber-insulated cables, and TVA conducted additional testing of thesa cables.
Following the testinc program, TVA replaced all cables at Unit 2 that were made by one specific manufacturer and that had the most significant failures l during testing. The NRC staff has accepted additional testing at Wyle i Laboratories that demonstrated the ability of cable with severely reduced I cable insulation thickness to function during and after a LOCA as adequate to l show the suitability of the remaining installed silicone rubber-insulated.
cables for 10 years of cervice. Additional testing to certify this cable for ,
the remainder of plant life will be conducted after tha restart of Unit 2.
34 9
Diesel G6nerator Operability As part of its Sequoyah corrective action program, TVA made major changes in the emergency diesel generator (EDG) load sequencing and timing. TVA load-tested the EDGs between July and November 1987. In January 1988, TVA reported that data from these surveillance tests raised significant questions about the operability of the EDGs and their conformance to voltage limits during loading.
A review of the test data by the NRC identified additional concerns with EDG transient voltage performance. After additional review by both NRC and TVA, it was concluded that, although current diesel generator performance was ade-quate for operation, enhancements should be provided after restart to increase performance margins.
TVA has committed to undertake these corrective actions, probably involving modification or replacement of major pnrtions of the exciter / regulator system, during the first refueling outage following Unit 1 restart. This will probably occur in late 1990. This schedule is acceptable to the NRC staff.
Future reports will be made as appropriate.
A A A A A A A A FUEL CYCLE FACILITIES 86-3 Rupture of a Uranium Hexafluoride Cylinder and Release of Gases This abnormal occurrence, involving Sequoyah Fuels Corporation (SFC), Gore, Oklahoma, was originally reported in NUREG-0090, Vol. 9, No.1, "Report to Congress on Abnormal Occurrences: January-March 1986," and updated in subse-quent reports in this series, i.e., NUREG-0090, Vol. 9, Nos. 2, 3, and 4, and Vol. 10, No. 3. It is further updated, and closed out, as follows:
I As discussed in the previous update report, on September 1, 1987, the NRC issued an Order to Show Cause and Notice of Violation and Proposed Imposition of Civil Penalty (Ref. B-12). The Order addressed the matte
- of Independent Oversight Team (IOT) coverage and concluded that the coverage could be reduced from the full coverage of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> per day to one random shift per day. In addition, the order concluded that several supervisors were aware of improper practices at the facility and did not fully disclose their knowledge of these practices to NRC investigators. The Order asked why SFC should be allowed to continue to operate while these supervisors are permitted to conduct licensed activities.
The Notice of Violation and Proposed Imposition of Civil Penalty addressed l an apparent false statement made by the licensee in a letter dated January 29, 1986. The NRC proposed a civil penalty of $8,000 for this apparent material false statement. ,
On Septeniber 25, 1987, the licensee responded to the NRC concerning the Notice of Violation, denying the violation (Ref. B-13). On October 1, 1987, the licensee responded to the NRC concerning the Order to Show Cause (Ref. B-14) describing changes and improvements made, the behavior of the four named super-j visors, and requesting that the NRC vacate the Order to Show Cause. ;
35 ;
l
On October 19, 1987, theNRCrespondedtothelicensee'sletters, requesting additional information regarding the supervisors and regarding the licensee s procedural and protocols for assuring that information provided to the NRC is complete and accurate (Ref. B-15). The licensee submitted the further infor-mation on November 13, 1987.
Based on the numerous inspections and investigations conducted by the NRC fol-lowing the January 4, 1986 accident, the responses received to the NRC Orders, the performance of the 10T, interviews with licensee workers, and meetings with licensee management, the NRC has determined that there is now a clee.rly improved attitude in the management and operation of Sequoyah Fuels Facility, emphasizing safe operation, following of procedures, and attention to detail.
This attitude, the expanded training program, and changes in organizational structure all reflect strong management control and improved interaction with corporate management. The 10T expressed positive views with regard to operat-ing procedures, managenent' controls and oversight.
The 10T also expressed positive views as to the competency, attitude, and candor of the four super-visors subject to the September 1,1987 Order.
Therefore, since the NRC is now satisfied that there is reasonable assurance that activities will be conducted properly, on February 10, 1988, the NRC sent a letter to the licensee which eliminated the 10T requirement (Ref. B-16).
Inspections will ensure that strong management oversight will continue. The requirement for all written submissions to be under oath of the President, SFC, was also removed and the Order to Show Cause as to the named supervisors was rescinded.
However, to emphasize the importance of providing complete and accurate infor-mation to the NRC, the February 10, 1988 NRC lett.er also included an Order imposing a civil monetary penalty, in the amount "f $8,000, for the previously mentioned material false statement. By letter dated March 1, 1988, the licensee enclosed a check for the imposed civil penalty, as committed to in a Settlement Agreement submitted on February 26, 1988.
This item is considered closed for the purposes of this report.
1 36
APPENDIX C OTHER EVENTS OF INTEREST The following item is described below because it may poss1 Sly be perceived by l the public to be of public health significance. The item did not involve a i
major reduction in the level of protection provided for public health or safety; therefore, it is not reportable as an abnormal occurrence.
l 1. Failure of Two Sets of Redundant Primary Containment Isolation Valves at l
Brunswick Unit 2 On January 2, 1988, Brunswick Unit 2 experienced a failure of four primary I
containment isolation valves which was outside of the design basis of the plant and outside previously analyzed conditions. The simultaneous failure of these redundant isolation valves could have, under certain accident ccnditions, re-suited in a direct, unmonitored, and unisolable radioactive release path to the environment. The event was reviewed by an dRC Augmented Inspection Team (AIT).
Brunswick Unit 2 is a General Electric-designed boiling water reactor, located in Brunswick County, North Carolina, and operated by the Carolina Power & Light Company (the licansee).
The event involved the failure of both of the redundant primary containment isolation valves for Unit 2 drywell equipment drains, and both of the redundant primary containment isolation valves for Unit 2 drywell floor drains, to auto-matica11y closa after receipt of a valid isolation signal. Both of the equip-mont drain primary containment isolation valves aiso failed ta close on a sub-sequent manual signal.
The containment isolation valves for the equipment drains and floor drains are part of the primary containment isolation system (PCIS), a safety-related sys-tem designed to prevent or limit the release of radioactive materials that may result from accidents through rapid, automatic closure of valvec in process lines which penetrata the primary contair. ment. Radioactive water which drains ;
from systems within the drywell or the drywell floor is collected and then '
pumped, via two separate pipes which penetrate the primary containment (drywell),
to the radwaste collector.
The two lines communicate directly with the drywell atmosphere, but not the reactor vessel, and hence they are considered class "B" penetrations requiring two automatic isolation valves in series outside containment. The drywell floor drain PCIS valve numbers are 2 G1ti-F003 (inboard) and 2-G16-F004 (out-board); the equipment drain PCIS valve numbers are 2-G16-F010 (inboard) and 2-G16-F020 (outboard). These four drywell drain PCIS valvcs receive an auto-matic closure signal when the drywell pressure exceeds tlte Technical Sp'ecifi-cation set point of 2 psig or when the reactor v9ssel low water level 1" goes below 162.5 inches.
The four PCIS valves are 3", 350 lb. gate valves. The valves are opened by an air actuation cylinder, controlled by solenoid valves, and are spring closed upon loss of air p) essure.
37 k
- l i
On the evening of January 1, 1988, Unit 2 was at 69 percent power and prepara- l tions were being made for a shutdown for a refueling outage due to fuel deple- l tion. At approximately 8:15 p.m. , the operators began an orderly power reduc- i tion utilizing a combination of recirculation flow reduction and control rod insertion. However, problems were encountered in maintaining condenser vacuum; therefore, the shift operations supervisor (505) made a decision to scram the- !
reactor and directed a further reduction in recirculation flow. At 12:18 a.m.,
the SOS directed a manual scram of the reactor. As a result of the scr g Unit 2 reactor vessel water level decreased below the low level "1" setpoint, which sent an automatic closure signal tr the four drvwell drain PCIS valves.
Following the scram, the actina sMft foreman (ASF) began directing the opera-tors through the Emergency Operating Procedure (EOP). In doing so, the control Oerator(CO)noted on the labeled control board, that the four drywell drain FJISvalveswerestIllinthe"0 PEN"positionwiththeredindicatinglights ,
illuminated. The C0 immediately attempted to manually close these PCIS valves )
utilizing the control board manual switches by moving these switches to the i "CLOSE" position. Upon receiving the manual close signal, valves F003 and F004 l closed as indicated by the light change from red to green. Despite efforts by l the C0 and later by the Shift Foreman (SF) from Unit 1, valves F019 and F020 :
could not be closed. l Eventually, at approximately 12:24 a.m., the F020 valve was observed to be in the "CLOSED" position. This closure provided a single isolation of the breach in primary containment that had existed for about six minutes. Another unsuc- j cessful attempt was made to close valve F019. Eventually, this valve was i observed "CLOSED" at approximately 12:26 a.m., which completed closure of the four PCIS valves. Initial investigation by the licensee indicated that the solenoid valves which actuate the containment isolating gate valves had failed.
For the January 2, 1988 event, isolation of the primary containment occurred at 12:24 a.m., about six minutes after the low water level in the reactor vessel generated a closure signal for the PCIS valves. However, it could not be assumed that either automatic closure, or manual closure from the control room, would have occurred during a design basis loss of coolant accident. In addi-tion, a 1000 rem radiation field in the area of the four PCIS valves would have impeded local closure attempts.
The NRC AIT concluded that the failure of the PCIS valves was probably due to multiple and diverse causes such as solenoid sticking and relay contact failures.
It was also noted that there was a weakness in the licensee's maintenance trend-
! ing program which resulted in the failure of the licensee to identify an adverse
- trend concerning past solenoid valve failures.
During the Unit 2 event, Unit 1 was operating at 100% power. To determine the operability of the four similar valves on Unit 1, on January 2, 1988, the licen-see manually cycled the valves successfully from the control room. In addition, nn January 4, 1988 the licensee conducted a special test on Unit 1 to verify the wiring logic for tha valves. After this verification, the licensee, with NPC AIT concurrence, closed the valves on Unit 1 and stated they would remain closed during plant operation, except when they were required to be opened dur-ing pumping of the floor and equipment drains.
i 38 i
. The licensee initiated a Scram Incident Investigation Team (SIIT) to investigate
- the event. This included performing operability tests on the four valves, dis-l assembly and inspection of the valves, inspection of the instrument air systems and inspection of PCIS relays. The licensee indicated that the investigation of the solenoid valve and electrical circuitry failures would continue and that the NRC would be promptly advised of any new information. The' licensee com-mitted to provide training to clarify the proper operation of the hand switches for the four failed valves. They also agreed to continue action to review and revise as necessary the maintenance trending program to ensure that specific failures are identified, and to revise the program to ensure that repetitive j failures are escalated to management attention if warranted.
l The NRC sent a fact-finding AIT to the site on January 5, 1988 to review the circumstances associated with the event and the actions taken by the licensee.
The AIT concluded that although the event d'id not result in a radiological re-lease or in exceeding any of the license technical specification safety limits, it has significant implications and is potentially generic to other operating facilities with similar valves and circuitry. The NRC plans a follow-up review after the licensee has completed its investigation of the incident.
The AIT report (Inspection Report Nos. 50-325/88-03 and 50-324/88-03) was for-warded to the licensee on January 27, 1988 (Ref. C-1). Action has also been undertaken to address the potential generic concerns regarding sticking ,
solenoid valves at Brunswick and at other nuclear power stations; '
i
- 2. Rapid Power Oscillations Leading to Automatic Reactor Scram at LaSalle Unit 2 On March 9, 1988, with LaSalle Unit 2 at about 85 percent power, an error by 1 instrument maintenance (IM) personnel led to the trip of both recirculation l pumps. Approximately five minutes later the reactor neutron power level began l
to oscillate widely for about two minutes until the upper level reached the l automatic shutdown point for the reactor at 118 percent power. The reactor then automatically scrammed. All systems functioned normally during the reactor shutdown.
l LaSalle Unit 2 is a General Eiectric-designed boiling water reactor, operated l l by Commonwealth Edison Company, and located in LaSalle County, Illinois. l l
l An NRC Augmented Inspection Team (AIT) performed a special inspection at the site from March 16-24, 1988, to evaluate the circumstances associated with the power level oscillations. The phenomenon had occurred previoutly at some for-eign BWRs and had been duplicated during testing at a domestic BWR. The LaSalle incident, however, was the first occurrence of the oscillation phenomena during normal operation of a domestic BWR.
The incident was initiated when IM personnel were in the process of performing an instrument surveillance. A valving error produced a pressure pulse which actur.ed the instrumentation which causes a trip of both necirculation pumps designed to decrease power in the event of an Anticipated Transient Without Scram (ATWS). These pumps control power level in the reactor by varying the flow rate of cooling water past the reactor fuel--the faster the water passes the fuel, the higher the power level.
39
The tripping of the pumps resulted in an immediate drop in power level as the flow rate decreased. Shortly thereafter the steam flow into the feedwater heaters started to isolate to prevent water from backi.y up into the turbine.
(The feedwater system pumps water from the condenser into the reactor. Feed-water is normally heated before entering the reactor to increase the unit's efficiency.) Without the feedwater heaters, cooler water was pumped into the reactor. Cooler water tends to increase the reactor power level.
At this point, the reactor power level was at a point of recognized instabil-ity (i.e., the control rods were withdrawn from the reactor core in their full power positions and the flow rate of water through the reactor was decreasing).
However, the reactor core was fully covered and the fuel rods were being ade-quately cooled by the feedwater flow into the reactor.
Under the conditions present, the reactor power level (measured by the produc-tion of neutrons) began to oscillate. The oscillations were measured from the 20 percent power level to the 95 percent level on a data recording system, l although the control room instrumentation, indicating average power levels, showed lesser swings. The oscillations occurred every 2 to 3 seconds and con-tinued to increase for about two minutes until the reactor shut down automat- l ically (scrammed) because of high power level. The reactor trip point is at 118 percent power (measured by neutron production)-
The power swings were too rapid to affect heat production significantly; therefore, the heat output remained at about 45 percent of thermal power.
Reactor operators were aware of the conditions occurring in the reactor, but their operating procedures were unclear on what actions should be taken. They knew that they needed either to increase the water flow rate or to decrease power level of the reactor. They tried unsuccessfully to restart the tecirc-ulation pumps to increase the flow rate. At the time the automatic shutdown occorred, reactor operators had decided to manually scram the plant.
The principal concern associated with the instability and power oscillations is that fuel damage could result if the power oscillations affected one part of the reactor more than others and caused localized overheating of the fuel.
The power instability could also affect the accuracy of reactor instrumentation which might be needed to monitor reactor conditions or initiate safety systems.
No fuel damage was expected or detected through periodic sampling of the reactor cooling system.
On March 17, 1988, NRC Region III issued a Confirmatory Action Letter to the licensee to document the licensee's agreement not to restart the plant until authorized by the NRC (Ref. C-2). After the initial review of the incident by the AIT, the unit was permitted to resume operation on March 18, 1988. An operating procedure change was put in place requiring that the reactor be scrammed manually in the event of a trip of both of the recirculation pumps, The NRC is currently evaluating this inciaent to deterrine if additional changes are needed in operating procedures, instrumentation, and Technical Specifications, as well as an assessment of the effects of this incident on other operating BWRs. This review will also include the procedures to be fol-lowed in the event of an Anticipated Transient Without Scram (ATWS), a situation
! 40 l
in which a reactor problem occurs but the control rods fail to insert into the reactor to shut the plant down. The current ATWS procedures for bailing water reactors call for tripping of the recirculation pumps to reduce the reactor power le 'el.
The AIT tvaluation of the LaSalle incident, documented in Inspection Report l
Nos. 50-373/88-8 and 50-374/88-8, was forwarded to the licensee on May 16, 1988 I (Ref. C-31 NRC issued Information Notice 88-039, "La Salle Unit 2 Loss of Recircula. ion Pumps with Power Oscillation Event," on June 15, 1988 to alert licensees to potential problems resulting from the thermal hydraulic instability
! of BWR cores when the plant is operating at certain unstable power / flow regions (Ref. C-4). On the same day, NRC issued Bulletin No. 88-07, "Power Oscillations in Boiling Water Reactors (BWRs)," to request that licensees ensure that ade-quate operating procedures and instrumentation are available and adequate oper-ator training is provided to prevent the occurrence of uncontrolled power oscil-lations during all modes of BWR operation (Ref. C-5).
k I
l l
l l
,1
I i
REFERENCES'(FOR APPENDICES) !
t
- B-1 U.S. Nuclear Regulatory. Commission, Inspection and Enforcement Information 'l Notice No. 81-21, "Potential loss of Direct Access to Ultimate Heat Sink," '
July ~21, 1981.* (Provides additional information to Ref. B-3.)
B-2 U.S. Nuclear Regulatory Commission, Inspection and Enforcement Notice !
No. 86-96, "Heat Exchanger Fouling Can Cause Inadequate Operability of j Service Water Systems," November 20, 1986.* l
-B-3 U.S. Nuclear Regulatory Commission,: Inspection an'd Enforcement Bulletin No. 81-03, "Flow Blockage of Cooling Water to Safety Sy' stem Components by 1
- Corbicula sp. (Asiatic clam) and Mytilis sp.-(Mussel), April 10, 1981.*
- i. 'B-4 Confiraation of Action letter, from.J. Nelson Grace, Regional Administrator, !
! NRC Region II, to'H. B. Tucker, Vice_ President, Nuclear Production Depart- ,
ment, Duke Power Company, Docket Nos. 50-413 and 50-414, March 11,' 1988.* l
) B-5 Letter from J. Nelson Grace, Regional Administrator,-NRC__ Region'II, to
- H. B. Tucker, Vice President, Nuclear Production Department, Duke Power !
Company, foivarding Augmented Inspection Team Report Nos. 50-413/88-14 and l j 50-414/88-14, Docket Nos. 50-413/88-14 and 50-414/88-14,-April 20, 1988 * ;
! B-0 U.S. Nuclear Regulatory Commission, NRC Information Notice No. 88-37,,"Flow
) Blocking of Cooling Water to Safety System Components," June 14, 1988. ,
! B-7 U.S. Nuclear Regulatory Commission, "A Prioritization of Generic Safety
! Issues," USNRC Report NUREG-0933, published December 1983,** with Supple- ;
ments issued periodically.** j l
B-8 U.S. Nuclear Regulatory Commission, "Safety Evaluation Report (SER) on j Tennessee Valley Authority Revised Corporate Nuclear Performance Plan," i
~
USNRC Report NUREG-1232, Volume 1, published-July 1987.**
l B-9 U.S. Nuclear Regulatory Commission, "Safety Evaluation Report (SER) on i j Tennessee Valley Authority Sequoyah Nuclear Performance Plan," draft !
j USNRC Report NUREG-1232, Volume 2, to be published.
. [
) B-10 Letter from James G. Keppler, Director, NRC Office' of Special Projects, !
4 to S. A. White, Manager of Nuclear Power, Tennessee Valley Authority, !
F
Subject:
"Items Identified by the Integrated Design Inspection Requiring i J_ Resolution Prior to Restart of Sequoyah Unit 2," Docket Nos. 50-327 and !
- . 50-328, October 9, 1987.*
l l
1 l l *Available in NRC Public Document Room, 1717 H Street, NW, Washington, D.C. I
- 20555, for public inspection and/or copying. !
i I j **Available for purchase from the Superintendent of Documents, U.S. Government !
- Printing Office, P.O. Box 37082 Washington, DC 20013-7082. Also available l l from the National Technical Information Service, SPBS Port Royal Road, ;
i Springfield, VA 22161. A copy is also available for public inspection and/ f or copying at the NRC Public Document Room, 1717 H Street, NW, Washington, D.C. 1 4
43
C-4 U.S. Nuclear Regulatory Commission, NRC Information Notice No. 88-39, "La Salle Unit 2 Loss of Recirculation Pumps with Power Oscillation Event,"
June 15, 1988.*
C-5 U.S. Nuclear Regulatory Commission, NRC Bulletin No. 88-07, "Power Oscilla-tions in Boiling Water Reactors (BWRs)," June 15, 1988.*
l t
i i
a i
l f
]
1 l
I I
l
.f T
A
- Available in NRC Public Document Room, 1717 H Street, NW, Washington, D.C.
20555, for public inspection and/or copying.
3 45 ;
v.=veti m iova, ,co .. .. o.<~vo. ... c .,noc. u ...,,,,
g,oa. >=
%" N! Bl8UOGRAPHIC DATA SHEET NUREG-0090 u s ,~irrvev.o.,i o~ ,. .ivi.n Vole 11. No. 1 3 f at ti .~o VG i s i t. 3ti.Y SL.~E Report to Congress on Abnormal Occurrences January - March 1988 . o.1.. ,0.,co.,s ,io
- , 1. , ..
July [ 1988
. <vi o. ,,,
e o.l . .o.t ISSvio
.O~f. ....
July I 1988
, ,. . . o.. ,~o o. m a . , ,o ~.. .~ o -,u ~o .oo. i u ,,~ <. c , ...wicia....ou v~a vou.
Office for Analysis and Evaluation of Operational Data . 4 o.c..~i~voie.
U.S. Nuclear Regulatory Commission
-Washington, DC 20555
- i. .o~.o.,~o o.a.~ 4.,,o . . .~o o,v ~ .oo.iu ,,-- e c , ii.1,, o, a.o.,
l l Same as 7., above Quarterly _
. . .. e co. . . o ,, .
January - March 1988 13 Sv,,Li .i ~ t .. . ~o t i s i3 088YalcT ,200 eer.s .r ases/
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 sinnificant from the standpoint of public health and safety and requires a Quarterly report of such events to be made to Congress. This report covers the oeriod January 1 to March 31, 1988. Durinn the report period, there were three abnormal occurrences at nuclear power plants licensed to operate: a potential for comon mode failure of safety-related components due to a degraded instrument air system at Fort Calhoun; common mode failures of main steam isolation valves at Perry Unit 1; and a cracked pipe weld in a safety injection system at Farley Unit 2. There were six abnormal occurrences at other NDC licensees; a diagnostic medical misadministration; a breakdown in management controls at the Georgia Institute of Technology reactor facility; release of polonium-210 from static elimination devices manufactured by the 3M Company; two theraceutic nedical misadministrations; and a significant widespread breakdown in the radia' tion safety program at Case Western Reserve University research laboratories. There was one abnormal occurrence reported by an Agreement State (Texas) involving radiation injury to two radiographers.
.. ooco.. ~, .~ .r o . . . ..o oroi .c..,1o..
s .,,, .,,gi .
Degraded Instrument Air System; Main Steam Isolation Valve Failure; Safety Injection System Pipe Weld Crackino; Medical Misadministrations; Management Control Breakdown; Release of Polonium-210; Breakdown in Unlimited .
Radiation Safety Program; Radiation Injury to Radiographers.
.,ei~,..ii.ro. ~ ~oioti.., (IE'cTassified VIcYssified 17 ~w.ii.O. ,aGi.
i i .. .c .
UNITED STATES s, mat rou . cc.ss me NUCLEATj RE EULATCRY COMMISSITN -
NS'"6js*J"' ' **
WASHINGTON, D.C. 20555 <
,, ,,,, % w OFFICIAL BUSINESS d
I PENALTY FOR PRIVATE USE. 6300 '
120555139217 1 1AN1001CV1CY1 l US NRC-0AFM-ADM @
I DIV FOIA & PUBLICATIONS SVCS
^RES-PDR t
NUREG P-210 WASHINGTON DC 20555 C l
4 1
5
. ,j O
i l
l l <
o L.
.s 0