ML20236A862
| ML20236A862 | |
| Person / Time | |
|---|---|
| Issue date: | 05/12/1987 |
| From: | Jordan E NRC OFFICE FOR ANALYSIS & EVALUATION OF OPERATIONAL DATA (AEOD) |
| To: | Beckjord E, Murley T, Thompson H NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS), Office of Nuclear Reactor Regulation, NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES) |
| Shared Package | |
| ML20236A814 | List: |
| References | |
| FOIA-87-377, RTR-NUREG-0090, RTR-NUREG-90 NUDOCS 8710230199 | |
| Download: ML20236A862 (68) | |
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4 o UNITED STATES g.
o NUCLEAR REGULATORY COMMISSION I ,> -[t g, y -
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MEMORANDUM FOR: T. E. Murley, Director, NRR H. L. Thompson, Director NMSS E. S. Beckjord, Director, RES L-
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H. R. Denton, Director, OGPA W. G. Mcdonald, Director, ARM '
,< J. G. Keopler, Director, OSP i
, W. C. Parler, General Counsel Regional Administrators FROM: E. L. Jordan, Director Office for Analysis and Evaluation of Operational Data i
SUBJECT:
ABNORMAL OCCURRENCE REPORT TO CONGRESS FOR l FOURTH QUARTER CY 1986 i i l
l Based on staff response to the AE00 January 30,'1987 memorandum to the Office Directors and Regional Administrators on this subject, we have prepared the enclosed draft Commission Paper (Enclosure 1), the letters of transmittal to Congress (Enclosure 2), the Fourth Quarter CY 1986 Abnormal Occurrence (AO)
Report to Congress (Enclosure 3), and "Other Events Considered for A0 !
Reporting" (Enclosure 4). '
The Enclosures show that there are three proposed A0s at the nuclear power J plants licensed to operate, and six at the other NRC licensees; there are no items for the Agreement States. There are five items for Appendix B (" Update of Previously Reported Abnormal Occurrences"), four ite:as for Appendix C
("Other Events of Interest"), and four items for "Other Events Considered for A0 Reporting." !
The dif ferences between the items shown in the Enclosures and the items suggested in our January 30, 1987 memorandum are described in Attachment A to this memorandum.
IE, in their memorandum dated March 18, 1987, suggested an additional' event as an A0 (i.e., a September 11, 1986 event at Zion 1 in which airborno radioactivity was drawn into the control room). The event is still being reviewed by AE00, NRR, and Region III for A0 reporting purposes. Therefore, it is not included in the draft report; it will be considered for reporting in the first quarter CY 1987 A0 report.
Information in this record was deleted g02pg;9g71019 in accordance with th edom of information GDRDON87-377 pyg gct m
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,7 'M itiple Addresses
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We ' request your review, comments and concurrence on this memorandum and its enclosur(s no later than May 27, 1987. Please provide updating information, if necessary, for all of the items in Enclosures 1 through 4 to reflect any cDanges in status. We plan to submit the report to the Commission by early June 1987. If you have any questions, please contact Paul Bobe at 492-4494.
In view of the NRC reorganization, please include in your reply the names, mail stops, and telephone numbers for the A0 Coordinator and the A0 Backup s Coordinator for your Office. We have enclosed (Enclosure 5) a list of the new Offices / Divisions and the present A0 Coordinators for your information. This is in accordance with NRC Manual Chapter 0212 ("A0 Reporting"). The Coordinator and his backup for AE00 are Paul Bobe (M.S. EWS 263A, 492-4494) {
and Jack Crooks (M.S. EWS 263A, 492-4425), respectively.
I In addition, please provide us the names of the other individuals in your Office who should routinely be placed on distribution for all A0 reporting correspondence.
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.L. dan, Director Office or Analysis and Evaluation of Operational Data Attachment and
Enclosures:
i As Stated cc w/ enclosures: 1 I
T. Rehm, E00 R. J. Brady, ARM /DS R. Gramann, NMSS J. M. Taylor, ED0/DEDRO S. A. Varga, NRR J. Glynn, RES J. Lubenau, ED0/E0 D. M. Crutchfield, NRR K. Murphy, RI T. Dorian, 0GC C. E. Rossi, NRR L. Bettenhausen, RI I J. C. Bradburne, OGPA/CA J. G. Partlow, NRR K. Landis, RII 4 J. J. Fouchard, 0GPA/PA H. J. Miller, NRR J. Strasma, RIII F. Ingram, OGPA/PA M. Caruso, NRR 0. Powers, RIV ;
J. R. Shea, 0GPA/IP J. W. Roe, NMSS J. Crews, RV C. C. Kammerer,0GPA/SLITP R. F. Burnett, NMSS R. Falkenberry, RV J. Lubenau, 0GPA/SLITP, R. E. Cunningham, NMSS J. A. Axelrad, OSP R. O'Connell, NMSS
? s Attachment A li Differences Between the AEOD'
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Memorandum of January 30, 1987, and the Draft Fourth Quarter CY 1986 A0 Report :
The following explains the differences between the items AEOD suggested to the staff (in our memorandum dated January 30, 1987) for inclusion in the Fourth I Quarter CY 1986 A0 Report, and the items actually included in Enclosures 1 through 4 of this memorandum.
The Offices mentioned below are those that existed prior to the April 12, 1987 NRC reorganization.
Abnormal Occurrences A0 86-21 (" Degraded Seft+y Systems Due to Incorrect Torque Switch Settings on Rotork Motor OperatoM. at Catawba and McGuire") is a new AO. NRR, in their memorandum dated M m h 18, 1987, suggested the event be reported as an AD. IE, in their memoranda dated March 18, 1987, suggested the event be reported in Appendix C. Bas 5d on discussions among AE00, NRR, IE, and Region II personnel, it was agreed that the item should be reported as an AO. The enclosed writeup is based on one provided by Region II.
A0 86-22 (" Secondary System Pipe Break Resulting in Death of Four Persons at Surry Unit 2") was suggested as either an A0 or Appendix C item in our January 30, 1987 memorandum, depending upon further evaluation. Both NRR and IE believed it to be an AD. Region II, in their memorandum dated February 17, 1987, believed that it should be an Appendix C ites. AEOD agrees with NRR and IE. Based on discussions with Region II, they indicated they would not object to reporting the event as an AO. The enclosed A0 writeup was prepared by AEOD based on the Appendix C writeup provided by Region II in their February 17, 1987 memorandum.
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A0 86-28,("Immediately Effective Order Modifying License and Order to Show Cause Issued to an Industrial Radiography Company") is a new A0 proposed by Region IV in their memorandus dated february 17, 1987. AE00 agrees that the item meets the A0 criteria for major deficiencies in management controls.
In our January 30, 1987 memorandum, we suggested a possible hand exposure of 512 rems at San Onofre Unit 3 as a possible A0. Region V, in their memorandum l dated February 6,1987, proposed deferring the item until the next quarterly report because the licensee's investigation and the Region's follow-up inspection activities were not yet completed. The licensee did not agree with the NRC findings. AE00 agrees with deferring the item'until a later quarterly report.
In our January 30, 1987 memorandum, we suggested that two overexposure at q Northwest X-Ray, a licensee of Idaho (an Agreement State) be reported as an A0 if the estimated extremity doses exceeded the A0 reporting threshold. The Office of State Programs, iri their February 9,1987 memorandum, stated that ,
the State's writeup would not be available in time for this quarterly report. l AE00 agrees to deferring the item until a later quarterly report. '
s Appendix B (Updating Items) 1 An update to A0 86-10 (" Willful Failure to Report a Diagnostic Medical 1 Misadministration") has been added by AE00 since close-out action was taken by the NRC in late December 1986. The enclosed writeup includes some information requested of Region I (J. Kinneman) by AE0D (P. Bobe).
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Enclosure 3 (to the Commission Paper) Items Item 4 (" Order to Show Cause Why License Should Hot be Revoked") was suggested l as a possible A0 in our January 30, 1987 memorandum. Based on the additional information provided by Region III in their March 9,1987 memorandum, AE0D l agrees with downgrading the item to an "Other Event Considered for A0 Reporting."
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Enclosure 1-
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For: The Commissioners Free: Victor Stallo, Jr.
Executive Director for Operations St& Ject: SECTION 208 REPORT TO THE CONGRESS ON ABNORMAL OCCURRENCES FOR OCTOBER - DECEMBER 1986 -
Purpose:
Approval of Final Draft Discussion: Enclosure 1 is a proposed letter to the Speaker of the House and the President of the Senate covering trans-mittal of the Section 208 report to Congress for the fourth quarter of CY 1986.
Enclosure 2 is a final draft of the quarterly report to Congress on abnormal occurrences (A0s). The report.
covers the period from October'l to December 31, 1986.
This draf t incorporates the major comments obtained from staff review of a previous draft. The draft report is similar in format to the third quarter CY 1986-report (published as NUREG-0090, Vol. 9, No. 3).
The draft report contains nine proposed A0s for NRC licensees. The items are:
86-20 Loss of Low Pressure Service Water Systems at Oconee.
86-21 Degraded Safety Systems Due to Incorrect Torque Switch Settings on Rotork Motor Operators at Catawba and McGuire Nuclear Stations.
86-22 Secondary Systes Pipe Break Resulting in Death of Four Persons at Surry Unit 2.
Contact:
Paul Bobe, AE00 X24494 h
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i T5'e Commissioners 2' N 86-24 Therapeutic Medical Misadministration (Cleveland Clinic Foundation; Cleveland, Ohio).-
86-26 Diagnostic Medical Misadministration (St. Luke's Hospital; Racine, Wisconsin).
86-27 Diagnostic Medical Misadministration (Toledo Hospital; Toledo, Ohio).
86-28 Immediately Effective Order Modifying License and Order to Show Cause Issued to an Industrial Radiography Company (Met-Chen Testin of Utah, Inc.;_ Salt Lake City, Utah)g Laboratories The three medical misadministration meet the Commission approved staff guidelines for selection of such events for .
A0s (i.e., Part II of NRC Appendix 0212). A0 86-24 involved a patient receiving a therapeutic dose greater than 1.5 times the intended dose. A0 86-26 and A0 86-27 involved patients receiving doses greater than five times the intended dose (in both cases, doses in the therapeutic range were received rather than diagnostic; damage to their thyroid glands is likely).
Tr.e.re were ne proposed A;s subcitted by the Agreement States for this report.
Appendix B of the draft report contains updating information -
for the following previously reported A0s:
Nuclear Power Plants 77 Environmental Qualification of Safety-Related-
-Electrical Equipment Inside Containment - This previously closed A0 is reopened to & scribe some new problems discovered during 1986. The item is then reclosed.
79-3 Nuclear Accident at Three Mile Island - Further information is provided and the item remains open.
86-15 Differential Pressure Switch Problem in Safety Systems at LaSalle Facility . Further information is provided and the item remains open.
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86-3 Rupture of Uranium Hexafluoride Cylinder and (
Release of Gaies (Sequoyah Fuels Corporation; Gore, Oklahoma) - Further infonnation-is provided for Sequoyah fuels Corporation, as well as for Allied-Signal Corporation, and the'ites is closed out. .
1 Other NRC Licensee.s.
86-10 Willful Failure to Report a Diagnostic'Nedical Misadministration (Mercy Hospital of Wilkes-Barre, Pennsylvania;' Valley Radiology Associates,'Inc.,
of Kingston, Pennsylvania) - Further. information' i is provided and the ites is closed out.
The draft report contains four items for Appendix C ("Other Events of Interest"). The items are:
- 1. Diesel Generator Problems (Calvert Cliffs, Zion..Palo Verde, Byron,.Braidwood, Nine Mile Point, and Cooper).
- 2. NRC Augmented Inspection Team Sen to' Hope Creek.
- 3. Conviction of International Nutronics, Inc., and One Employee in Federal District Court.
- 4. NR0 Aegr.ented Inspecticr. Team ~ 5 tnt to Hatch Facility.
When Commission _ approval is received, the report will be g-dated, if necessary,_ before it is published. The report-will be designated as NUREG-0090, Vol. 9, No. 4.
There are four items'for Enclosure 3 (i.e., items which ;
were candidates for inclusion as A0s, but which in the i staff's , judgement did not meet the criteria for A0 report-ing after further stu@) to this Commission Paper. In addition, they do not appear to meet the guidelines (i.e.,
Part III of NRC Appendix 0212) for Appendix C items.
Therefore, the staff does not plan to include the items in -
the quarterly report. The items are:
- 1. . Incineration of Molybdenum-99/ Technetium-99m Generator at a Hospital (Henry Heywood Hospital; Gardner,
' Massachusetts).
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( Commissioners N 3. Inhalation of Iodine-125 from Leaking. Sources '(IE .
Information Notice No. 86-95 regarding some leaking Lixi, Inc. imaging devices).
- 4. Order to Show Cause Why License Should Not'Be Revoked: 1 (regarding 'a idoctor' administering radiopharmaceuticals L to patients at an unauthorized facility). _
Recommendations: That'the Commission:
- 1. Approve the contents of the proposed Fourth. Quarter .l CY 1986 Abnormal Occurrence Report to Congress, and
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- 2. Note that upon approva1'and publication, forwarding letters to the Speaker of the House _and the President of the Senate will be provided to the Chairman fori signature. Congressional Affairs will then arrange.
for appropriate distribution to Congress. A. Federal Register notice will be issued to announce the avail-ability of the' quarterly report. In addition, a sepa-rate. Federal Register notice (describing details of-the ;
events) will be issued'for all of the A0s at NRC l licensees. No' press releases 'are planned.
Scheduling: While no specific circumstances require Commission action by a particular date, it is desirable to disseminate these quar-te '.y re::c ,: as socn a reasonably possible. It is'ex--
pected that Comniission action within two weeks of-receipt of the draft would permit publication and disse:aination'about '
three weeks later, if no'significant revisions are. required.
Victor Stello, Jr.
Executive Director for Operations _
Enclosures:
- 1. Proposed Letters to Congress
- 2. Draft of Fourth Quarter.CY 1986 Abnormal Occurrence Report to Congress .
- 3. Other Events Considered for Abnormal Occurrence Reporting
Enclosure 2 !
Page 1 of 2
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DRAFT-
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\ (Enclosure 1 of Commission Paper)
The Honorable Jim Wright Speaker of the United States House of Representatives Washington, DC 20515
Dear Mr. Speaker:
Enclosed is the NRC report on abnormal occurrences at licensed nuclear facili-ties, as required by Section 208 of the Energy Reorganization Act of 1974 (PL 93-438), for the fourth calendar quarter of 1986.
In the context of the Act, an abnormal occurrence is an unscheduled incident or event which the Commission determines is significant from the standpoint of pt&lic health or safety. The report states that for this report period, there were three abnormal occurrences at the nuclear power plants licensed to oper-ate. The events were (1) loss of low pressure ~ service water systems at Oconee, (2) degraded safety systems due to incorrect torque switch settin motor operators at Catawba and McGuire Nuclear Stations, andsecondary (3) gs onsys-Rotork
- tem pipe break resulting in death of four persons at Surry Unit 2 six abnorm lic o ratio ne era tag-one invo . mme ye ect' ve order modifying licenta en creer .s show cause istued to at, industrial radiography company.
There were no abncrmal occurrences reported by the Agreerient States. '
The report also contains information updating some previously reported abnormal occurrences.
In addition to this report, we will continue to disseelnate information on reportable events. These event reports are routinely distributed on a timely basis to the Congress, industry, and the general public.
Sincerely, Lando W. Zech, P.
Chairman
Enclosure:
Report to Congress on Abnormal Occurrences (NUREG-0090, Vol. 9, No. 4)
[ Enclosure 2 Page 2 of 2 <
i DRAFT 1 1
N (Enclosure 1 of Comission Paper) '
s The Honorable George H. W. Bush President,of,the Senate Washi 20510 ,
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Dear ident:
Enclose NRC report on abnormal occurrences at licensed nuclear facili-ties, as ired by Section 208 of the Energy Reorganization Act of 1974 (PL 93-438), for the fourth calendar quarter of 1986.
In the context of the Act, an abnormal occurrence is an unscheduled incident or event which the Comission determines is significant from the standpoint of public health or safety. The report states that for this report period, there were three abnormal occurrences at the nuclear power plants licensed to oper-ate. The events were (1) loss of low pressure service water systems at Oconee, (2) degraded safety systems due to incorrect torque switch settings on Rotork motor operators at Catawba and McGuire Nuclear Stations, and (3) secondary sys-tem pipe break resulting in death of four persons at Surry Unit 2. re were six abnormal occurrences at the o icensees. One inv -
nistrations one thera eutic nd two diag-s nostic ne in one M t.-
nvo ved an immediately effective or er mod' fy' ng
' license and o r to show cause issued to an industrial radiography company.
There were no a:en:rtal occurrences repeeted by the Agreement States.
The report also contains information updating some previously reported abnormal occurrences.
In addition to this report, we will continue to disseminate information on reportable events. These event reports are routinely distributed on a timely basis to the Congress, industry, and the general public.
Sincerely, Lando W. Zech, Jr.
Chairman
Enclosure:
Report to Congress on Abnormal Occurrences (NUREG-0090, Vol. 9, No. 4)
,, .. ' ' Enclosure 3 NUREG-0090 l Vol. 9 No. 4 ,
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DRAFT (Enclosure 2 of Commission Paper) .i REPORTTOCON5RESS ON ABNORMAL OCCURRENCES' OCTOBER - DECEMBER 1986 Status as of March 1987 Date Published: -July 1987 4
4 Office for. Analysis.and Evaluation of Operational Data United States Nuclear Regulatory Commission Washington, D.C. :20555
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N ABSTRACT Section 208 of the Energy Reorganization Act of 1974 identifies an abnormal occurrence as an unscheduled incident or event which the Nuclear Regulatory Commission determines to be significant from the standpoint of public health or safety and requires a quarterly report of such events to be made to Congress.
This report covers the period from October 1 to December 31, 1986.
The report states that for this reporting period, there were three abnormal occurrences at the nuclear power plants licensed to operate. The events were (1) loss of low pressure service water systems at Oconee, (2) degraded safety systems due to incorrect torque switch settings on Rotork motor operators at Catawba and McGuire Nuclear Stations, and (3) secondary system pipe break resulting in death of four persons at Surry Unit 2. were s 1 occurrences at the other NRC licensees.
[MME volved me ' cal misade'n's n e three n-era e c and two dia s c; one one an amediately effective or r mod conse and or show cause issued to an industrial radiography company. There were no abnormal occurrences reported by the Agreement States.
The report also contains information updating some previously reported abnormal occurrences.
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N CONTENTS P, age, ABSTRACT ........................................................ iii P RE F A C E . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . vii INTRODUCTION ............................................... vi1-THE REGULATORY SYSTEM ...................................... vii REPORTABLE OCCURRENCES ..................................... vili' AGREEMENT STATES ........................................... .ix FOREIGN INFORMATION ........................................- x REPORT TO CONGRESS ON ABNORMAL OCCURRENCES, OCTOBER-DECEMBER 1986.................................................. 1 NUCLEAR POWER PLANTS ....................................... 1 86-20 Loss of Low' Pressure Service Water 86-21 Systems at 0conee...................................
Degraded Safety Systems Due to Incorrect 1
Torque Switch Settings on Rotork Motor Operators at Catawba and McGuire Nuclear Stations... 5 86-22 Secondary System Pipe Break Resulting in Death of Four Persons at Surry Unit 2. . . . . . . . . . . . . . . 7 FUEL CYCLE FMILITIES (Othe than Wear Pwer Plants) .... 13 OTHER InstitutiNRCIndustrial LICENSEES Users,(Industrial Radiog)raphers, Medical etc. .................... 13
.. 13 a s .............. 16 86-26 Diagnos
............... 18 f Misadmi istra ............... 20 86-27 Diagnostic Medical Misadministration ............... 21 86-28 Immediately Effective Order Modifying License and Order to Show Cause Issued to an Industrial . Radiography Company. . . . . . . . . . . . . . . . 23- i AGREEMENT STATE LICENSEES .................................. 25
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REFERENCES ...................................................... 27 -I APPENDIX A - ABNORMAL OCCURRENCE CRITERIA . . . . . . . . . . . . . . . . . . . . . . . 31 f,
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1 O i CONTENTS (continued)
.P,ag j ss APPENDIX B - UPDATE OF PREVIOUSLY REPORTED ABNORMAL OCCURRENCES . 33 NUCLEAR POWER PLANTS ....................................... 33' 77-9 Environmental Qualification of Safety-Related Electrical Equipment Inside' Containment.............. 33 ;
79-3 Nuclear Accident at Three Mile Island................ 34 l 86-15 Differential Pressure Switch Problem l in Safety Systems at LaSalle Facility................ 2Hi FUEL CYCLE FACILITIES ...................................... 37 l
86-3 Rupture'of Uranium Hexafluoride Cylinder and Release l of Gases ............................................ 37 1
i OTHER NRC LICENSEES ........................................ 39 86-10 Willful Failure to Report a Diagnostic Medical Misadministration............................ 39 l APPENDIX C - OTHER EVENTS OF INTEREST ........................... 41 REFERENCES (FOR APPENDICES) ..................................... 51 l
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i 's PREFACE
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INTRODUCTION i The Nuclear Regulatory Commission reports to the Congress each quarter under' provisions of Section 208 of the Energy Reorganization Act of 1974 on any abnormal occurrences involving facilities and activities regulated by the NRC. ,
An abnormal occurrence is defined in Section 208 as an unscheduled incident or i event which the Commission determines is significant from the standpoint of public health or safety. ,
Events are currently identified as abnormal occurrences for this report by the NRC using the criteria delineated in Appendix A. These criteria were promul- i gated in an NRC policy statement which was published in the Federal Register on February 24, 1977 (Vol. 42, No. 37, pages 10950-10952). In order to provide wide dissemination of information to the public, a Federal Register notice is I issued on each abnormal occurrence with copies distributed to the NRC Public Document Room and all Local Public Document Rooms. At a minimus, each such notice contains the date and place of the occurrence and describes its nature and probable consequences. i i
The NRC has reviewed 1.icensee Event Reports, licensing and enforcement actions (e.g. , notices of violations, civil penalties, license modifications, etc.),
generic issues, significant inventory differences involving special nuclear material, and other categories of information available to the NRC. The NRC has determined that only those events, including those submitted by the Agree-ment States, da:ted in tMs repcrt aset the' criteria fer abnormal occurrence reporting. This report covers the period from October 1 tc Dececber 31, 1986.
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Information reported on each event includes: date and place; nature and prob-able consequences; cause or causes; and actions taken to prevent recurrence.
THE REGULATORY SYSTEM The system of licensing and regulation by which NRC carries out its responsi-bilities is implemented through rules and regulations in Title 10 of the Code of Federal Regulations. To accomplish its objectives, NRC regularly conducts licensing proceedings, inspection and enforcement activities, evaluation of operating experience and confirmatory research, while maintaining programs for establishing standards and issuing technical reviews and studies. The NRC's role in regulating represents a complete cycle, with the Mtc establishing stan-dards and rules; issuing licenses and permits; inspecting for compliance; en-forcing license requirements; and carrying on continuing evaluations, studies and research projects to improve both the regulatory process and the protection of the public health and safety. Public participation is an element of the regulatory process.
In the licensing and regulation of nuclear power plants, the NRC follows the philosophy that the health and safety of the public are best assured through the establishment of multiple levels of protection. These multiple levels can vil
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'be achieved-and maintained through regulations which specify requirements 'which will assure the safe use of. nuclear materials. The regulations include design and quality assurance criterit appropriate'for the;various activities licensed'
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[ . by NRC. nan inspection -and enforcement program helps assure compliance with the- 1 regulations.
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Most NRC licensee employees who work with or in the vicinity of radioactive materials are required to utilize personnel monitoring devices such 'as film badges or TLD (thermuluminescent dosimeter) badges. These badges are processed' periodically and the exposure results'normally serve as the official and legal-record of the extent of personnel exposure to radiation during the period the' badge was' worn. .If a'n individual's past exposure history .is known and has been sufficiently low, NRC regulations permit an individual in a-restricted area to l receive up to three rems of whole body exposure in a calendar quarter. : Higher values are permitted to the extremities or skin of the whole body. For unre-stricted areas, permissible levels of radiation are considerably smaller. ' Pe r-missible doses for restricted areas and unrestricted areas are stated in 10;CFR Part 20. In any case, the NRC's' policy is to maintain-radiation exposures to-levels as low as reasonably achievable.
REPORTABLE OCCURRENCES 1
Actual operating experience is an essential input to the regulatory process for; assuring that licensed activities 'are conducted safely. Reporting requirements-exist which require that licensees report certain incidents or events to the NRC. This reporting helps to identify deficiencies early and to assure.that--
corrective actions are taken to prevent recurrence.
For nuclear power plants, dedicated groups have been formed both by the NRC and l by the nuclear power industry for the detailed review of operating experience to help identify s!fety concerns early,.to iciorove dissemination of'such infer-l s.ation, and to feea back the experience into licensing, regulations, and ]-j operations.
In addition, the NRC and the nuclear power industry have ongoing efforts to improve the operational data system which include not only the type, and qual-ity, of reports required to be submitted, but also the method used to analyze the data. Two primary sources of operational data are reports submitted by the licensees under the Licensee Event Report (LER) system, and under the j Muclear Plant Reliability Data (NPRD) system. The former system is under the I control of the NRC while the latter system is'a voluntary, industry-supported i system operated by the Institute of Nuclear Power Operations (INP0), a nuclear utility organization. '
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Some form of LER reporting system has been in existence since the first nuclear power plant was licensed. Reporting requirements were delineated in.the' Code of Federal Regulations -(10 CFR), in the licensees' technical specifications, and/or in license provisions. In order to more effectively collect, collate, store, retrieve, and evaluate the information concerning' reportable events, the Atomic Energy Commission (the predecessor of the NRC)' established in 197,3 a computer-based data file, with data extracted from licensee reports. dating from j i 1969. Periodically, changes were made to improve both the: effectiveness of- j data processing and the quality of reports required to be submitted by the l licensees. '
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Effective January 1,1984, major changes were made 'to the requirements to re-port to the NRC. A revised Licensee Event Report System (10 CFR $ 50.73) was y established by Commission rulemaking which modified and codified the former LER - i system. The purpose was to standardize the reporting requirements for all I nuclear power plant licensees and eliminate reporting of events which were of i low individual significance, while requiring more thorough documentation and :
analyses by the licensees of any events' required to be reported. All such re . ,
ports are to be submitted within 30 days' of. discovery. The revised system-also' !
permits licensees to use the LER procedures for various other< reports required under specific sections of 10 CFR Part 20 and Part 50. TheJasendment to the ,
Commission's regulations was published in the Federal- Register (48 FR 33850)' on 1 July' 26,1983, and is described in NUREG-1022, " Licensee Event Report System," j and Supplements 1 and 2 to NUREG-1022. j q
Also effective January 1,1984, the NRC amended its immediate notification' re- ,
quirements of significant events at operating nuclear power reactors (10 CFR
$ 50.72). This was published in the Federal Register (48 FR ~39039) on l August 29, 1983, with corrections (48 FR 40882) published on September 12, j Among the changes made were the use of terminology, phrasing,- and 1983. l reporting thresholds that are similar to those of 10 CFR 5 50.73. Therefore, most events reported under 10 CFR $ 50.72 will also require an-in-depth follow-up report under 10 CFR l'50.73.
l The NPRD system is a voluntary program for the reporting of reliability data 1 t,j nuclear power plant licensees. Both engineering and failure' data are to be l submitted by licensees for specified plant components and systems. In the past, industry participation in-the NPRD system was limited and, as a result, 1 the Commission considered it may be necessary to make participation mandatory I in order to make the system a viable tool in analyzing operating experience.
However, on July 8,1981, INPO announced that because of its role as' an active user of NPRD systo data, it wcu c assume responsibility for management and i
funding of the NPRD system. INPD reports that significant improvements in licensee participation are being made. The Commission considers the NPRD sys-tem to be a vital adjunct to the LER system for the collection, review, and !
feedback of operational experience; therefore, the Commission periodically moni- i tors the progress made on improving the NPRD system.. j Information concerning reportable occurrences at facilities licensed or other-wise regulated by the NRC is routinely disseminated by the NRC to the nuclear ,3 industry, the public, and other interested groups as these events occur.-
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 than 100 local. public document rooms throughout the Unit,ed States and to the IRC Public Document Room M in Washington, D.C.
The Congress is routinely kept informed of reportable events occurring in lice ised facilities.
AGREEMENT STATES Section 274 of the Atomic Energy Act, as amended, authorizes .the Commission to enter into agreements with States whereby the Commission relinquishes and the ix r
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States' assume regulatory authority over byproduct, source and_special-nuclear-materials (in ' quantities. not capable of. sustaining a chain reaction). Compara -
ble and compatible programs are the basis for agreements.
.I N i Presently, information on reportable occurrences.in Agreement State. licensed 'I activities is publicly available at the State level. Certain information.is also provided to the NRC under exchange of information provisions in the agreements.
In early 1977, the Commission determined that abnormal occurrences happening at .
facilities of Agreement State licensees should be included in the quarterly. . l
' reports to Congress. The abnormal occurrence criteria included in Appendix A' !,
is applied uniformly to events at NRC and Agreement- State licensee facilities. .;
Procedures have been developed and implemented and abnormal occurrences reported .{
by the Agreement States to the NRC are included in these quarterly reports-to ;
Congress. {
l' FOREIGN INFORMATION The NRC participates in an exchange of information with various foreign govern- i ments which have nuclear facilities. This foreign information is reviewed and j 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 ab .ormal occurrences are reported. l l
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REPORT T0. CONGRESS ON ABNORMAL OCCURRENCES OCTOBER-DiCEMBER 1986
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NUCLEAR POWER PLANTS The NRC is reviewing events reported at the nuclear power plants licensed to 'l operate during the fourth calendar quarter of 1986. As of the date of this report, the NRC had determined that the following events were abnormal occurrences.
86-20 Loss of Low Pressure Service Water Systems at Oconee The following information pertaining to this event is also being reported con- '
currently in the Federal Register. Appendix A (see the third general crite-rion) of this report notes that major deficiencies in design, construction, 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") I 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 - On October 1,1986, Duke Power Company (the licensee) twice attempted an electrical load shed surveillance test of. circuits'on Oconee Unit-2, which was shut down for refueling at the time. During both tests, the low-pressure service water (LPSW) system was lost. ' Investigation revealed a ques-tionable design feature which was also'applicaole to Units T and 3. Therefore, the LPSW systems for all three units were considered inoperable and on October 2,1986, orderly shutdowns of Units 1 and 3 (both operating at 100% power at the time) were commenced.
Oconee Units 1, 2, anc 3 eacF ;.41he e Eebca:k & Wilcox-designed pressurized water reactor; the facility is located in Oconee County, South Carolina.
Background - At Oconee, the condenser circulating water (CCW) system takes suc-tion from Lake Keowee and supplies water to the main condensers. Unlike most nuclear power plants, the CCW pumps at Oconee also perform safety-related func- ,
tions which include: supplying cooling water to the LPSW system, the cooling water pump for the standby shutdown facility emergency diesel' generator, and the turbine-driven auxiliary feedwater pump for. long-tern cooling. The LPSW system supplies cooling water for the decay heat removal (OHR) system and other safety-related equipment. The LPSW system pumps take suction on the upstream >
side of the condenser from the CCW system crossover lines between Oconee Units 1, 2, and 3.
Each of the four CCW pump motors for each Oconee unit is capable'of being powered from either of two emergency hydro generators. However, the Oconee plant is designed to accommodate a loss (shedding) of the CCW pumps and still provide LPSW pump suction through a siphon arrangement. The siphon is neces-sary because of a high point in the CCW piping just downstream of the CCW pumps.
and upstream of the LPSW pump suction. This high point may be as much as 25 feet above the level of Lake Keowee (depending upon lake level). '
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$ture and Probable Consequences .On October 1,1986, while Unit 2 was'in a refueling outage, the Unit 2 load shed _ test' was performed.' . At Oconee, a load shed of non-essential loads is initiated when emergency power is required via the underground. feeder from Keowee Hydro Station. . -The load shed protects 'this power path from overload. _l When the load shed test was initiated, the condenser ci'rculating water. pumps- .
.were deenergized. Normally,;this causes' the gravity flow system to automat-1 ically align and to allow the flow of water from the.LakeL Keowee intake 'struc-ture through the condenser and discharging to the Keowee'tailrace into Lake . ']
Hartwell. The elevation difference and a . siphon effect are used to cause' the '
condenser circulating water to continue to flow. This mode of. sperat Sn of - the- l CCW system is . referred to as the emergency condenser circulating wates IECCW)-
system. For this. test, the condenser gravity drain to the Keowee tailrace was blocked because this was not part of -the test.
After about an hour, the LPSW pumps began to cavitate and. stop pumping. One ,
LPSW pump was stopped by the control' room operator, and a second LPSW pump'was I observed to have low discharge pressure and cycling amps. Various high tem- l perature alarms for the components cooled by LPSW were received in the control- 1 room. CCW flow was restored by restarting a CCW pump and the plant was restored to its normal power condition without any plant damage or system upsets. Prior to the occurrence, two LPSW pumps were operating with approximately 13,000 gpm/ pump.
The CCW crossover head, which provided suction for the LPSW pumps, was being i supplied by the Unit 2 CCW pumps at the time' .
In the evening of October 1,1986, the test was repeated, but this time the gravity drain feature.was also tested. The results were the same as Lin the first test, i.e. , loss of LPSW system function due to loss of LPSV pump suc-tion. NRC Region II was advised of these results late. in the evening, and I concurred with tne licensee that Units 1 and 3 (operating at'100K power)- could I continue to operate until the test data'could be fully evaluated. : At 9:00 a.m.
. on October 2,1986, evaluation'of the tests revealed that the operation of this design feature (the ECCW system) was ' questionable for Units 1 and 3, and .that this resulted in inoperability of all LPSW systems .for Oconee. As a result, an i orderly shutdown of the two operating units was begun as required by Technical l, Specification 3.3.7. Both units reached cold shutdown conditions by October 3, J 1986.
Investigation showed that the loss of suction to the LPSW pusps was caused by I a loss of the previously described siphon. . The CCW pump discharge flange is normally nine feet below the surface of_ Lake Keowee when the lake.is at full i level. However, because of drought conditions, the lake level was'about six feet below the flange at the time of the load shed test. _(Technical. Specifi- 1 cations permit plant operation with lake:1evels as low as 16 feet below this - !
flange.). 'During operation at these reduced. lake levels, water leakage at' the flange had been observed. This flange was not originally designed to be leak-tight, but the' amount of water leakage was insignificant during plant operation.
However, with the CCW pumps.off (shedded), air inleakage caused the high' point-in the CCW system piping to drain and caused a loss of- siphon flow in about 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. ,
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siphon flow, if initiated, could not be sustained in the system,'as originally designed and built, during. low lake level conditions 'because of. air inleakage at the' CCW pump discharge flange. It appears that previous; surveillance tests.
were not of sufficient duration to determine that siphon flow was_ sustained.
'Since.the large volume of water contained in-the CCW lines provided LPSW flow for about an hour before the loss of LPSW suction, it appears that load shed -
testing personnel,.in the past, may have been misled into thinking siphon flowj had been sustained.
As mentioned briefly in the " Background" information above, the Oconee CCW system is designed to also provide suction and discharge (heat sink) for the.
cooling water pump for an emergency diesel generator-(EDG) used in the standby shutdown facility (SSF). The SSF was designed to be an independent DHR system. '!
Analysis performed subsequent to the above load shed test showed that if siphon-
flow were lost in the CCW system pipe,-the CCW system could not provide an adequate heat sink for SSF operation to meet its design basis of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of operation. In addition, when the CCW pumps are not ' operating, the CCW system:
should provide emergency gravity-siphon ~ CCW system flow to the main condensers-to recover condensate for DHR.following certain postulated events until"the DHR system is in operation. The gravity flow is possible because the CCW system discharge from the main condenser is shifted to an. alternate pipe.that dis-charges downstream of Lake Keowee dam at an elevation well below the CCW system- ;
intake. The siphon is required for the same reason as required by the LPSW l system. This feature of the CCW system also was disabled by the loss of siphon.
The safety significance of the October 2,1986. event at Oconee is that the' event revealed an unanticipated failure mode for a. diverse, passive means for ensuring adequate DHR following design basis accident. scenarios as' well as a Station Blackout condition. This means of losing the' safety-related functions of the LPSW system has existed ever since .the first Oconee unit went-into operation in 1973.
Cause or Causes - The root cause of. this incident is the inadequate design _.and testing of the ECCW system. This led to a failure of the ECCW system to perform its intended function as described in the FSAR under all. assumed !
conditions. Inadequate original design evaluation of the ECCW system and the !
lower than normal lake level due to extreme drought conditions of Lake Keowee are contributing factors to the cause of this incident.
Actions Taken to Prevent Recurrence l Licensee - The licensee modified the discharge flange on all CCW pumps to pre-vent air inleakage when the lake level is below the discharge flange. The LPSW pumps were successfully tested for several hours with the CCW pumps off and the lake level below the dischart,3 flange. The meergency CCW gravity-siphon flow j
to the main condensers and the EDG cooling water pump also were successfully 1 tested under the above conditions. In addition, the SSF cooling water pump was I muodified to take a separate and independent suction from Lake Keowee. ' '
J The licensee inspected each continuous vacuum priming (CVP) line at the CCW !
system intake for blockage. Unit 3 lines were clear and vacuum was' established d on Unit 3 intake high point vents. . Unit I and Unit 2 lines were found blanked' off with blind flanges which prevented the CVP pumps from developing adequate t vacuum'on the CCW system intake high point vents to overcome air. inleakage at j g
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tNe~ pumps. These flanges had apparently not been removed at the completion of
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the original system hydrostatic testing.~ -The flanges were removed and a vacuum us established on Units 1 and 2 intake high point vents. ,
N Successful testing was performed on the condensate steam air ejector, the tur-bine bypass valves (TBV), and the siphon effect. By October 23, 1986,'a11 three units were returned'to service.
.Further corrective actions planned by the licensee include: (a) review and analyze the seismicity of the CCW system, (b) develop a program to include CCW system piping in a routine. inspection, (c) review the validity. of the testing program to ensure that systems.and components are ' tested aiequately, and (d) review Technical Specifications to determine if any revisionsJare necessary.
NRC - The NRC Resident Inspectors: for.the Oconee site were'. observing the Unit 2 a load shed test'in progress on October 1,1986, and. observed the' failures. They observed licensee actions to assure that the Units remained in a stable'condi -
tion and notified Region II. Initial investigation of the circumstances associated with this event began while surveillance . test data.was being analyzed.
On October 4,1986, the Deputy Director of the Division of Reactor Projects, .
Region II, went to the site to observe the licensee's repairs for. a fix..of- the l
problem, to review th'e potentia 1 'of the proposed fix to solve the problem, 'and' to assess the overall significance of' the event, j On October 8,1986, a meeting between NRC.and licensee personnel wa's held in l j
the NRC Region II Office to discuss actions being taken and planned by the _ j licensee to repair and demonstrate operability of the.0conee units. _ Actions to "
be taken and required conditions prior to restart of the' units were also l discussed.
g l The NRC Resident Inspectors witnessed the repairs'and the subsequent testing to confirm the overall adequacy of the licensee's corrective action prior to . l restart of the units.
On October 14, 1986, an NRC Management Meeting with the. licensee was held in Bethesda, Maryland, to review the completed modifications and test results.
The licensee presented the results of the surveillance which were conducted following the described modifications to' the pumps.
NRC Inspection Reports No. 86-26 and No. 86-33 concerning the incident were forwarded to the licensee on October 23, 1986,- and December 1, 1986,.respec-tively (Refs. I and 2). On December 22, 1986,> an NRC Enforcement' Conference j
was held at the NRC Region II Office to discuss the event with the licensee" 1 '
(Ref. 3). On February 5,1987, the NRC issued a Notice of Violation to the licensee (Ref. 4), regarding the operability of the ECCW system.. The viola- <
tion was classified as Severity Level IV (on a scale-where Levels I and V are 1 the most and least severe, respectively).
j On January 30, 1987, the NRC issued Inspection and Enforcement Information I Notice No. 87-06 (Ref.~ 5) to all nuclear power reactor facilities holding an operating license or a construction permit to inform them of.the Oconee event.
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1 This item is considered closed for the purposes of this report. '
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" l aaaaaaanam 86-21 Degraded Safety Systems Oue to Incorrect Torque Switch Settings onQotork Motor Operators at Catawba and McGuire Nuclear Stations The.following information pertaining to this event is also being reported con-currently in the Federal Register. Appendix' A (see Example 10 of "For All-Licensees") of this report notes that a. major _ deficiency in design, construction, or operation having safety implications requiring immediate remedial action can be considered an abnormal occurrence.
Date and Place - On October 23, 1986,' Duke Power Company (the licensee) dis- 'l covered that numerous safety-related valves at Catawba Nuclear Station were degraded due to incorrect torque switch settings. On October 28, 1986, a similar situation was also found at the licensee's McGuire Nuclear Station. Catawba Units 1 and 2, and McGuire Units 1 and 2 are Westinghouse-designed pressurized ,
water reactors. The Catawba Station is located'in York County, South Carolina .j and the McGuire Station is located in Mecklenburg County, North Carolina. 1 1
Background - At Catawba and McGuire, and many other plants, Rotork Actuators are used for remote control of plant valves. Many of these valves are in safety-related systems. The actuators are driven by electric motors. The size of the motor and the actuator depends on the size of the valve and the- force or torque necessary to open and close the valve. Rotork Actuators have five torque switch settings which the licensee had assumed represented 40, 55, 70, 85 and 100 percent of the maximum rated torque output. The required torque switch setting for each actuator is determined based on the maximum differential pres- '
surre expected on its associated valve during anticipated events.
If incorrect torque switch settings are used, the valves may not. perform as designed (e.g. , the actuator mctor may switch off before the- valves complete thieir travel) thereby possibly degrading their function to avoid, or mitigate, the consequences of a transient or a design basis accident. ' Improper torque switch settings, on motor operators manufactured by various vendors, have been a contributing cause in a number of significant events, one of the most serious of which was the complete loss of main and auxiliary feedwater event at Davis-Besse on June 9,1985. The Davis-Besse event was' reported as abnormal occurrence No. 85-7 in NUREG-0090, Vol. 8, No. 2 (" Report to Congress on Abnormal Occurrences: April-June 1985").
On November 15, 1985, the NRC issued IE Bulletin No. 85-03 (Ref. 6) to all holders of nuclear power .aactor operating licenses or construction permits for action. The Bulletin described various events (including the June 9,1985 event at Davis-Besse) during which motor-operated valves failed on demand, in a common mode, due to improper switch settings. The Bulletin'also described numerous previously issued NRC reports, Information Notices,: Bulletins, and Circulars '
(as far back as '1972) involving problems with; torque. switches.
Bulletin No. 85-03 requested licensees to develop and' implement a program to ensure that switch settings on certain safety-related motor-operated valves are !
1 selected, set, and' maintained correctly to accommodate the maximum differential.
pressures expected on these valves during both normal and abnormal events within the design basis. It was during the licensee's followup to the Bulletin that' the problem at Catawba and McGuire was discovered.
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' Nature and Probable Consequences Catawba Nuclear Station (CNS) l On October 23, 1986, with both CNS units in cold shutdown, a valve was being repaired. The motor and worm gear had been replaced and the valve's actuator was being calibrated per CNS plant procedure. : After setting the. torque switch to.the specified setting, per procedure, while the valve was on the test bench, the torque output was checked. The result indicated the. torque output was lower than required. Subsequently, a performance curve (percent torque output versus torque switch setting) was obtained. As stated above, the licensee had assumed
.that setting 1 represents 40. percent of rated torque and setting 5 represents 100 percent of rated torque.
A linear or straight curve had been assumed by the licensee for the range from
, 40 percent to 100 percent. However, results of the performance curve indicated a non-linear relationship between percent- torque output'versus torque switch l setting. Also, the curve obtained was generally lower than the linear rela .
tionship previously thought to be correct. l 1
l Based on these results, the valve being tested at CNS was determined to be 1 incapable of performing its intended function. Subsequent review indicated !
that at CNS this situation was not-unique to the single valve discussed. I Fifty-three valves were determined to potentially be affected. Systems which -
1 contained safety-related valves (many of which were containment isolation valves) l which could potentially- be affected were: 4
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(a) Chemical and . volume control systems (b) Component cooling system (c) Residual heat removal system .j (d) Ice condenser ref rigeratin syste- ,
(e) Safetyinjectionsystem ,
I (f) Nuclear service water system i (g) Containment hydrogen purge system i (h) Breathing air system (1) Instrument air system (j) Containment air release and addition system McGuire Nuclear Station (MNS)
Based on the above event at CNS, it was determined at MIS on October 28. 1986 that the charging line outside containment isolation valves for both MNS units were technically inoperable. In fact, analysis of the two valves, which would. ,
be required to close during safety injection, indicated that they may not be able to do so under differential pressure conditions which could be encountered following a loss-of-coolant accident.
Unit I was operating at 100 percent and Unit 2 was operating at 47 percent of full power. With both trains of the emergency core cooling system thus inoperable, the licensee commenced shutdown of both Units. A plan was estab-lished to inspect all safety-related Rotork motor operators and perform a-detailed engineering evaluation on each affected valve to determine the appro-priate corrective action to ensure its' operability.
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- C'ause' or Causes .- The cause. of the problem appear to' be poor communication d beteen the motor operator vendor and the licensee. ~The vendor subsequently;
' stated that whenever the factory torque l setting is changed in the field, ~an '
individual calibration curve or bench test is required to accurately determine I torque output. In fact, based on verbal communication with:the vendor,"the" j licensee utilized a linear or straight line curve for the relationship between percent torque versus ' torque switch setting.
Actions Taken to Prevent Recurrence Licensee - The licensee took the following corrective steps: *l
.1 a) ' Operability requirements for all safety-related.Rotork motor operatorst were evaluated individually. -l b) Corrective action for each motor operator evaluated was performed,Jif required.
c) Corrective maintenance procedures for Rotork motor operators were revised to require a bench test of the operator when setting torque switches.
d) A detailed inspection of each safety-related Rotork'. motor. operator installed was performed, e) All significant discrepancies found during inspection were corrected.
f) A new document which specifies the required torque output (in terms of-torque output, and not torque switch setting) for each valve with a Rotork motor operator was completed and approved.
.I g) The licensee committed to implement a program by. November 15, 1987, to ensure that torque switch settings on all safety-related motor' operated valves are selected, set, and maintained correctly. This commitment was made in accordance with the previously discussed.NRC'IE Bulletin No. 85-03.-
p i SRC - The NRC monitored the licensee's corrective action to assure that they :I l
were responsive. Subsequently, NRC met with the licensee and Rotork to follow 1 progress on the issue.
On November 3,1986, the NRC issued IE Information Notice No. 86-93 (Ref. '7), H which described the McGuire event, to all nuclear power reactor facilities a holding an operating license or a construction permit. l This item is considered closed for the purposes of this event. I a*aaaaaaa=
l 86-22 Secondary System Pipe Break Resultina in Death of Four Persons at Surry Unit 2 j The following information pertaining to this event is also being reported concurrently in the Federal Register. Appendix A (see Example 10 of "For All Licensees") notes that a major deficiency in design,' construction, or _ operation. '
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having safety implications requiring immediate remedial action can be considered an abnormal occurrence.
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Date and Place - On December 9,1986, with bbth!Surry Units 1 and 2 -operating . ,/ '
at 100% reactor power, a Unit 2 reactor trip (stras) followed by a main feedwatei ' e (MFW) line rupture occurred. Eight individuals were injured by the escaping steam and water. Four of those injured subsequently died. ,
.The Surry Power Station consists of tw westinghouse-designed pressurized water I '
reactors. The station is operated by Virginir.iflectric and Power Company (the r' licensee) and is located in Surry County, Virginia.
' ' l Nature and Probable Consequences - On December 8,1986, Unit 2 had completed a refueling outage and had returned to fu11' power operation. Unit l was also operating at 100% power.
On December, 9,1986, at about 2:20 p.m. (EST), a. low-low level in the "C" steam \I generator (S/G) of. Unit 2 caused an automatic reactor trip and an automatic start of the two Wotor-driven auxiliary feedwater pumps. The reactor trip also resulted in a trip of the Unit 2 main turbine generator, s /
The control room operators noted the S/G code safety valves lifting and regulated S/G pressure through the atmospheric dump valves. Approximately 30 seconds after the trip, the unit's electrical busses auto-transferred to offsite power.
A small steam release noise was heard followed by a very loud noise approxi-mately five seconds later. [.)j .
ir A shift supervisor, who was in the turbine building observing construction.
activity areund the MFW pumps, realized that a large break /
had o to the control to alert the control room operators. ( ('
He also told.them that people had been injured. . All seccMary pumn were secured and thel >reak .
isolated. Water to tne S/Gs us supplied by the auxiliary f eenater system. l The primary system responded normally to the loss of load transient. Reactor coolant temperature, pressure, and pressurizer level were stabilized in the desired band. '
A notification of unusual event was declared at 2:27 p.m. At 2:30 p.m. , ground and air ambulances were called to take the injured people to the hospital. At 2:40 p.m. , the unusual event was upgraded to an Alert in order to ensure accountability of all station personnel.
Unit 2 was stabilized by 2:34 p.m. with two reactor coolant pumps running and primary systes pressure and temperature being maintained by relieving steam to the atmosphere. No radioactive releases resulted from the event and a cooldown was initiated. (Unit 2 was in the. cold shutdown condition by 7:04 a.m. on December 10,1986;)
At 3:48 p.m., accountability of personnel was completed and at 4:25 p.m. , the Alert was terminated. At 3:55 a.m. on December 10, 1986, Unit 2 was placed on the residual heat renova (system; at <7:04 a.m. , the Unit achieved cold shutdown conditions. '
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'" k ^ ' Later investigation sh' owed that.an 18-Ir.ch suction line to the' A train tr s i rO feedwater pump had ruptured at the elbow where the line connects to the-n 24-inch condensate header. At the time rof the. rupture, contracter personnel il '
' ~ h' f/ employed by Daniel Construction Compant of. Greenville, South C t
Insuhtion Specialties, Inc., of Hopasell. Virginia, were doing instrument l
'lline 4jipe.
relocation and pipe @nsulation,wWk in the general: area of As a result 'of the escaping: steam and water, six of these individuals:
sere , hospitalized forithaisent of severe bu,rns. Three were evacuated directly 1 frusth'e- site by helic /pter, and three others were taken off site by amb'uiance. The other wg who were less. severely injured, were treated at a.
clinic thd releasedi .
One of tisse hospital} zed died the afEernoon of December 10, 1986, and another victim died on December 11, 1986. Two others died several days later. The other,tp remained lirl seilaus to criticall condition. 1 and was' later releaseddu;t ,the other was'still in serious. condition more thanj One of the two' improved a month after th( ace,ident; J Iq j (
The. escaping steam and water also caused various system interactions which complicated 4he complication Vas,licentpe's handling of the event. For example, one important that '. vater and steam saturated a security card reader which -
shorted out th'e entire plant card readef system. As a result,, key-cards would not open phnt doors. Security personnel responded to the control room and 4
provided'9 pus control, while doors into'the control room were opened for easy access ar# to imorpve control room ventilation. Guards admitted personnel' on -
the badh 'er pe'rnna1 recognition and excluded non essential personnel. The 1
card reader system returned to service approximately 20 minutes after the pipe break and functioned normally thereafter.
An operator ported being delayed in the stairway outside the control room as a result of the card reader failure. Due to the hot water conditions on the turbine building basement floor and the discharge of the Halon fire suppression system in the emergency switchgear rooms below the control room and.the carbon dioxide fire suppression system in the cable tray. rooms above the control room fi (see discussion below), the operator had no safe way to exit the stairway other
};, than the control room itself. 1 someone 6pening the door from inside the control room.The operator was admitte The licensee is con-sidering installlr,g%1ectronic override switches which would permit the opening of electronically locked doorf in emergency? situations.
Another ,important4horpiication was that within minutes of the feedtater pipe rupture event inlthe Unit 2 turbine building, portions of the Unit 7. turbine building sprinkler system actuated. Sixty $wo sprinkler heads opened in the immediate area of the feedwater pipe ruptum due to the high heat levels associated with' the event. As they opened,,th'ese sprinkler heads issediately begargI 'M$ charging water to cool the turbine building atmosphere.
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Water shoft-circuitedz ct,ntroi systM,f r
S in the area for the Halon and carbon d' ioxidtOff re suppression 'sys}qus; as a result,Mhn systems activated.. Control .i room habitability became aponcern because doors were blocked open to allow; better,ontrol room access without recognizing that' carbon dioxide had been dischsged in the areas above.
The carbon dioxide was apparently coming into s
r thf c6ntrol room from the turbine building hallway. Control room personnel'in l,fj) /'
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the main control room annex,and near the main contro11Fcce door experienced '
shortness' carbon of breath, dioxide was dizziness present, the andcontrol nausea.room However,Lonce' operators took' they approprMe nyoprizr # that; correc- q tive act4Sns and initiated control room. emergency air; supply fans which placed '
the main control rcom at a higher pressure than the turbine building.i This-action assisted in d!1uting and exhausting the ' existing carban dioxide. levels-and kept apy additional carbon dioxide from infiltrating into the g' ma!q pontrol.
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About- 3:30 p.mi o' n D#emtier 9g 1986, a decision was -made by N fteiNn'IIL1 .
l nanapsent to sen1 an'inspec. tion team to the site. Thisteamconsk,sted'of 1 Regional-baud personne1 Land the senior resi~ dent inspectors faos North Anna and' Surry. The IyJion.II inspection team arrived on site abouto:30 p.dont i December 9,1580, to assess the operational status of the~ unit and to ins; Wet ' i the-damaged area of the Unit 2 turbine building. :During the morning of ? 4 1 December 10, 1986, the team's status was upgraded to.an Augmented-Inspection A
- g Team (AIT), and an engineer from.the NRC Office of Nuclear. Reactor Regclation,- !'
knowledgeable ir/ water hammer phenomena, was assigned to the team. The AIT was j
- t. asked with the. job of examining the licensee's response to;the incident and ~'g, performing a separate investigation r ,
V' I Following the termination of the Alert classification $n December 9,1986h I licensee management, initiated recovery activities. An organization wh established and resources identified'for evaluating the incident and recoscend-ing recovery actims. .The 'licenste's. preliminary findings resulting f rom the ,
Unit 2 main feed p@ e6ction pipe rupture indit.ated that there may have teen > :-
si nificant thinning'of the pipe Will due to a corrosion / erosion mechanisk not ' >t 1 fu ly understood at the time. Since the same mechanism could!similarly affect- I ]<
. ' Unit 1; at ]2:30 p'.,m. on December 10, 1986, the licensee decided to shut down .
) s Unit A 3 Thti shutdown was initiated at 5:30 p.m. on December 10, 1986, and the *'
unit us off t~ en lire at 10:47 p.m. ' '
Subsequent.ly, the unit was placed in a ,\*l cold sddown cordition.
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- i 3 Based on the preliminary results of the licensee's investigation an'd the en NRC Regian II AIT fr,$pection/ investigation,'it has been concluded that the\g(in ;
following factors attributed to the main fe'edwater pipe rupture events.
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- 1. Pipe Wall ThinnUg on "A" Main Fee (, Pump Suction Line 3 The 18-inch suction line which supplies'the fA" Train Main Feed Pump was fabricated using ASTM A-106, Grade B, Extra Strong, carbon' steel seamless pipe and ASTM A-234, Grade B ,Evtra Strong, WPB ~ carbon steel. wrought fit-tings which had a nominal wall thi4 ness of:.50 inches t10% at installation.
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i Since instal)ation, the bulk ringit phase corrosion / erosion mechanism had t subsequently reduced the originil vill thickness. . Ultrasonic wall thick-t ness measurements and micrometer, measurements taken on the elbow following. ,
.I vthe failure showed a gradually sloping wall thickness loss'ovep much of l t. A the suction line. At several locations, usually near welds, localized
- / cavities had been formed in the elbow inner surface by the corrosion /
/ eronion process. The maaining wall thickness of these localized areas has' been measured as 1d as .048 ' inches while adjacent lot.atior..: were .090 l t4440 inches in thickness. Using the code minimum wall equation and '
assuming an intejnalgipe pre:.sdre ef 600 psig, a temperature of 370'F, 1
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ness of .090 inches and a yield thickness of .173 inches. This difference-in the pipe thickness directly contributed to the pipe failure.
- 2. Ma h Steam Trip Valve and Instrument Air Pressure Surry Unit 2 tripped as designed when a low-low.S/G water level protec-tion signal occurred on the C" Steam Generator Reactor Protection System.
Instrumentation. This occurrence was a result of the unplanned closure of the"C"sainsteamtripvalve(MSTV). The MSTV closure was initiated by a slight reduction in instrument air pressure to a steam flow assisted' closure. The "C" MSTV would not be ex reduction in instrument air )ressure; pected to close with the slighthowever, thej that the valve disc had not' >een in a fully open position because of mis - l alignment' of the valve bonnet'that occurred when the valve was overhauled .q during the past refueling outage. This slight reduction in instrument air pressure allowed enough deflection for the steam flow to force the' disc l shut. This directly contributed to the' trip, but not toithe pipe rupture.
- 3. Main Feed Pump Discharge Check Valve
]
1 The licensee's inspection of the 2A main feed pump check valve revealed 1 two hardware' deficiencies, i.e., a missing disc hinge pin and a displaced seat-disc assembly. Further investigation showed that these deficiencies did not contribute to the cause of the break, but may have permitted an additional amount of water to be discharged through-the t,reak.
Cause or Causes j I The investigations have indicated that the rupture of the suction line elbow resulted from the combination c' wall thinning due to bulk sinole phase corrosion / erosion and a normal feed pump suction pressure transient. The root i causes appear to be a design deficiency (associated with piping configuration R and flow velocity) and an operational problem (associated with water chemistry).
Actions Taken to Prevent Recurrence J Licensee - The licensee developed a comprehensive plan for inspection, evalua-tion, -and modification / replacement, as necessary for components (e.g. , pi main steam trip valves, main feed pump discharge check valves),e.systems g. , (ping, fire protection s
- communications).The ystem) andalso licensee procedures developed(e.g., security a station plan, startup emergency.
plan. On plan and January 14, 1987, Unit 2, Reactor Trip and Feedwater Pipe FailureRef. Report"the. licensee fontarde 8), which pro-vided detailed information on the event together with the' station recovery-plan and corrective actions for NRC revlew and concurrence prior to station -
1 restart. As discussed below under "NRC Actions," the licensee's plans and ac-L tions were acceptable to the NRC. Subsequently, Surry Units 1 and 2 were !
'o returned to service on February 23, 1987 and March 20, 1987, respectively. !
The licensee also operates the North Anna nuclear power station which consists of two units similar to those of Surry. The facility is located.in Louisa County, Virginia. Subsequently, when pipe wall thinning was found at Surry 1
11 i 1
U6f ts 1 and 2, the licensee decided to inspect similar pipin at North Anna Unit 1. Therefore, on December 25, 1986, power was reduced rom 100%, reach-ing 20% on December 26, 1986. Ap i made on Worth Anna Unit 1 piping. proximately 4900 No measurements ultrasonic indicated p e wallinspections thick- were ness below the required minimum. The feedwater pump suction p ing and header wall thicknesses were within original pipe manufacturing speci cations, and the high pressure drain pump discharge piping was no more than 15 percent below H the original specifications. Since no abnormal conditions were found, Unit'l a was returned to full power on December 27, 1986. North Anna Unit 2 and addi-' a tional Unit 1 piping inspections will be performed during future outages.
NRC - As previously mentioned, an NRC AIT was sent to the Surry facility on:
E cember 9, 1986. The AIT conducted inspections during the remainder of the -
week ending December 12, 1986, to ascertain the circumstances involved in the 1 accident. An executive summary was transmitted to the Region II' office on .1 December 17, 1986. This summary provided the significant facts concerning the event. The AIT did not conclude its inspection at that time due to the ongoing activities by the licensee to develop a root cause analysis, which required subsequent inspection activities.
AIT activities continued during the weeks of December 22 and 29,1986, and January 12,1987. An AIT-exit meeting with plant management was held on
. January 14, 1987, after review of the licensee's investigative report (Ref 8) and proposed corrective actions which were presented to the NRC on January 12, ,
1987. In addition to the AIT inspection activities, inspectors knowledgeable i in security, fire protection systems, water chemistry and check valve design were assigned to review specific concerns in these areas. Where applicable, their bspection findings were incorporated into the AIT report.
The AIT Inspection Report was forwarded to the licensee on February 10, 1987 (Ref. 9). The forwarding letter stated that the AIT concurred in the licensee's planned actions in order to restart the plant. The forwarding letter also included a Notice of Violation regarding maintenance procedures for overhauling a main steam trip valve. The violation was classified as Severity Level IV (on ;
a scale in which Levels I and V are the most and least significant, respectively).
On December 16, 1986, tion and Enforcement Information the NRC Notice No.86-106 ("Feedwater issued Line Break Insy) to all nuclear power reactor facil-ities holding an operating license or a construction permit (Ref.10). The Notice alerted addressees of a potentially generic problem with feedwater pipe thinning and other problems related to the Surry Unit 2 event. Supplement 1 to that Notice, issued February 13 1987, provided additional. information about 4 thinning of piping walls which led to tie pipe break:(Ref.11). Supplement 2 to the Notice, issued March 18, 1987, arovided information about potentially 1 generic systems interaction problems t1at were caused by the release of large' quantities of feedwater (Ref.12).
This item is considered clos.d for the purposes of this report. <
12 i A
FUEL CYCLE FACILITIES (OtherThanNuclearPowerPlants)
The NRC s reviewi events reported by these licensees durir, endar quarter of 1 determined that As .
anyof the date ofwere events this report abnormalhad occurrences.
not , the
[;
AA A A AAAAAA OTHER NRC LICENSEES i (Industrial Radiographer, Medical Institutions, ;-
Industrial Users, etc.)
There are currently more than 8,000 NRC nuclear material licenses in effect in '!
the United States, principally for use of radioisotopes in the medical, indus- ;l trial, and academic fields. Incidents were reported in this category from I
licensees such as radiographer, medical' institutions, and byproduct material users.
The NRC is reviewing _ events reported by these licensees during the fourth cal-endar quarter of 1986. As of the date of.this report
- t. hat the- following events were abnormal occurrences. , the NRC had determined-i 1
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86-24 Therapeutic Medical Misachsinistration The following information pertaining to this event is also being reported con-currently in the Federal Reaister. 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 - On October 6-8, 1986, a patient at'the Cleveland Clinic Foun-dation, Cleveland, Ohio, received a series of cobalt-60 therapeutic radiation exposures which resulted in a radiation exposure that was about 67 percent greater than the prescribed dosage.
Nature and Probable Consequences - A 58 year-old female patient received two radiation treatments a day for three consecutive days, October 6-8 1986, for treate.ent of bone marrow disease. Because of an error in calculating treatment time, these treatments resulted in the patient receiving a radiation dose of approximately 2,000 rads head-to-waist, as opposed to the intended 1,200 rads.
16 e -
The patient was discharged from the hospital on October 10,1986, but was re-admitted on October 20, 1986, for symptoms believed to result from the radi--
ation exposure (unable .to swallow, fever, and chills). .She was discharged af ter trlatsent, but later admitted to Cleveland Metropolitan Hospital. Burn Clinic on November 10, 1986, with skin burns. The patient died.on' November 18,-
1986. ,
The licensee did.not discover the therapeutic treatment error until November 11, !
1986, when a dosimetrist reviewed the patient's treatment records and checked l the calculations. NRC regulations stipulate that such misadministration' be reported to the.NRC within 24-hours after they are discovered; however, the:
licensee did not report it to the NRC until November 17, 1986. . The delay was apparently due to the licensee's failure to realize that a misadministration of this type requires immediate notification.
A panel of HRC medical consultants reviewed the case and concluded th'at the radiation treatments had " minimal effect, if any, upon'the fatal outcome of her disease." The skin burns were not attributable to the radiation treatment, but rather. to a variety of drugs (i.e. ,-chemotherapy) given to the patient prior to and in addition to her radiation treatments.
Cause or Causes - The misadministration was caused by an error in the calcula-tions performed to determine the exposure time to deliver the desired radiation dosage. The physicist who performed the calculations used the distance from the cobalt-60 radiation source to the patient, instead.of the distance from the exterior of the radiation therapy device to the patient. The physicist ;
entered the measurement into a programmable calculator that already accounted for the internal distance from the radiation source to the exterior.of the de-vice. Therefore, the internal distance was added twice with the result that a longer treatment time was scheduled. (The further the source is from the pat'.e nt, the longer the treat. ment time required. )'
In 1982 Cleveland Clinic adopted new procedures as a result of a therapeutic misadministration at that time. These new procedures included a system of dual verification of all dose calculations prior to the first day.of treatment.
In this case, however, the procedure was not followed and there was no recheck of the physicist's calculations prior to treatment.
Actions Taken to Prevent Recurrence Licensee - The licensee has adopted revisions to .its procedures providing that all dose calculations will be independently performed by two qualified individ-uals and that, prior to the~ first treatment, the technologist will . verify that the duplicate calculations have been performed. In addition, the treatment data will be reviewed weekly by the chief technologist. A quality assurance audit by the licensee's Radioisotope Committee is to be performed quarterly _for a year and then annually thereafter.
NRC - On November 20, 1986, NRC Region III issued a Confirmatory Action Letter .
documenting the licensee's agreement to institute the improvements in its procedures listed above (Ref.16).
17
.I i
i special NRC inspection was conducted beginning November 20, 1986 (Ref. 17).
The inspection identified two violations of NRC requirements, i.e., failure to report the therapeutic misadministration within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and failure to obtain approval of the licensee's Radioisotope Committee for physicians to use NRC-licensed materials. (This second violation is not directly related to the misadministration.) Enforcement action on these violations is pending.
The NRC retained a special medical panel to review the case, consisting of two physicians and a physicist. As previously mentioned, the panel concluded that the patient's deterioration of the misadministration. g condition, ending in her death, was not the result This item is considered closed for the purposes of this report.
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86-26 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 arc Place - Or C:teber 21,19M, a patient at St. Luke's Hospital, Racine, Wisconsin, receiveo a whole body iodine-131 ciagnostic scan while the intended procedure was to be a thyroid scan.
Nature and Probable Consequences - On October 6,1986, a patient received a diagnostic thyroid scan using iodine-123, an accelerator-produced radioisotope (accelerator produced radioisotopes are not subject to NRC regul6 tion, but are under State jurisdiction). The attending physician then gave oral instructions for an iodine-131 scan because the previous scan was not definitive. The nuclear medicine technologist erroneously arranged for a whole body scan instead of a thyroid scan as intended by the physician. The whole body scan involved 1.53 mil 11 curies of iodine-131, which is approximately 30 times the normal 50 microcurie dosage for a thyroid scan.
After the scan was performed on October 21, 1986, the attending physician dis-covered the error. The whole body scan, however, did provide the physician with the diagnostic information desired.
The radiation exposure, while in excess of that intended, did not result in any immediate medical effects, according to the licensee. Had a typical dosage of iodine-131 for a therapeutic procedure been administered (i.e., 4 to 6 milli-curies), rather than the 1.53 mil 11 curies actually administered, a significant reduction in thyroid activity could have resulted. Thyroid damage, however, can be compensated for through the use of medication.
20
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1 C$'use or Causes - The misadministration was caused by the medical;technolog'ist's misinterpretating the attending physician's oral instruction.- The physician requested an " iodine-131 scan," which the_ technologist incorrectly assumed to be whole4 ody scan. Typically, the licensee uses iodine-123 for' thyroid scans and iodine-131 for either thyroid scans or whole body scans.
Actions Taken To Prevent Recurrence Licensee - The licensee has revised its procedures for prescribing radioiodine for medical procedures and provided training on the revised procedures. All; prescriptions are now to be in written form and will be reviewed by a nuclear medicine physician and verified by the_ technologist prior to administration off the radiopharmaceutical to_the patient.
NRC - The NRC conducted a special inspection on December 15, 1986, to review the circumstances of the misadministration (Ref. 21).' The inspection did not-identify any violations of NRC. requirements,'but determined.that improvements were needed.in the patient prescription process to preclude similar misadmini- i strations in the future.
NRC Region III issued a Confirmatory Action Letter on _ January 9,1987- (Ref. 22),
documenting the licensee's agreement to change its procedures.' The changes will be incorporated into the facility's NRC license.
The NRC also retained a medical consult' ant to evaluate the misadministration and its possible medical effects. The consultant's report is pending.
- 1. Unless new, significant information becomes available, this item is considered closed for the purposes of this report.
- x x aaaa===*
l 86-27 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 - On November 18, 1986, a patient at Toledo Hospital, Toledo,
, Ohio, received a misadministration of a radiopharmaceutical when the' wrong radio-l active material was. administered. It is estimated that the patient's thyroid received a dose of about 6,760 rads.
Nature and Probable Consequences - The physician of a 62-year ol'd female' patient planned a bone scan for the patient.as an outpatient at the Diagnostic Center.
at Toledo Hospital. The bone scan normally involves a 20 millicurie dose of technetium-99m MDP. , The hospital's procedures provide that the. referring.
physician's office notify the Diagnostic Center by telephone of the scheduled procedure. The procedure is then scheduled, .and the hospital's nuclear medicine department is notified to order the radiopharmaceutical.
21 l
L
In this instance, the physician's office notified the Diagnostic Center, but kept no record of the telephone conversation. The intended procedure was a bone.sca% but the Center's receptionist recorded a " total body scan, rule. out metastases, carcinoma."' This was interpreted by the nuclear medicine depart-ment as an order for a thyroid metastatic disease scan, which is also known as a " total body scan." Toledo Hospital. normally uses a 20 millicurie dose of iodine-131 for such a procedure, which is usually performed on patients.who.
have had their thyroid removed. (The' organ principally affected by an iodine.
dose is the thyroid.) The nuclear medicine department confirmed with the Cen-ter's receptionist.that the thyroid metastatic disease scan was the prescribed procedure. The receptionist, however, did not verify the procedure with the referring physician's office.
On November 18,1986, the pat'ient was administered the iodine-131. .She. returned ;
to the Diagnostic Center the .following day and said.she was ~ scheduled for. a. bone.
scan. Since the Center had no bone scan' scheduled, the error was consequently discovered.
q The patient had previously been diagnosed as having mild hypothyroidism (under- i active thyroid) and was taking medication to make up for the decreased thyroid function. The. iodine-131' dosage was estimated to cause a 6,760 rad dose to the.
thyroid, while other organs received a relatively small dose. (Aradisa standard measure of absorbed dose.) This dosage to the thyroid is less than' would nomally be expected for 20 millicuries of. iodine-131, because of the patient's reduced thyroid function. If the patient.had had a nomally func-tioning thyroid, the expected dosage would have been three to seven' times what this patient is estimated to have actually received. 1 Nevertheless, the 6,760. rad thyroid dose is expected to significantly decrease the patient's thyroid function, riecessitating an increase in the medication .j (thyroxin) the patient was already receiving. The prescribed thyroxin dosage was increased to three times the original prescribed dose. Both the hospital and the patient's physician plan to continue to monitor the patient.
Cause or Causes - The apparent cause of the misadministration was a failure to accurately communicate the prescribed procedure to the hospital's Diagnostic Center. The precise method of failure could not be determined since the pa-tient's physician did not have a record of the telephone conversation in which j the procedure was scheduled. '
i Actions Taken to Prevent Recurrence Licensee - The hospital has instituted a change in its procedures for. scheduling outpatient diagnostic doses. All prescriptions for nuclear' medicine: procedures are to be in written form and reviewed by a nuclear medicine physician and veri- >
fled by a technologist prior to the administration.'of. the radiopharmaceutical to' the patient..
NRC - NRC Region III conducted a special inspection 'at Toledo Hospital on F6vember 25,1986, to review the circumstances of the misadministration (Ref. 23). -(
No violations of NRC requirements were found during the inspection. NRC Re- '
.gion III issued a Confirmatory Action Letter to the hospital on November 21, i
1986, documenting the hospital's agreement to change its procedures for 22
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s[heduling procedures involving radiopharmaceuticals (Ref. 24). The NRC also 1
retained a medical consultant to review the possible health effects of the misadministration.
N .
Unless new, significant information becomes available, this item is conside' red closed for the purposes of this report.
A A A A A A'* AA A 86-28 Immediately Effective Order Modifying License and Order to Show Cause Issued to an Industrial-Radiography Company The following information pertaining to this event is also being reported con-currently in the Federal Register. Appendix A (see the third general suberiter-ion) of this report notes that major deficiencies in management controls for licensed facilities or material can be considered an abnormal occurrence.
Date and Place - On December 30,1986, the NRC issued an Order to Met-Chen Test-'
.ing Laboratories of Utah, Inc. of Salt Lake City that in effect prohibits the l
l company from involving a senior management employee in the performance or supervision of any NRC licensed activities (Ref. 25).
Background - The licensee 'is the holder of both a general license pursuant to 10 CJR 6150.20 and a specific license (License No. 43-26821-01)' pursuant to l 10 CFR Part 30, issued by the NRC. The general license authorizes the licensee '
to conduct the same activity in non-Agreement States pursuant to the provisions of 10 CFR 6150.20 as the licensee is authorized to conduct by its' specific license from the State of Utah, an Agreement State. The NRC specific license authorizes the licensee to use the licensed materials in industrial radiography -
and replacement of sources, and to use an EON Model 64-764 calibrator (which contains a radioactive source) for caHbration cf survey instruments at locations where NRC maintains jurisdiction. The NRC license for industrial radiography was issued in July 1986.
Nature and Probable Consequences - The NRC Order was issued to remove the senior vice president from any assignment or position influencing or involving the performance or supervision of any licensed activities. This action was taken following an NRC investigation initiated in 1985 as a result of inspector obser-wations made during a routine inspection. The NRC decided to issue the Order j after an NRC investigator obtained a sworn statement on August 21,- 1986, from l the senior vice president in which he admitted that while employed as the office manager for the predecessor radiography company (Met-Chen Engineering Labora-tories, Inc.), he had typed a letter and forged on it the signature of a radio- '
grapher for the purpose of explaining away an overexposure indicated on the radiographer's film badge.
The overexposure, while not clinically significant, was reportable according to NRC regulations. The letter falsely stated that the radiographer's dosimeter and film badge were left in a shirt' pocket and the shirt was placed in an area near a radiation source resulting in an overexposure reading, but not an overexposure to the radiographer himself.
1 23
___x_-_ _ _ _ _ - . .
I Had the NRC been provided with correct information,. inspection actions regarding the overexposure would have been taken against Met-Chem Engineering Laboratory, Inc.. . the%ow defunct former coigwy. .Further, had the NRC known that a senior management employee of the licensee had withheld reportable.information concern-ing radiation exposures, the specific . license for the present company would not have been issued. The false statements made by the . senior vice president. call into question.his candor in deeiing with the NRC, and demonstrate that there was no longer reasonable assurance that the licensee would comply with NRC requirements while the individual was involved in licensed activities.
Cause or Causes - The employee willfully made false statements. to, and withheld )
information from, the NRC. On August 13, 1986, the employee denied to an NRC inspector and an NRC investigator any knowledge of how the forged letter was generated. However, on August 21, 1986, he admitted that he had indeed ,
generated, and signed, the letter. i The employee has stated that the reasons he wrote' the forged letter were (1) he did not want anything to stop the sale of certain Met-Chem Engineering Labora-
, tories, Inc. properties to a third party, and (2) he did not'want the NRC to -
know about the overexposure since it would not have been desirable to have.the NRC looking into the matter during the sale negotiation period.
l Actions Taken to Prevent Recurrence
- j Licensee - The licensee responded to the NRC Order on January 15, 1987. The i licensee stated that the employee terminated employment at Met-Chen Testing Laboratories during November 1986, to accept employment with a company which neither has a radioactive materials license nor handles any rautoactive materials.
The licensee held meetings with all authorized users of radioactive materials to restate the instructions they are given during training, which includes total compliance to NRC requirements and to be honest and cooperate totally with NRC personnel, On January 5,1986, the authorized users of radioactive materials signed a statement that they have read and understand the December 31, 1986 NRC Order.
NRC - The NRC Order contained the following provisions, effective imediately:
(1) License No. 43-26821-01 is amended by adding the following condition:
The employee shall be removed from any assignment or position influencing 3 or involving the performance or supervision of any licensed activities 1 (e.g., as an authorized user), including the supervision of any Radiation 4 Safety Officer (RS0).
l 1
(2) The licensee shall show cause in the manner hereinafter provided why the license amendment set out in' paragraph (1) above should not becce parmanent. 1 I
(3) The employee shall be removed from any assignment or position influencing )
or involving the performance or supervision of any licensed activities i pennitted under the general license issued pursuant to 10 CFR (150.20.
24
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(4)' The licen'see shall show cause in the manner hereinafter provided why.thef provisions in' paragraph (3) above should not become permanent.
'(5)' Prior to conducting any licensed' activities after receipt of-this Order, the licensee shall (a) notify in writing all personnel involved in the performance and. supervision of licensed activities at. Met-Chem Testing Laboratories of Utah,-Inc. of this Order and the importance of strict '
adherence to NRC requirements and complete candor with NRC personnel, and (b): certify-to the NRC that each Authorized User and RSO has. read the notification and Order and understands-its' contents.
(6) The NRC Region. IV Regional Administrator may relax or rescind any.of'the:
above provisions for good cause shown by the licensee.'
The.NRC is evaluating the licensee's response to the Order, to determine'whether-it is satisfactory, and/or whether further enforcement -action. is required.
. Future reports will be made as appropriate.
=aaannaaaa AGREEMENT STATE LICENSEES Procedures have. been developed for the. Agreement States to screen unscheduled incidents or events using the same criteria as the NRC (See' Appendix A).and.
report the events to the NRC for inclusion in this report. -During the fourth calendar quarter of 1986, the Agreement States reported no abnormal occurrences to the NRC.
^ '
25 LJ
I
.f REFERENCES
- 1. Letter from Virgil L. Brownlee,' Chief,' Reactor Projects Branch 3. Division of. Reactor Projects. NRC Region II,: to H. B. Tucker, Vice President, .
Nuclear Production Department, Duke Power Company, forwarding Inspection Report Nos. 50-269/86-26, 50-270/86-26 and 50-287/86-26,: Docket Nos. 50-269, 270, and 287; October 23, 1986.*-
2.
~
Letter from J. Nelson. Grace, Regional Administrator, NRC Region LII. to 3 H. B. Tucker, Vice President . Nuclear Production Department, Duke Power-
{
Company, forwarding Inspection Report Nos. 50-269/86-33,_50-270/86-33, and 50-287/86-33, Docket Nos. 50-269, 270, and 287, December 1,1986.*
- 3. Letter from J. Nelson Grace, Regional Administrator,;NRC Region II, to H. B. Tucker, Vice President, Nuclear Production Department, Duke Power-i Company, forwarding " Meeting Summary - Report Nos. 50-269/87-07, 50-270/87-07, and 50-287/87-07," Docket Nos. 50-267,'270, and 287, j
February 11, 1987.* 1 i
4.
I Letter from J. Nelson Grace, Regional Administrator, NRC Region II, to- }
H. B. Tucker, Vice President, Nuclear Production Depart'nent Duke Power i l
Company, forwarding a Notice of Violation, Docket Nos. 50-269.-50-270, 1 and 50-287, February 5, 1987.*
l
- 5. \
U.S. Nuclear Regulatory Commission, Inspection and Enforcement Information (
Notice No. 87-06, "Less of Station to Low Pressure Service Water System Pumps Resulting from Loss of Siphon," January 30, 1987 * {
]
- 6. U.S. Nuclear Regulatory Commission, Inspection and Enforcement Bulletin No. 85-03, " Motor-Operated Valve Common Mode Failures During Plant Tran- I 1
sients Due to Improper Switch Settings " November 15, 1985.*
i 7. U.S. Nuclear Regulatory Commission, Inspection and Enforcement Information Notice No. 86-93 " Inspection and Enforcement Bulletin No. 85-03 Evaluation of Motor Operators identifies Improper Torque Switch Settings," November 3, 1986.*
l
- 8. Letter from R. F. Saunders, Station Manager Virginia Electric and Power:
Company, to U.S. Nuclear Regulatory Commission, Document Control Desk, forwarding (1)LicenseeEventReportNo. 86-20-01; and (2) "Surry Unit 2, Reactor Trip and Feedwater Pipe._ Failure Report " Revision No. O, dated January 14, 1987;' Docket No. 50-281 January 14, 1987.*
l 1
1
- Available in NRC Public Document Room.1717 H Street, NW, Washington, DC 20555, for inspection and copying (for a fee). i l
27 y
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- 9. Letter from J. Nelson Grace. Regional Administrator NRC Region II, to W. L. Stewart, Vice President, Nuclear Operations, Virginia Electric and PoweNCompany, forwarding (1) Notice of Violation; and (2) NRC Augmented Inspection Team Report Nos. 50-280/86-42 and 50-281/86-42; Docket Nos. 50-280 and 50-281, February 10, 1987.* l
- 10. U.S. Nuclear Regulatory Comission Inspection and Enforcement Information Notice No.86-106, "Feedwater Line Break," December 16, 1986.*
- 11. Supplement 1 to Infonnation Notice No.86-106 issued February 13, 1987.*
- 12. Supplement 2 to Information Notice No.86-106 issued March 18, 1987.*
l l
1 l
- 16. Confirmatory Action Letter from James G. Keppler, Regional Administrator, NRC Region III, to Frank Thomas, M.D., Cleveland Clinic Foundation, License No. 34-00466-02, Docket No. 30-00394, November 20, 1986.*
- 17. Letter from Jack A. Hind Director, Division of Radiation Safety and Safeguards, NRC Region III, to William S. Kiser, Chairman. Board of Governors, Cleveland Clinic Foundation, forwarding Inspection Report No. 86-01, Docket No. 30-00394, February 12, 1987.*
- Available in NRC Public Document Room 1717 H Street, NW, Washington, DC 20555, for inspection and copying (for a fee).
28
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- 21. Letter from W. L. Axelson, Chief, Nuclear Materials Safety Section 2. NRC-Region III, to Ray L. Dilulio.- President, .St. Luke's Hospital, forwarding .
. Inspection Report No. 86-01 . Docket No. 030-03425, January 15, 1987.*
i
- 22. Confirmatory Action Letter from James G.' Keppler, Regional Administrator, I NRC Region III, to Ray L. 011u11o President St. Luke's Hospital, License ;
- No. 48 02096-01, Docket'No. 030-03425, January 9, 1987.*- l 4
1
- 23. Letter from W. L ' Axelson, Chief, Nuclear Materials Safety and Safeguards.
Branch, NRC Region III, to William Jeffries Radiation Safety Officer, l Toledo Hospital, forwarding NRC Inspectidn Report No. 86-01, Docket !
No. 030-02685, January 9, 1987.* !
- l
- 24. Confirmatory Action Letter from James. G. Keppler, Regional Administrator. .
NRC Region III, to William F. Jeffries, Radiation Safety Officer, Toledo l 1
Hospital, License No. 34-01710-05, Docket No.- 030-02685, November 21, .i 1986.*
- 25. Letter from James M. Taylor, Director.- NRC Office of Inspection and Enforcement, to J. W. Lewis, Chairman of the. Board, Met-Chem Testing
. Laboratories of Utah,-Inc., forwarding an Order Modifying License and Order -
to Show Cause (Effective Imediately), Licensee lNo. 43-26821-01, Docket ~
No. 30-29056, December 30, 1986.* ']
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- Available in NRC Public Document Room,1717 H Street, NW, Washington 00 !
20555, for inspection and copying (for a fee).
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APPENDIX A N ABNORMAL OCCURRENCE CRITERIA The following criteria for this report's abnormal occurrence determinations were set forth in an NRC policy statement published in the Federal Register on i February 24,1977 (Vol. 42, No. 37, pages 10950-10952).
Ar. event will. be considered an abnormal occurrence if it involves a major re-duction in the degree of protection of the public health'or safety. Such an event would involve a moderate or more severe impact on the public health or safety and could include but need not be limited to:
- 1. Moderate exposure to, or release of, radioactive material licensed by or otherwise regulated by the Commission;
- 2. Major degradation of essential safety-related equipment; or !
- 3. Major deficiencies in design, construction, use of, or management controls for licensed facilities or material.
Examples of the types of events that are evaluated in detail using these crite--
ria 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 cf the feet, ankles, hands or forearms ofanyindividualto375remsormoreofradiation(10CFR620.403(a)(1)),
or equivalent exposures from internal sources. .
- 2. An exposure to an individual in an unrestricted area such that the whole-bodydosereceivedexceeds0.5reminonecalendaryear(10CFR620.105(a)).
- 3. The release of radioactive material to an unrestricted area 'in concentra- l 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 520.403(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 stem per hour three feet from the surface of a package containing the radioactive material, or (b) release of radioactive mate-rial 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.
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Any substantia' ed loss of special nuclear material or any substantiated.-
t inventory. discrepancy.which is judged to be significant relative to nor-mally 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 thef t, diversion, or sabotage.:
-9. An accidental criticality (10 CFR $70.52(a)).
10.
A major deficiency in design, construction, or operation having safety j
implications requiring immediate. remedial action.
IL Serious deficiency in management or procedural controls in major areas.- !
12.
Series of events -(where individual events are not of ' major importance), 1 I
recurring incidents, and , incidents. with' implications for similar facili- !
ties (generic incidents), which create major safety concern.
For Commercial Nuclear Power Plants L Exceeding a safety limit of license technical specifications-3 i
H (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 p:stulated 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. i l
S.
Personnel error or procedural deficiencies which result in loss of plant capability to perform essential safety functions such that'a potential 1 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 exceeded and a plant shutdown is required (10 CFR $50.36(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.
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APPENDIX B 1 y . UPDATE OF PREVIOUSLY REPORTED' ABNORMAL OCCURRENCES During the October through December 1986, period, the NRC, NRC licensees, ' Agree-ment States, Agreement State Licensees, and other involved parties, such as reactor vendors and architects and cagineers, 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. l Those occurrences not now considered closed will be discussed in subsequent '
reports in the series.
NUCLEAR POWER PLANTS i 77-9 Environmental Qualification of Safety-Related Electrical Equipment Inside Containment This abnormal occurrence was originally reported in NUREG-0090-10. " Report 'to Congress on Abnormal Occurrences: October - December,1977" and updated in subsequent reports in this . series, i.e. , NUREG-0090, Vol.1, No.1; Vol.1, No. 2; Vol. 2, No. 2; Vol. 3, No. 2; Voi. 4, No. 2; Vol. 5, No. 2. Vol. 6, j It is being reopened to report the k.1; and closed out in Vol. 8, No. 2.
following new information; f.he information is current as of the end of 1986.
\
The NRC is currently conducting an Environmental Qualification (EQ) Inspection 1 Program which consists of inspections at utility engineering offices and nuclear plant sites to evaluate implementation of EQ programs in compliance with ;
10 CFR $50.49. The program also includes inspections at EQ test laboratories and qualified eovipment manuf acturers' test facilities te evaluate-test methods and practices, and verify compliance with quality assurance, EQ technical, and other regulatory requirements. 3 During the course of this inspection series, the NRC discovered a number of-instances in which heat-shrink electrical insulation sleeving manufactured by Raychem, the principal supplier of this type of nuclear environmentally qua11fied-insulation systes, was improperly installed. The sleeving is primarily used to insulate and environmentally seal electrical connections which may be subjected to a harsh environment during postulated plant accidents such as. loss of coolant' accidents (LOCA) or high energy. line breaks (NEL8). Rachen sleeving was found -
installed in configurations which deviated from the configurations.which had -
been tested and qualified. The improper installations were found in use with various qualified components in potentially harsh environmental zones both inside.
and outside containment. In most instances, the deficiencies would not be ex-pected to cause equipment failure under' accident conditions. .However, the full-
~
extent of safety systes degradation could not be accurately. assessed.
As a result of these discoveries, the NRC issued Inspection and Enforcement Information Notice No. 86-53 on June 26~, 1986, (Ref. B-1). which alerted remain-ing plants of the problem. Corrective actions are in progress or have been completed at plants where Raychem deficiencies have been identified and the EQ inspection program routinely examines this issue to verify that the utilities 33
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are effectively conducting their own inspections and taking appropriate correc-tive actions. Additionally, the NRC has issued a Temporary Instruction for use-by NRC regional inspectors and resident inspectors to expedite verification that all plants are finding and correcting Lthese deficiencies.
In an effort to reduce the number of required repairs, and to show that most installed splice configurations are qualifiable as they are, several utilities have sponsored additional EQ testing. This testing has qualified a much wider i range of splice configurations than was originally qualified by Raychem and has significantly reduced the scope of the problem. Pursuit of this issue will continue until resolved at all plants and effective measures to prevent recurrence are implemented.
During the EQ inspection at Commonwealth Edison Company's (CECO) Dresden Nuclear- i Power Station, NRC inspectors found that General Electric type F-01 Electrical ~
Penetration Assemblies contained wire butt: splices using insulated crimp connec-tors made by the Amp Company. The splice insulation sleeves were made of nylon material with open ends which did not seal tightly to the wire insulation.
Prompted by NRC questioning of the applicability of CECO's.EQ test documentation to the installed configurations of the splices, CECO sponsored additional EQ testing to qualify the splices for their applications. In these tests, the Amp connectors failed by causing short circuits due to degradation of the nylon insulation.- The tests indicated that these open end nylon insulated Amp butt splice connectors were unqualified for their identified harsh environmental applications.
Connectors of this type were found in similar applications at CECO's Quad Cities Nuclear Power Station and Iowa Electric's Duane Arnold Energy Center. In all cases, licensees have taken prompt action in effecting repairs using qualified tape or heat-shrink sleeving irsulation and sealing methods. To alert other licensees to this problem, NRC issued Inspection and Enforcement Information Notice No.86-104 on December 16, 1986 (Ref. B-2). The actions of licensees in response to the Information Notice will be monitored by regional and resi-dent inspectors and will be addressed during the future inspections in the ongoing EQ Inspection Program.
Unless new, significant information becomes available, this item is considered closed for the purposes of this report.
1 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 Abnomal occurrences: January-March 1979," and updated '
in each subsequent report in this series, i.e., NUREG-0090, Vol. 2. No. 2 through Vol . 9, No. 3. It is further updated for this report period as follows.
Reactor Building Entries During the fourth calendar quarter of 1986, 88 entries were made into the TMI-2 a reactor building, bringing the total number of entries since the March 1979 accident to 1135. Entries are currently made by several crews each day to per-form defueling and supporting tasks. Specific reactor building activities 34
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conducted during this period included core drilling using the core boring machine l as a-defueling tool,. mapping and-video examinations-of the core debris bed, q removal of end fittings and broken drill. strings', robotic decontamination of. 1 basement Wils, and efforts to. improve water clarity in the reactor coolant: .,
system. j l
Reactor Vessel Defuelina Operations In October 1986, the drilling equipment that had been used earlier for core sampling activities'was modified and reinstalled on the defueling platform.- l Full-scale drilling operations. were conducted between late October and . late -
November. A total of 425 holes were. drilled into.the hard crust region of ~ the core to condition this' material for subsequent removal. The entire cross ~sec-tional area of the core, except for a 2 foot ring at the periphery,;was drilled to a depth of between 14 and 4 feet. While this effort was- successful in' break-i ing up the hard crust, the subsequent removal of: the resulting debris proved j difficult, due to the number of relatively large pieces of core material-' remain-ing in the vessel. Consequently, defueling progress during'the quarter was limited. Through December 1986, approximately 61,000 pounds (20%) of the core debris had been removed from the reactor vessel.
In late November 1986, the licensee, General Public Utilities Nuclear Corporation (GPUNC), performed topographic mapping and video examinations of the core debris ~
bed to develop more effective methods of removing the core debris generated by j the drilling operation. Concurrent with these' activities,'GPUNC conducted I numerous tests on coagulant additives intended to improve filter performance in l the defueling water cleanup system (DWCS). By early January 1987,'an effective )
coagulant compound was identified .and was added to the reactor vessel water tof i l be used in conjunction with a diatomaceous earth filter aid in the DWCS. This system has greatly improved the DWCS filter performance and is expected to main-tain good water clarity for the duration- of defueling activities. Additionally, new tooling and techniques appear to be ef fective for . removal of the larger pieces of core debris. Through the end of January 1987, 72,000 pounds (24%) of 1 core debris had been removed from the reactor vessel, an indication that the I pace of defueling has increased. The licensee projects that debris removal i from the reactor vessel will be completed by December 1987.
Decontamination and Waste Disposal Activities In early December 1986, GPUNC conducted a decontamination experiment in ,the reactor building basement. A high pressure hydrolazer mounted on a robotic vehicle was used to scarify the concrete basement walls. The licensee is i' continuing to examine the use of robotics technology.in decontamination applications.
During the fourth quarter of 1986, the licensee continued to apply proven decon-tamination techniques in the auxiliary and fuel handling building (AFHB). By year's end, decontamination of approximately 66% of. the contaminated surface area in the AFle had been completed. The licensee expects to complete AFHB,
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decontamination during CY 1987.
During the fourth quarter of 1986, 10 EPICOR-II spent resin liners were shipped '.
offsite to Hanford, WA for disposal as low-level waste.
35
In December 1986 and January 1987, two additional shipments of core debris were transported by rail to the Idaho National Engineering Laboratory (INEL) for examination and interim storage. A total of.48,000 pounds of core debris has been shipped to INEL to date.
In December 1986, the NRC staff issued, for public comment, draft Supplement i i
No. 2 to NUREG-0683, " Programmatic Environmental Impact Statement (PEIS) for i the decontamination of TMI-2" (Ref. B-3), which deals with the disposal of slightly contaminated accident generated water (AGW). Ten alternative methods ;
of disposal were evaluated.in some ' detail in the draf t supplement, including 4 the licensee's proposal for forced evaporation of the water with disposal of the solidified residue at a low-level waste burial site. . The potential environ .
mental impacts for all ten alternatives evaluated were determined to be very small. Following receipt of comments on the draft supplement, the NRC staff will issue a final supplement to the PEIS and provide a recommendation to the l Commission. The Commission will ultimately approve the method of disposal .for j
the AGW. I TMI-2 Advisory Panel Meetings The Advisory Panel for the. Decontamination of THI-2 met on October 8 and-December 10, 1986, in Harrisburg, Pennsylvania. At the October meeting, the Panel was briefed by the licensee on the progress of defueling and the status of funding for the remainder of the cleanup. The Environmental Protection l Agency precented the results of its review of the TMI offsite radiation monitor- 1 ing program. At the December meeting, the Panel was briefed by the licensee on their proposal for the Post-Defueling Monitored Storage of THI-2.
Future reports will be made as appropriate.
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I 86-15 Differential Pressure Switch Problem in Safety Systems at LaSalle Facility l This abnormal occurrences was originally reported in NUREG-0090, Vol. 9, No. 3, l
" Report to Congress on Abnormal Occurrences: July-September 1986." It is l updated through December 1986 as follows. -
On Decelaber 18, 1986, Commonwealth Edison Company (the licensee) reported that five SOR, Incorporated (formerly the Static "0" Ring Pressure Switch Company) .
differential pressure switches utilized in safety-related applications had failed l in the previous 14-month period at LaSalle. Prior to this report, the' licensee '
had considered the failures to be " random failur'es". This new report indicates that all five failures were related to failures of the diaphragm in the switches '
(i.e. , pin hole leaks or tears) which raises reliability and design concerns.
Failures that occurred prior to the June 1,1986 event were found during normal surveillance testing. Failures af ter June 1986 have been found during the aug-mented test program initiated following the June 1,1986 event. Systems and components in which failures have occurred include: Unit 2 ADS (automatic de-pressurization system) permissive (Level 3); Unit 2 RCIC (reactor core isolation.
cooling) Line Break Protector; and Unit 1 RHR (residual heat removal) mini-flow 1
35 j
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alare. The root cause of the switch diaphragm failures'is still under investigation by the licensee and the switch manufacturer.
Future reports will be made as appropriate.
- aa FUEL CYCLE FACILITIES ,
y 86-3 Rupture of Uranium Hexafluoride Cylinder and Release of Gases !
This abnormal occurrence, involving Sequoyah Fuels Cor was originally reported in NUREG-0090, Vol. 9, No.1, Report goration, Gore, Oklahoma, to Congress on- l Abnormal Occurrences: . January-March 1986." and updated in subsequent reports 1 in this series, i.e. , NUREG-0090, Vol. 9, No. 2 and Vol. 9, No. 3. Some-other i events involving overfilled uranium hexafluoride cylinders at Allied-Signal:Cor- '
poration's facilities in Metropolis, Illinois were also discussed as an Annex to the abnormal occurrence in NUREG-0090,. Vol. 9, No. '1. - The' abnormal occurrence j and the-Annex are updated through mid-March 1987, as follows. j Sequoyah Fuels Corporation (SFC) 1 As discussed in the previous report (i.e. , NUREG-0090, Vol. 9,L No. 3), on !
November 14, 1986 the NRC authorized SFC to restart production activities. 1 The NRC provided 24-hour oversight of; restart and production activities and continued the requirement for an independent oversight organization to~ provide f a third party examination of operations.
On December 15, 1986, the licensee began filling the first UFs cylinder (14 ton) which was completed on December 17. The remodeled steam chest lwas successfully used on December 22 to heat an out-of-specification (high chromium '
level) UFs cylinder. By later December, SFC.was operating'at a production rate of 300 stu per month. The independent oversight organization released-two reports to SFC staff describing its activities, findings, and recommendations, y
Although minor equipment problems were encountered as each piece of equipment '
was started up, SFC believes that they have not had as many problems' as they originally anticipated. t Continuing into early 1987, no significant safety concerns have been noted, and ;
SFC continued to operate without incident. On Februar 28, 1986, NRC Region IV 1
reduced their oversight activities-from a 24-hour per yday basis to one shift- '
per day, seven days a week. On February 24,1987,. SFC requested a reduction in -
the coverage by the independent oversight organization.' This request is under review by the NRC.
As discussed in NUREG-0090, Vol. 9, No. 3, on October 14', 1986 the NRC' issued' I to SFC a Notice of Violation and Proposed Imposition of Civil Penalties in the -
amount of $310,000 (Ref. B-4). The violations which were directly associated with the January 4,1986 accident were categorized as a Severity. Level .I problem (on a scale in which Levels I and V are the most and the least significant, respectively) and accounted for $300,000 of the proposed civil penalty.
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On November 4
protesting the 13,1986,' the licensee' civil penalties responded and in support by presenting of remission or mitigation a number of argo.nents of ther civil penalties. , The NRC has- reviewed the licensee's response and concluded that the alleged violations;did' accur as stated in the Notice of Violation ap'd '
no mitigation of the civil penalties is warranted. Therefore, on february % ,
1987, the NRC issued to SFC an Order Imposing Civil Monetary Penalties. in the ,, .i amount of $310,000 (Ref. B-5). 'In early March 1987, the licensee paid the civi) penalty in full.
1J Annex Update: Allied-Signal Corporation. N
[h On November 3-7,1986,' representatives of the NRC Office of Nuclear Mate' rial ~
Safety and Safeguards, NRC i Region II, and.the. Illinois Department of Nuclear ~l Safety conducted an announced team inspection at the Allied-Signal facility in' a s Metropolis, Illinois. Theinspection"focusedon'completionofopsnlitemsiden-tified-in previous inspections;' comparison of actions taken at Allied with thoser 'j taken at SFC; and. Allied responses to.the Lessons-Learned Recommendations.t - [Asr s discussed in previous reports, the: Lessons-Learned Recommendations wereddinaloped j
by a Lessons-Learned Group," formed by the NRC Acting Executive Directorifora h Operations on February 20, 1986, with a: goal to' identify actions to preventV i incidents similar to the January 4.-1986 SFC accident and to improve response / '
9 follow-on activities by licensees and regulatory agencies. .The _ Group's report '
was issued during June 1986 as .NUREG-1198 -(Ref. B-6).]f The results of the inspection were discussed with Allied nianagement in an'eIit
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meeting on November 7,1986. The inspection report wss' formally sent to,the e !
licensee on November 18, 1986'(Ref. B-7). The inspectbn identified one apparent l violation (regarding the frequency of monthly.trainik 6f the fire brigade "
l emergency response team)-and various open items to be resolved. The NRC forward-1 ing letter noted that Allied's Hazard Review Committee, established during the , M' first half of 1986, has made considerable progress in identifying and resolving i '
safety issues.
As discussed previously in NUREG-0090, Vol. 9, No.1, on June 27. 1986 - the NRC-forwarded to the licensee a Notice of Violation and Proposed Imposition of Civil Penalty in the amount of $25,000 (Ref. B-8). The. violations included failing to report a December 7,1984 incident (involving overfilling of, and subsequent damage to an uranium hexafluoride cylinitet; however, there were no releases of . is J
gases or personnel injuries involve.1) tFthe NRC, and three instances of failing "
to follow procedures during the March 23,'198 ovsrfill incident.
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On August.21,1986, the licensee forwarded adheck' for $25,000; however, the l' licensee questioned the assignment of Severity Level,III to-thescollective vio-lations, as well as the magnitude of the' proposed tivil penaltyi The NRC re viewed the licensee's response and in a December 19,'1986 lette'r(Ref.'B-9)
concluded tion of the that civil apenalty category wasofwarranted.
Severity Level III.wasLproper and that no.'mitiga-q ij ,
Unless new, significant information becomes. available, .the SFC abnonnal occur-rence, as well as the Allied-Signal Corporation incident',"are considered closed for the purposes of this report. -,
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( r, y g 86-10 Willful Fai_lpre to Report a Diagnostic Medical Misadministration .
b This abnormal; cecurrence was originally reported in NUREG-0090, Vol. 9, No. 2, .
1
" Report to Congress on Abnormal Occurrences:' April-June 1986." It is updated as follrys. ' '
As d cu'ssadinthepreviousreport,onJune17,198fntheNRCiorwardedto 1 Mercy Hospital of Wilkes-Barre, Pennsylvania. (1) ant ordar requiring the licensee j to show cause why the Chief Nuclear Medicine Techni'c fan and the Radiation Safety '
should not be prohibited from the peNermance or supervision of I Officer any licensed (RS0)4ctivities, and (2) a NoMcfof Violation and Proposed 1 Imposition '
of Civif Pent Ity in;the amount of $5,000 (Ref. B-10). The reason was'that.the ,
individuals Willfully: failed to report a diagnostic medical misadministration j totthe NRC as required bf 10 CFR K35.43.' In addition..the R$0 'at Mercy Hospital. i
/#also listed as an authorizd user of NRC licensed material on thetlicense'of- 5
' Valley Radiology Associates. Inc., Kingston, Pennsylvania. Therefore,'on.
June 17, 1986, the NRC issned a similar Order to this licensee-(Ref. B-11).
The ~ technologist invol/ed at Mercy Hospital no longer works in the field'of-Nuclear Medicine. On July 15, 1986, NRC Region I, with the consent of'the licensee, issued an amer,dment to'the Mercy Hospital license which' tid'not. in-clude the RSO involved. While the individual will continue to work at the hos-M pital, his activities will be under the supervision of an authorized user. . He .
1 will not tranage the Nuclear Medicine program and is no longer the RSO. LRegion I has agreed to reconsider adding the individual as an authorized use d if the.
hospital formally requests ~f t, after a period of one year. ;0n October-17,:1986, the licensee sent a check in full payment of the $5,000 civil penalty' On .
December 24, 1986, the NRC found that the . licensee's corrective actions pro-vided adequate cause why the licensee should not be modified (Ref. B-12).
. f' <
On December 15, 1986, the license of Valley Radiologi" Associates,, knc.iwas amended to delete the same individual as an 'authorizeNuser, with the consent.
of.the licensee. On December 24, 1986, the NRC found that the licensee's '
7 corrective actions provided adequate cause why the / license should not be .i j; ' sodified (Ref. B-13).
This' Item is considered c1csid(.ior the purposes of this report.
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i APPENDIX C OTHER EVENTS OF INTEREST The following items are described below because they may possibly be perceived ..
I by the public to be of public health significance. The items did not involve a major reduction in the. level of protection provided for public health or l safety; therefore, they are not reportable as abnormal occurrences. )
- 1. DgselGeneratorProblems During tnis report period, several significant events involving problems with standby diesel generator (D/G) units occurred at various rxlear power plant sites. Stsndby D/G units are provided as a source of electric power for safety-related equipment in the event of an accident situation in which a total loss of offsite power also occurs.) The events are. described below. ,
1 .
1 Gas Leakage into the Jacket Water Cooling System J l During September 1986 it was reported by Baltimore Gas and Electric Company that-diesel engine Na.12 at the Calvert Cliffs Nuclear Power Plant (NPP)* was , ;
experiencing leckage of gasas into the jacket water cooling system (JWCS). 4 l Excess gas in the JWCS can result in cavitation of the jacket water pump, loss
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of cooling for the engine, and ultimately diesel engine failure. The problem 1 was initially observed at a very low level of leakage in 1984. )
After extensive investigation and analysis of the potential sources of leakage, /
including the installation of additional instrumentation and provisions for venting, the licensee in conjunction with vendor representatives identified i adapter penetration seals and a turbocharger'61r. cooler (one of two)'as :he j sources of gas ,inleakage. There are four adapter. penetrations in each cylinder that provide for fuel injection, starting air input, and a spare. A cracked )
cylinder liner was also found. All cylinder liners, adapter seals, and the i faulty air cooler were replaced and the engine was operating satisfactorily as Y of mid-December 1986. I e l The diesel engine manufacturer, Colt /Fdirbanks Morse Engim Division, hat issued I a special manual, " Maintenance and Surveillance Testing Program," for neclear !
l applications (Rev. 1,7/18/86). The manual calls for leak testing adapter seals '
I at intervals no greater than 18 months. The vendor has also reconsnended tha* I the cylinder liners be replaced every 10 years. j zi .j v ,
Failure of a Connecting Rod -
On October 24, 1986, Commonwealth Edison Company reported the failure of a con- !
necting rod in diesel engine No. IB at the Zion NPP.* The connecting rod had ' j
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- 1 been thrown through the inspection door along with oil, water, and various other parts of the engine.
N Prior. to the failure, the diesel engine had been undergoing preventive mainte- !
nance testing. The testin (cylinders 4L,SR,and7L)gshowedslightlylowcompressioninthreecylinders
. To correct this problem, the heads on these three 1
i cylinders were removed, and the pistons, rings, cylinder liners, and gaskets were replaced. The engine was reassembled and post-maintenance testing was initiated. After the engine had operated for approximately one and one-half J
hours at 1000 kilowatts (kW) load (25 percent of full capacity), loud knocking noises were heard that escalated rapidly to massive engine failure. 'I Review of maintenance and reassembly activities revealed that the bolts.that -
fasten the articulated connecting rod to its rod pin were torqued to 690 foot-pounds (f t-lb) instead of the required 1140 ft-lb. Inadequate torquing of these bolts resulted in a gradual loosening and failure of the bolts. ~
The. engine was repaired in place with the assistance of the engine manufacturer ,
and consultants. The engine completed its 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> test run satisfactorily on February 18, 1987, and is available for service. NRC issued two notices of I violation to the licensee as a result of this engine failure.
Failure of Connecting Rods On December 23, 1986, Arizona Public ' Service Company reported a catastrophic _
failure of two connecting rods (rods 9R and 9L) in diesel engine No. 3B at the Palo Verde NPP.*
At the time of the failure, the No. 3B diesel engine had operated about eight minutes of a scheduled two hour rur at 120 percent of full power. The failure resulted in the connecting rods, pistons, counterweight and other parts i being ejected from the engine. Even though the fuel oil supply to the engine p was terminated when the engine was tripped, the combustion process was sustained
'4in the remaining 18 cylinders by the hot lubricating oil mist in the crankcase s
,i mixing with air and drawn into the intake manifold through the now empty 9R and
'9L cylinders.
The engine centinued to run at approximately 250 rpm for about one hour. Opera-tion was finally stopped by the plant fire department by spraying fire suppres-3s ,
sion foam into the lubricating oil sump through the hole in the side of the engine.
Investigation and analysis of the event resulted in the conclusion that the cause of the connecting rod failure was stress concentration at the oil hole between the main'and articulating connecting rod pin bores aggravated by the composition and thickness of the iron plating. The rod pin bores had been over-machined during manufacture and the correct diameters were restored by the use of iron plating of 50 to 60 mils in thickness. Unfortunately, the plating
- Palo Verde NPP is a three Unit facility (as of late March 1987. Unit 3 was not yet operational) located in Maricopa County, Arizona. All three Units utilize Combustion Engineering-designed pressurized water reactors.
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. a e'xtended Linto the oil holes joining the- rod pin bores and was;not removed tol restore the required radius. This, in combination with the. iron' plating's .
columnar grain structure which has-a tendency- for cracking..and the. area of.
stress concentration caused by the oil holes, resulted .in cracking of the iron plating and the parent metal and ultimate catastrophic failure of the engine.
The engine manufacturer, Cooper-Bessemer, reviewed its manufacturing records and determined that a total of four other connecting rod pin bores were over-bored and iron plated in' this manner. - Two (including the failed rod) were installed in Palo Verde engine No. 3B, one was.1nstalled.in Palo Verde engine
.No. 2A, and one was installed at Commonwealth Edison Company's Byron Nuclear.
Power Station."
All of the affected connecting rods were removed and replaced with correctly machined connecting rods. In addition, the' manufacturer has notified the.
licensees for two other plants that their diesel engines contain five connecting.
rods with iron plating but not in areas that are subject to high stress loads.
Four of the five connecting rods are located at Commonwealth Edison Company's Braidwood NPP** and one at Niagara Mohawk Power Company's Nine Mile ' Point Unit 2t. 'The connecting rod has been replaced at Nine Mile Point Unit 2.. - The ~
1 licensee for Braidwood is determining.the corrective-action to be taken.
Cracked Heads on Diesel Engines On October 7,1986, during a station refueling' outage, Nebraska Public Power District discovered while conducting an annual inspection of station diesel No. 2 at Cooper Nuclear Stationtt that.12 of 16 cylinder heads were cracked.
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' The cracked heads were the result of the use of heads originally designed to-accommodate fuel consisting of 94 percent natural gas and 6 percent. diesel fuel oil. The heads contain extra passageways for higher volume gas flow which resulted in a smaller stress margin to withstand engine upset conditions-(e.g.,
primarily engine cooldown and incorrect fuel timing). The heads were of the gas / diesel configuration considered in 1970 when the engines'were manufactured.
As of mid-February 1987 all 16 cylinder heads had been magnaflux tested and hydrostatically tested with the result'that 10 heads.were replaced with heads of a new design, 4 heads were found crack-free and were returned to the engine, ,
and 2 heads were replaced with crack-free heads of the old design. '
- Byron Nuclear Power Station is a two Unit facility located in Ogle County, Illinois. Both Units utilize Westinghouse-designed' pressurized water reactors.
tNine Mile Point is a two Unit facility located in Oswego County, New York.-
Both Units utilize General Electric-designed boiling water' reactors.
ttCooper Nuclear Station is a single Unit facility located in Nemaha County, Nebraska. 'The Unit utilizes a General Electric-designed boiling ' water reactor.
43
. i Station diesel engine No. I was also inspected. All of the cylinder heads' were crack-free and were returned to_ the engine. .The licensee will replace all heads of the old, design for both engines with heads of the new design at the earliest practical opportunity. The manufacturer of these diesel engines Cooper- .
Bessemer, has reviewed its records and has determined that the only other J 1
engines subject to this potential problem are the five engines at the Zion NPP. However, all of the cylinder heads for those engines have already been '
replaced several years ago with heads of the new design..
- 2. NRC Augmented Inspection Team Sent to Hope Creek On September 24, 1986, NRC Region I sent an Augmented Inspection Team (AIT) to the Hope Creek Nuclear Power Plant to investigate numerous equipment problems 3
and test anomalies experienced during loss of offsite power (LOP) tests performed !
on September 11 and 19,1986. Hope Creek, operated by Public Service Electric and Gas Company (the licensee), utilizes a General Electric-designed boiling water reactor. The plant is located in Salem County, New Jersey. The plant first achieved criticality on June 28, 1986 and first generated electricity on August 1,1986. ')
l During-September 1986, the plant was in the power ascension testing program.
An important part of the program is the LOP test. Its purpose is to demonstrate whether the plant response is satisfactory and in accordance with the plant 3 design for concurrent loss of the turbine generator and all offsite power J '
sources.
The LOP test was initiated on September 11, 1986 from about 21.5% power' and l with the turbine generator loaded to 165 MWe. The first indication of an un- I satisfactory plant re'sponse was the failure of the "C" emergency diesel genera-tor output breaker to close automatically. Socn after,. an observed failure of i the reactor auxiliary cooling system coincident with-increasing crywell pres- l sure resulted in the test being aborted by the licensee. Normal offsite power was then manually restored to the station. Twenty-four observations were made by the licensee during this test. These observations occurred during the time i from initiation of the test until the reactor vessel water level and pressure i were controlled and the reactor scram was reset.
The most significant observations on September 11,1986 were: (1) emergency diesel generator "C" output breaker failed to close; (2) main steam relief valve position indication was lost; (3) power supplies for the source and intermediate range neutron detector drives and main steam line acoustic monitors were lost; (4) 17 control rods did not provide a normal full-in position indication; (5) reactor auxiliary cooling system flow was lost; (6) emergency diesel generators "A" and "B" governors transferred isochronous (frequency control) to speed droop (load contro') mode without operator. action; and (7) the "B" safety auxiliary cooling system pump failed to auto-start.
On September 19, 1986 the licensee performed a cold LOP test. The purpose of this test was to demonstrate that the plant response was in'accordance with plant design for loss of all offsite power sources after the licensee had assured that the previous test observations had been investigated and resolved. This LOP test was initiated with the reactor at cold (T 200"F) shutdown temperature and pressure conditions with the reactor mode switcE in shutdown.
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i- '.s-During the September 19, 1986 test,-.the: licensee. identified a total-of'17 ob servations. The most significant observations were: (1) the "B" safety.auxi '
liary. cooling ' system loop head tank level indicator' failed; (2) one control room emergency ventilation (air recirculation) system fan failed to start; and (3) one drywell fan also failed to' start.o As a result of the unsatisfactory test results, an NRC AIT was formed and sent
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.to' the. site to: (1) independently assess the. root' cause of each' observation; (2) review the effectiveness .of the correctiv'e actions. planned or taken;. and
.(3)' assess the 'overall implications of the test;results. The AIT inspection
' began on September 25 and ended October 3,1986.
A 'second cold LOP test was conducted on October 2,1986. . The AIT witnessed this-test and assessed the results. One test observation was a repeat of'a previous observation and involved a Bailey' logic module.
Of the total of 41 observations reported from the September 11 and 19,1986. LOP tests, the overall safety significance was concluded to be relatively minor except for the Bailey solid state logic module failures. These modules,lmanu-factured by the Bailey Meter Co. , are multipurpose electronic ' devices used extensively throughout the plant for control and. safety functions. Of eight:
hardware failures identified. during this review, six were attributable to various malfunctions with Bailey logic modules.
Three weaknesses with the Bailey logic modules .were' found: L(1) the' dependency on common equipment for accomplishment of automatic and manual safety actions for the actuated safety system equipment; (2) limited test' provisionsito assure the online operability of the Bailey logic modules after their installation into the equipment cabinets'; and (3) the usefulness of the bench test equipment in assuring that the Bailey logic modules are operable.. The AIT was also con-cerned that the failure rate cf the Bailey logic modules appeared high. These weaknesses are especially significant since all of the balance'of plant safety .
related systems (and a part of one nuclear steam supply system) use Bailey modules to develop the safety system logic'and actuation functions.
A number of minor plant design,. construction, and manufacturing problems were.
also identified. Several specific weaknesses in the scope of.various system preoperational tests were revealed since the LOP tests were the first integrated demonstration of the plant response to this type of event. Several subtle interactions involving the dependency of various systems on cooling and instru-ment air supporting systems were revealed.
A number of observations resulted because instruments or other equipment-lost power during the test. A number'of these instances involved the apparent failure to meet Final Safety Analysis Report (FSAR) commitments to provide reliable power to specific instruments or equipment.
In summary, the AIT determined that the results of the LOP tests indicated cer- ..
tain weaknesses in the design', construction,' and testing programs for Hope Creek. j With the exception of. the Bailey modules, the AIT found the weaknesses to be .
minor in nature; 'For the Bailey modules, however, the'AIT identified concerns with the adequacy of bench and surveillance testing' and a failure rate which '
was higher than expected.
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- 7. _ - ,
~Fo'11owing the.AIT. inspection, the licensee presented a program to resolve con-cerns related to the Bailey modules. The program will include:
Anln-house data assessment programwhich will include a' review of each in plant module failure 'and. a. determination by the ' manufacturer of the '
individual component which failed and, to the degree possible, the cause of the failure.
An assessment, by the manufacturer, of module failures at installations 'of other. users.
An accelerated aging and cycling test program, with a final reliability analysis report by the end of the second _ quarter of CY 1987.
A monthly trending progrem that will provide a report bi-monthly indicating:
(1) the number of module failures having an adverse affect onLsystem.func-tion; (2) resulting time in a Limiting Condition for Operation; and (3) the number of failures. determined by surveillance. This program will apply to both 1E and non-1E systems.
A report of Bailey's recommendations to improve module reliability based upon their observations at Hope Creek of site environment. handling, and testing techniques.
The modification of existing module test equipment and procedures to permit module testing without staple jumper removal.
The development and procurement of a test rig capable of bench-testing modules for all utilized functions prior toLNovember 1987. Testing.would j.
oe concucted without removing staple jumpers or the field programmable logic array chips.
The determination _of the feasibility and implications of modifying the existing Bailey system to permit in-situ testing.
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This program is underway and the results are under continued NRC review.
The findings of the AIT are contained in NRC Inspection Report No. 50-354/86-50 which was forwarded formally to the licensee on October 31,1986 (Ref. C-1).
The NRC forwarding letter also listed several' issues which the AIT identified as potentially warranting enforcement action in . terms of violations or devia-tions. The issues will be reviewed during future NRC inspections, and if-warranted,' enforcement actions will be taken as' appropriate.
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- 3. Conviction of International Nutronics, Inc. , and One Employee in' Federal District Court International- Nutronics, Inc. , (INI) a California corporation, and Eugene.0'Sullivan, a Corporate Vice. President and Corporate Radiation' Safety Officer of INI,' were convicted on October 29, 1986, in Federal' District Court in Newark, New Jersey. They had been charged with two counts of willful viola- l-tion of the incident notification' requirements of the Atomic Energy. Act, one'-
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count of willfully furnishing false information to a government agency, one count of conspiracy to conceal the incident.from the NRC, and five counts of mail and Wre fraud. Bruce Thomas the Plant Manager Land. Radiation Safety Officer of IN!'s Dover, New Jersey, facility was acquitted on all nine counts.
INI was fined $35,000, the maximum fine. Mr. O'Sullivan was given a suspended
-sentence and two years probation. Both convictions have been appealed.
The charges resulted from a December 1982 incident involving a spill of radio-actively contaminated water at the INI irradiation facility in Dover, New Jersey.
The spill resulted in widespread contamination of the. facility, including the ground immediately under and adjacent to it. Decontamination was begun in 1983 and completed in early 1986. The. facility has been released for unrestricted u use, and the INI license has been terminated at their request.
- 4. NRC Augmented Inspection Team Sent to Hatch' Facility i
On December 4,1986, NRC Region II sent an Augmented Inspection Team to the E.I. Hatch Nuclear Plant to investigate a release of radioactive water from the plant refueling floor into the environment discovered on December 3,1986.
Hatch is a two-unit facility, both of which utilize a General Electric-designed boiling water reactor. The facility is operated by the Georgia Power Company (the licensee) and is located in Appling' County, Georgia.
Prior to December 2,1986, the air regulator (100/20 psi) supplying the inflat- 1 able seals on the seal assembly, which seals the gap between the Unit I and !
Unit 2 reactor. buildings, failed and the lever valve upstream of the regulator was placed in a throttle position to compensate. This throttle almost shut. On December 2,1986, at approximately 10:00 p.m. (position was EST),this throttled valve was closed when a Plant Equipment Operator (PE0),'who was re-storing equipment maintenance clearances on the air system, noticed that the valve was almost shut; assumed it to be slightly out of position in the open direction; and placed'the valve in the closed position. This action secured the pressurization air to all six inflatable seals on the transfer canal seal assembly.
At this time, a slow depressurization of the seals commenced. The inflatable seals, not being entirely leak tight, slowly deflated creating a leak path from !
the refueling floor into the gap between the two reactor buildings and from there to the environment and to areas of Unit I and 2 reactor buildings Unit I and 2 turbine buildings, the control building, the hot machine shop, and the nitrogen storage area.
On December 3,1986 the licensee discovered unidentified water leakage at several i locations in the reactor and turbine buildings and searches were initiated to l locate the source of the leakage. Several low fuel pool level alarms were received during this period but the operating personnel-did not correlate the various water leaks to the' reactor and turbine buildings with the low fuel pool level alarms. (Nonnal fuel pool makeup had been about once per shift prior.to the leak.) When at 9:37 a.m., December 3, the fuel pool cooling pumps tripped on low-low surge tank level, along with the low fuel pool level alarm, operations personnel went to the refueling floor to investigate.
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-The flow past the transfer canal was seen by the operators and the air supply
.was restored to the seals which stopped the leak. Fuel pool level was observed to be about five and one-half feet below normal. No significance change in radiation levels on the refueling floor was observed. Calculations indicate that approximately 140,000 gallons -had leaked from the fuel' pool.- Of this, about 17,500 gallons were collected in the buildings.and processed through the radwaste system. About 80,000 gallons were released to the swamp located on the licensee's property east of the plant. The rest of the. water was absorbed in 'the three and one-half inch gap between the Unit I and Unit 2 buildings.
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During the night of December 3 and the morning of December 4,1986, the licensee. I took immediate corrective action to determine the extent of, and to limit the spread of, contamination. A series of water samples were collected to determine which outfall areas of the swamp contained contaminated water. Additionally, a ,
series of dikes (dams) were constructed to. minimize the spread of the contami- '
nation. The licensee promptly performed radiation surveys of. the affected buildings and the outfall area of the swamp to determine the extent of the release. The licensee estimated that the maximum amount of radioactive material that could have been released was approximately 0.4' curies.
During the afternoon of December 4,1986, direct radiation readings along the outfall which led to the swamp ranged from 1-2 mrem /hr. Direct radiation read-ings along the shoreline of the swamp ranged from 150-250 uR/hr (normal readings l were typically 8-10 uR/hr). Water sample results at the location identified as a potential pathway to the river indicated no detectable activity. The water ] j from behind the dikes and at the entrance to the swamp pool was pumped back to I tanker trucks, filtered through demineralizers, and then discharged through the plant radwaste system with all normal discharge precautions in place.
Fuel pool to transfer canal gates were installed on December a,1986, and by j December 5,1986, the air supplies were configured such that the transfer c6nal
}
gate seals were fed from both units such that the inner seals were supplied from one unit and the outer seals were supplied from the other unit, thus redundancy )
1 was achieved. Also, by December 6,1986, the transfer canal seals were pres-surized from each unit such that a single loss of air supply would not cause the seals to deflate.
In addition to the cleanup efforts required (for the inplant areas and the swamp area located near the cooling towers on the east side of the plant) and establishment of sampling stations at various locations, the licensee planned an augmented environmental sampling program to detect any significant future.
migration of the radioactive material in excess of permissib'e levels.
The NRC AIT concluded that the organization, staffing, controls and coordination of the recovery efforts were rapid and effective. The licensee's Technical ;
Support Center functioned well as the recovery center. The staffing of the 1 recovery center was adequate, with all needed disciplines represented. The personnel at the spill area in the swamp were well managed and effective in l j
.containing the extent of the contamination within the site boundaries and in minimizing the area of the swamp which was contaminated.
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The AIT' inspection was conducted from December 4 through December 7,1986. The 'j inspection findings are contained in NRC Inspection Report Nos. 50-321/86-41 i and 50-366/86-41. The report was forwarded formally to the license on -
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,o Jamuary 8,1987 (Ref. C-2). Apparent violations were identified during the inspection involving design and procedures. The violations are being evaluated and will b'e forwarded by separate NRC correspondence.
On February 24, 1987, the NRC issued Inspection.and Enforcement Notice No. 87-13 (Ref. C-3) . The Notice described the Hatch event and alerted addressees to the pctential for high radiation fields following loss of water from the fuel pool.
As described in the Notice, analysis by the licensee after the event has shown that if water had been completely lost from the fuel transfer canal, high ra- ,
diation fields would be high enough such that remedial measures may have been i difficult to take. Some irradiated control rod blades, stored on short hanger ,
rods clipped over the side of the spent fuel pool, would become completely I uncovered, resulting in general area radiation levels of about 100 R/hr at the edge of the spent fuel pool and about 1 R/hr six feet from the pool edge. [To avoid this potential hazard, the licensee is shipping the control rod blades off the site.] However, no fuel damage of the stored spent fuel components in the pool would be expected because even if the water level dropped to the bot-torn of the fuel transfer canal, about two feet of water would remain over the top of the spent fuel.
The Notice also mentioned that concern has been raised about the design of the i i leak detection system for the seals. The NRC is currently evaluating the design '
to determine whether it is adequate.
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REFERENCES.
FOR APPENDICES B - 1. U.S. Nuclear Regulatory Comission, Inspection and Enforcement Information Notice No. 86-53, " Improper Installation of Heat Shrinkable Tubing,"
June 26, 1986.*
B-2 .U.S. Nuclear Regulatory Comission, Inspection and Enforcement Information Notice No.86-104, " Unqualified Butt Splice Connectors Identified in Qualified Penetrations," December 16, 1986.*
B-3 U.S. Nuclear Regulatory Comission, " Programmatic Environmental Impact Statement Related to Decontamination and Disposal of Radioactive Wastes Resulting'from March 28 1979 Accident Three Mile-Island Nuclear Station,. 1 Unit 2," NUREG-0683 Docket No. 50-320,' Draft Supplement No. . 2- (dealing : i with disposal of accident-generated water) issued for public coment during December 1986..
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B-4 Letter from James M. Taylor, Director, NRC Office of Inspection and' Enforcement, to J. G. Randolph, . President, Kerr-McGee Center, Sequoyah Fuels Corporation, forwarding a Notice of Violation and Proposed. Imposition-of Civil Penalties, Docket No. 40-08027, October 14, 1986.* '
B-5 Letter from James M. Taylor, Director, NRC Office of Inspection and Enforcement, to J. G. 'Randolph, President : Kerr-McGee Center Sequoyah Fuels Corporation, forwarding an Order Imposing Civil Monetary Penalties, Docket No. 40-08027, Februa ry . 5, 1987.*
B U.S. Nuclear Regulatory Comission, " Release of UF, from a Ruptured Model 48Y. Cylinder 'at Sequoyah Fuels Corporation Facility: ' Lessons :- Learned Report," USNRC Report NUREG-1198, published June.1986.**
B-7 Letter from Jack A. Hind, Director, Division'of Radiation Safety and Safe-guards, NRC Region III, to J. C. Bishop, Plant Marager, Metropolis Works, Allied Chemical Company, forwarding. (1) a Notice <>f Violation,-and (2) Inspection Report No. 86-06, Docket No. 40-3392, November 18, 1986.*
I B-8 Letter from James G. Keppler, Regional ' Administrator, NRC Region 'III, to L. L. Taunton, Vice President, Operations, Engineered Materials Sector, Allied-Signal Corporation, forwarding a Notice of Violation and Proposed Imposition of Civil Penalties, Docket No. 40-3392.-June 27, 1986.*
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- Available in NRC Public Document Room,1717 H Street, NW, Washington, DC 20555, for inspection and copying '(for a fee).
. Single copies of NRC draft reports are available free, to the extent of .
supply, upon written request to the Division of Technical Information and Document Control, U.S. Nuclear Regulatory Comission, Washington, DC 20555.
- Available in NRC Public Document Room,1717 H Street, NW, Washington, DC-20555, for inspection. Available for. purchase from.the GPO Sales Program, '
Superintendent of Documents, U.S. Government Printing' Office, Post Office
. Box' 37082, Washington, DC 20013-7982.
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..- 4 B Letter from James M. Taylor, Director, NRC Office of Inspection and En-
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.forcement, to L. R. Taunton. Vice President, Operations Engineered Mate-
' rials Sector, Allied-Signal Corporation, Docket No. 40-3392, December-19, 1986.*-
B-10 Letter from James M. Taylor, Director, NRC Office of Inspection and En-forcement, to W. David Keating, Vice President, Ancillary Services, Mercy Hospital, forwarding (1) an Order to Show Cause Why the License Should Not Be Modified and (2) a Notice of Violation ~ and Proposed Imposition of Civil Penalty, Docket No. 30-02971, June 17. 1986.*
B-11 Letter from James M. Taylor Director, NRC Office of Inspection and En-forcement, to Salvatore M.' Imperiale, M.D., Director of Nuclear. Medicine -
Valley Radiology Associates, Inc., forwarding an Order to Show Cause Why the License Should Not Be Modified, Docket'No. 30-15110, June 17, 1986.* {
'B-12 Letter from James M. Taylor, Director, NRC Office of Inspection and 1 Enforcement, to Robert J. Moylan, Executive Vice President, Mercy Hospital, j Docket No. 30-02971, December 24, 1986.*- '
B-13 Letter from James M. Taylor, Director, NRC Office of Inspection and En-forcement, to Salvatore M. Imperiale, M.D., Director of Nuclear' Medicine, Valley Radiology Associates, Inc., Docket No.~ 30-15110, December 24, 1986.*
C-1 ~)
Letter from William F. Kane, Director, Division of Reactor Projects, NRC-1 Region I, to C. A. McNeill,'Jr., Vice President-Nuclear Public. Service j Electric and Gas Company, forwarding (1) potential enforcement issues. l and (2) Region I Augm(nted Inspection Team Report No. 50-354/86-50, d Docket No. 50-354, October 31, 1986.*
I C-2 Letter from J. Nelson Grace, Regional Administration NRC Region II, to J. H. Miller, Jr., President, Georgia Power Company, forwarding Inspection.
Report Nos. 50-321/86-41 and 50-366/86-41, Docket Nos. '50-321 and 50-366 January 8,1987.*
C-3 U.S. Nuclear Regulatory Connission, Inspection and Enforcement Notice j No. 87-13, " Potential for High Radiation Fields Following Loss of Water from Fuel Pool," February 24, 1987.* )
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- Available in NRC Public Document Room,1717 H Street, NW, Washington, DC 20555,-for inspection and copying (for a fee).
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(
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l . Enclosure 4 (Enclosure 3 to the Commission Paper)
OTHER EVENTS. CONSIDERED FOR ABNORMAL OCCURRENCE REPORTING-The following incidents are samples of the incidents seriously' considered for abnormal occurrence-(AO) reporting. The incidents are briefly discussed and the reasons why they are not being reported are stated. The. incidents were judged not to have involved any major reduction'in the level.of protection provided for public health or safety.
This enclosure is provided to the Commission per Commission ccaments on SECY-76-471, dated December 2,1976; the enclosure is not intended to be'a part of the published report.
- 1. Incineration of Molybdenum-99/ Technetium-99m Generator at a Hospital On October 21, 1986, Henry Heywood Memorial Hospital of Gardner, Massachusetts, I reported that a. molybdenum-99/ technetium-99m generator containing 880 millicuries .
of molybdenum-99 (as of noon October 19,-1986) had bean inadvertent 1yLincinerated in the hospital's incinerator on the ' evening of October 19, 1986. Initial. l) surveys of the incinerator performed by the. licensee revealed only backg,round.
radiation levels. Therefore, the licensee-assumed that the molybdenum had vaporized and was released through the stack. The licensee's~ surveys of the-grounds surrounding the hospital:likewise revealed only background radiation -
levels. The incinerator was cleaned out and the debris held just in case'it - a was contaminated. The licensee subsequently used the incinerator two more times.
Radiation surveys conducted by NRC inspectors;upon arrival at the site revealed' radiation levels exceeding 200 mR/hr in the~ incinerator and the container hold <
ing the debris from the incinerator. Immediate' corrective action taken by the licensee included roping off the area surrounding the incinerator, removing the l container containing the debris to a restricted access area, and shielding it '
with lead sheets. Use of the incinerator was suspended until radiation levels decreased to background. About 1492 mil 11 curies of technetium-99m were esti-anated to have been released. The resulting concentration of technetium-99m released, 0.22 E-7 uC1/m1, when averaged over a one year period was less than the limit specified in 10 CFR Part 20, Appendix 8, Table II, Column I for "
technetium-99m (5 E-7 uCi/al). When averaged over a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period, the result-ing concentration was about 16 times the limit specified in.10 CFR Part 20, Appendix B, Table II.
A number of factors contributed to the-incident, including inadequate training
~
of the nuclear medicine technician who performed the initial surveys and of the personnel who were expected to handle and control radioactive materials. Also, inadequate management involvement in the program contributed to the licensee's ineffectiveness in correctly evaluathg the effects' of the event.
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Oh' December 11, 1986, the NRC forwarded to the ifcensee a Notice of Violation and Proposed Imposition of Civil- Penalty in the amount of $2,500. On January 2, ll 1987, the licensee responded by enclosing a' check for $2,500, taking exception I
to some of the findings, and describing corrective actions taken. By letter dated March 19, 1987, the NRC stated that the licensee's response was carefully l considered, but it was concluded that the violations did occur as stated in the '
December 11, 1986 NRC letter. The licensee's implementation of corrective I actions would be examined during 'a subsequent NRC inspection. i No personnel exposures were attributable to this event. The potential hazard to personnel in unrestricted areas was relatively low due to technetium-99m's shcrt half life (about 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />) and the low energy of its radiation. There-fore, the event is considered below the threshold for A0 reporting.
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- 3. Inhalation of Iodine-125 from Leaking Sources On November 10, 1986, the NRC issued Inspection and Enforcement Information ;
Notice No. 86-95 to all NRC licensees authorized to use Lixi, Inc. imaging I devices or any other bone mineral analyzer with a sealed source containing l iodine-125. The Notice alerted licensees that a recent incident indicated that the normal means of testing the sealed source in such devices for leakage was ineffective and resulted in several people inhaling small amounts of the ;
radioactive material. '
As described in the Notice, Lixi, Inc. imaging devices include a sealed source containing 220 to 450 mil 11 curies of iodine-125 and function much like an x-ray i fluoroscope. The source has to be tested every 6 months to determine if there is any leakage of the contained iodine-125. Licensees who keep sources longer than 6 months are required by license to perform this test and send the test samples out for analysis. This test is now performed by using an alcohol-unoistened Q-tip and a dry Q-tip to wipe certain portions of the source holder as specified in the instructions. The purpose of the test is to determine if any particulate iodine-125 is on the sample Q-tips, which would indicate the source was breached and would have to be replaced to avoid a contamination probles. However, the recent incident at the Lixi, Inc. facility showed that the alcohol and dry wipes were not an adequate means of detecting a leaking source.
3-3 l ,
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. during the investigation of the incident, it was. found that 'two sources that jE had been returned for exchange were leaking.' Alcohol and dry wipes did not.
reveal any removable iodiae-125 on the various surfaces because the escaping iodine-1(5 came out in gaseous form rather than as particulate matter.. As a result, about 15 people inhaled small' amounts of iodine-125 which deposited in ,
1 their thyroids. The iodine-125 was apparently adsorbed on carbon containing material such as cardboard, rubber bands, styrofoam, and charcoal that was' near the escaping gas. A survey of- these materials revealed elevated radiation levels. Through these direct surveys and a series of air samples using filter
~
media containing charcoal, the ruptured sources were located and the airborne contamination problem was resolved.
The distribution of the sources was' temporarily stopped and then resumed when Lixi, Inc. . decided to put charcoal or carbon in the delivery packages. The charcoal would show whether the sources were leaking. A review of the design and reevaluation is being done. by Atomic Energy of Canada, Ltd. , (AECL) to ;
redesign the source to preclude leakage of iodine-125. At this writing, no !
new design has been developed, although AECL said they would resubmit a design sometime in early 1987.
In addition, recipients of the iodine-125 sources have been advised regarding the method for leak testing, e.g. , placing charcoal or carbon near the sources to detect leakage. Also, companies that perform a leak-testing service were informed about using carbon to check the sources.
This event was considered for conormal occurrence reporting, but was rejected because the amounts of radioactive material inhaled were below the NRC limits.
- a ,
- 4. Order to Show Cause Why License Should Not Be Revoked On December 23, 1986, the Director of the NRC Office of Inspection and Enforcement ordered Dr. Kedarnath B. Joshi of Livonia, Michigan, to show-cause why his NRC license should not be revoked. The Highland Waterford Medical i Services (HWMS) in Pontiac, Michigan also was issued an order to show-cause why its NRC license should not be modified to prohibit Dr. Joshi from engaging in any NRC licensed activity at its facility. Dr. Joshi is licensed by the ;
NRC to use radiopharmaceuticals for diagnostic studies at his private :
clinics. HWMS was not one of his clinics.
An NRC inspection between June 12 and July 1,1986, detemined that Dr. Joshi administered radiopharmaceuticals to patients at the HWMS facility from May 21 through June 11, 1986, although his license did not authorize use at that location.
In addition, even though a technologist and two consultants. informed Dr. Joshi he would need an amendment to his license to use radiopharmaceuticals' at HWMS, and even though the radiopharmaceutical supplier refused to deliver material to HWMS for Dr. Joshi, the NRC staff determined he. " willfully and egregiously circumvented the conditions" of his license by requesting the supplier to deliver material to a vacant building in Pontiac, Michigan, where he formerly conducted radioisotope procedures. (The supplier did not know the building 3-4 i
, . - e j
\
n
.wa!, ' vacant.) Dr. Joshi then transported.the radiopharmaceuticals to HWMS and i administered them to patients on June 9,10, and 11,1986.
On January 14, 1987, Dr. Joshi responded in writing to the NRC show-cause Order and requested a hearing. The Highland Waterford Medical Services did not respond to the show-cause Order, and has terminated Dr. Joshi's services.
The enforcement action resulting from Dr. 'Joshi's activities did not have any direct safety significance. He was authorized to perform the medical proce-dures using licensed materials, but the authorization was restricted to his 9 private practice clinics. The Show Cause Order was issued because he willfully 1 violated his license by performing diagnostic procedures at Highland Waterford l
-Medical Services. l
)
This matter was considered as a possible abnormal occurrence in regard to deficiency in management or procedural controls; however, it was judged not to ;
seet the criteria because of the small size of the medical program and the 1 absence of any direct safety significance. !
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,, Enclosure 5 l
A0 Coordinator Information 1 Office / Division Present A0 Coordinator (s)
Office of the General Counsel / T. Dorian [ ]
Office of Governmental and Public Affairs l Public Affairs F. Ingram 3 State, local, and Indian Tribes Programs. J. Lubenau (Agreement '
States)
Deputy Executive Director for Regional .None, at present-l Operations l Office of Enforcement None,.at present Office of Special Projects None, at present i >
! Office of Administration and Resources Management Division of Security 4. Filger l
Office of Nuclear Material Safety and Safeguards R. Gramann R. O'Connell Division of Fuel Cycle, Medical, None, at present Academic and Commercial Use Safety Office of Nuclear Reactor Regulation M. Caruso and G. Holahan
( Associate Director for Projects . None, at present >
l Associate Director for Inspection and None, at present
! Technical Assessment Division of Operational Events Assessment None, at present Division of Reactor Inspection and Safeguards None, at present !
Office of Nuclear Regulatory Research J. Glynn Region I K. Murphy Region II K. Landis Region III J. Strasma l
Region IV D. Powers l Region V J. Crews c _ - - _ __. _ - - - _ _ - - - _ _