ML20237G332
| ML20237G332 | |
| Person / Time | |
|---|---|
| Site: | 05000000, Palisades |
| Issue date: | 09/30/1986 |
| From: | James Keppler NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | Heltemes C NRC OFFICE FOR ANALYSIS & EVALUATION OF OPERATIONAL DATA (AEOD) |
| Shared Package | |
| ML20237F649 | List: |
| References | |
| FOIA-87-377, RTR-NUREG-0090, RTR-NUREG-90 NUDOCS 8709020206 | |
| Download: ML20237G332 (1) | |
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September 30, 1986 MEMORANDUM FOR:
C. J. Heltemes, Jr., Director, Office of Analysis i
and Evaluation of Operational Data FROM:
James G. Keppler, Regional Administrator
SUBJECT:
ABNORMAL OCCURRENCE REPORT TO CONGRESS FOR SECOND QUARTER CY 1986 We have reviewed the subject report for the items involving Region III I
facilities. Several minor additions and editorial changes were made on Pages 12, 13, 33, 39, and 48. These pages are attached with pen and ink
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changes reflecting our coments.
J In additier, the Office of Nuclear Reactor Regulation has indicated to us that it believes that the Palisades e designation as a Abnorrnal Occurrence.quipment failures issue does not warrant Upon review, we support this position.
The equipment failures primarily involved non-safety-related systems, and there were no violations (and consequently no enforcement action) developed in the i
l Augmented Investigation Team (AIT) report. A more appropriate designation for the issue would be as an Appendix C ltem, since an AIT was involved. This would be consistent with the position stated in your July 1,1986, memorandum.
If you have any questions, please contact Jan Strasma of my staff (FTS 388-5674).
I trtginal itsnod by A, Scrt Davis y James G. Keppler Regional Administrator
Attachment:
As Stated i
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l MEMORANDUM FOR:
H. R. Denton, Director, NRR J. G. Davis, Director, NMSS D. F. Ross, Acting Director, RES J. M. Taylor, Director, IE G. H. Cunningham, Deputy General Counsel J. J. Fouchard, Director, PA G. W. Kerr, Director, SP Regional Administrators i
FROM:
C. J. Heltemes, Jr., Director Office for Analysis and Evaluation of Operational Data j
SUBJECT:
ABNORMAL OCCURRENCE REPORT TO CONGRESS FOR l
SECOND QUARTER CY 1986 I
Based on staff response to the AEOD July 1, 1986 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), and the Fourth Quarter CY 1985 Abnormal Occurrence (AO)
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Report to Congress (Enclosure 3).
l shows that there are four proposed A0s at the nuclear power plants licensed to operate, five at the other NRC licensees, and two for the Agreement States. There are ten items for Appendix 8 (" Update of Previously Reported Abnormal Occurrences") and six items for Appendix C ("Other Events of Interest").
The differences between the items shown in Enclosure'l and the items suggested i
in our July 1, 1986 memorandum are described below.
l l
Abnormal Occurrences l
A0 86-10 (" Boiling Water Reactor Emergency Core Cooling System Design Deficiency")
is a new A0 proposed by NRR. The significant design deficiency involves Pilgrim, Quad Cities Units 1 and 2, and Dresden Units 2 and 3.
1 A0 86-11 (" Continuing Multiple Equipment Failures at Palisades") is a new A0 proposed by Region III based on significant deficiency in management and pro-cedural controls, particularly in the area of maintenance.
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A0 86-13 (" Therapeutic Medical Misadministration") involving Maryview Hospital of Portsmouth, Virginia, is a new A0 proposed by AE0D based on an enforcement l
action dated August 7, 1986. The described event clearly meets the guidelines for reporting medical misadministration as A0s.
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A0 86-15 (" Diagnostic Medical Misadministration") involving Robert Packer
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Hospital and Guthrie Clinic of Sayre, Pennsylvania is a new A0 proposed by
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A Multiple Addressees 2
i Region I.
A0 86-16 (" Diagnostic Medical Misadministration") involving Tripler Army Medical Center, Tripler AMC, Hawaii, is a new A0 proposed by Region V.
Both of these events involved doses exceeding five times the intended doses.
A586-5 (" Uncontrolled Release of Krypton-85 to an Unrestricted Area") is a new A0 submitted by the State of Arizona.
In our July 1, 1986 memorandum, we suggested that management deficiencies at Byron Unit 1 be considered as an AD.
Both IE and Region III feel that the deficiencies were not sufficient to meet the threshold for A0 reporting. We agree.
Our memorandum also reminded Region IV that they were still considering alleged violations at Pathfinder Mines Corporation as a possible AO.
In their memoran-dum to us dated July 30, 1986, they state that based on further discussions with the licensee, they are recommending to Headquarters that the severity level be reduced and no civil penalty be imposed. Therefore, they believe the item should be dropped. We have no objections; however, if the situation changes, Region IV should reconsider whether the item should be reported.
Appendix B Items A0 85-1 (" Premature Criticality During Startup") has been modified to not only include the April 13, 1986 similar event at San Onofre Unit 2 (as suggested in
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our July 1,1986 memorandum), but also a similar event at Fermi Unit 2 on July 1, 1985. AE00 prepared the Fermi Unit 2 writeup based on the information contained, l
l or referenced, in the significant enforcement action taken on July 3,1986, and l
in Region III's response to our July 1, 1986 memorandum.
A0 86-6 (" Breakdown of Management Controls at an Irradiator Facility") involving Radiation Technology, Inc., Rockaway, New Jersey, was added by AE00 to describe Information Notice No. 86-54, issued June 27, 1986.
Appendix C Items Two new items, " Fire in Charcoal Filter Tanks at Perry Unit 1" and " Water Level Instrumentation Problems at Lacrosse," were provided by Region III based on suggestions by P. Bobe of AE00 to J. Strasma of Region III.
In our July 1, 1986 memorandum, we stated that after an Augmented Inspection Team (AIT) issues their report, the report (and the associated problems which initiated the creation of the AIT) should generally be discussed in the quarterly reports as Appendix C items, unless the events, the Team's findings, and actions taken (NRC and/or licensee) warrant upgrading to A0 status.
In Region I's July 25, 1986 reply to our July 1, 1986 memorandum, they state that "---we suggest that AEOD consider each AIT on its own merits and to make a decision as to whether it is of interest to Congress.---Region I bclieves that additional criteria should be developed for deciding whether AIT's are of sufficient interest to be reported as Appendix C items."
i Mul'tiple Addressees 3
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While we agree that each AIT should be evaluated on its own merits, we do not believe that additional criteria are needed.
Part III of NRC Appendix 0212 presents the guidelines for selecting Appendix C items. Congressional interest is only one of the many guidelines.
It is noted that the Pilgrim problems i
required considerable corrective actions (both by the licensee and by the NRC in its inspections and followup reviews). How well the licensee performs its review and corrective actions are not germane, except if they are particularly 1
faulty it may be an added basis for reporting as an AO.
Also, the problems j
have resulted in a considerable amount of downtime.
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In addition, leakage past valves forming the isolation barrier between the high pressure reactor coolsnt system and the low pressure piping of. the RHR system has been a recurrent problem at BWRs. A possible interface LOCA could bypass containment. This recurring problem was the subject of an AE00 Case Study (AE00/C502) issued in September 1985. The problem was also reported as A0 84-8 in NUREG-0090, Vol. 7, No. 3.
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Finally, it would not be unreasonable to assume that the mere activation of an AIT could well be perceived by the public (or Congress) to be of public health or safety concern (regardless of whether or not the NRC had a real concern regarding safety when the AIT was activated).
We request your review, comments and concurrence on this memorandum and its i
enclosures no later than September 30, 1986. Please provide updating informa-(
tion, if necessary, for all of the items in Enclosures 1 through 3 to reflect l
any changes in status. We plan to submit the report to the Commission by mid October 1986.
If you have any questions, please contact Paul Bobe at 492-4494.
$w D.. Auve$. I&
J. Heltemes, Jr., Director Office for Analysis and Evaluation of Operational Data
Enclosures:
As Stated cc w/ enclosures:
J. Sniezek, DEDROGR J. Glynn, RES K. Murphy, Region I R. Whipp, SEC/ADM L. Cobb, IE L. Bettenhausen, Region I T. Oorian, OGC E. Weiss, IE K. Landis, Region II G. Holahan, NRR J. Callan, IE R. Aiello, Region II M. Caruso, NRR (3)
R. O'Connell, NMSS J. Strasma, Region III J. Lubertau, SP R. Gramman, NMSS
- 0. Powers, Region IV
1-Multiple Addressees 4
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1 For:
The Commissioners from:
Victor Stello, Jr.
Executive Director for Operations l
Subject:
SECTION 208 REPORT TO THE CONGRESS ON ABNORMAL OCCURRENCES FOR APRIL - JUNE 1986 1
Purpose:
Approval of Final Draft l
l Discussion: 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 second quarter of CY 1986. is a final draft of the quarterly report
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to Congress on abnormal occurrences (A0s). The report covers the period from April 1 to June 30, 1986.
This draft incorporates the major comments obtained from staff review of a previous draft. The draft report is similar in format to the first quarter CY 1986 report (published as NUREG-0090, Vol. 9, No. 1).
The draft report contains nine proposed A0s for NRC licensees. The items are:
86-8 Decay Heat Removal Problems at U.S. Pressurized Water Reactors (many plants over the years).
86-9 Out-of-Sequence Control Rod Withdrawal at Peach Bottom Unit 3.
86-10 Boiling Water Reactor Emergency Core Cooling System Design Deficiency (Pilgrim, Quad Cities Units 1 and 2, Dresden Units 2 and 3).
86-11 Continuing Multiple Equipment Failures at Palisades.
86-12 Willful Failure to Report a Diagnostic Medical 1
Misadministration (Mercy Hospital; Wilkes-Barre, Pennsylvania).
Contact:
Paul Bobe, AEOD X24494 1
l 3
The Commissioners 2
86-13 Therapeutic Medical Misadministration (Maryview Hospital; Portsmouth, Virginia).
86-14 Willful Failure to Report Diagnostic Medical Misadministration (Bloomington Hospital; Bloomington, Indiana).
86-15 Diagnostic Medical Misadministration (Robert Packer Hospital and Guthrie Clinic; Sayre, Pennsylvania).
86-16 Diagnostic Medical Misadministration (Tripler Army Medical Center; Tripler AMC, Hawaii).
I In regard to the five medical A0s, A0 86-12 and 19 86-14 rep-resent serious deficiencies in management and procedural controls.
The other three meet the Commission approved staff guidelines for selection of medical misadministration for A0s (i.e., Part II of NRC Appendix 0212). A0 86-13 in-volved a patient receiving a therapy dose in a chemical
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form other than that intended, resulting in an unintended dose of several hundred rads to the patient's bone marrow.
I A0 86-15 and A0 86-16 involved diagnostic doses greater than five times the intended doses.
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The draf t report contain's two proposed A0s submitted by the Agreement States. The items are:
A585-5 Uncontrolled Release of Krypton-85 to an Unre-stricted Area (Micro-Rel Division, Medtronic, Incorporated; Tempe, Arizona).
A586-6 Contaminated Radiopharmaceutical Used in Diagnostic Administrations (Scripps Memorial Hospital; Encinitas, California).
Appendix B of the draf t report contains updating informa-tion for the following previously reported A0s:
Nuclear Power Plants 79-3 Nuclear Accident at Three Mile Island - Further information is provided and the item remains open.
85-1 Premature Criticality During Startup - This pre-viously closed A0, involving Summer Unit 1, is i
reopened to describe two somewhat similar events at Fermi Unit 2 and San Onofre Unit 3.
Th'e item is then reclosed.
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j The Commissioners 3
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85-7 Loss of Main and Auxiliary Feedwater Systems (Davis-Besse) - Further information is provided and the item remains open.
85-14 Hanagement Deficiencies at Tennessee Valley Au -
thority - Further information is provided and the
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item remains open.
J 85-20 Management Deficiencies at Fermi Nuclear Power.
Station - Further information is provided and the item remains open.
86-1 Loss of. Power and Water Hammer Event (San Onofre i
' Unit 1) - Further. information is provided and l
the item is closed-out.
86-2 Loss of Integrated Control System Power and Over-cooling Transient (Rancho Seco) - Further informa-tion is provided and-the item remains open.
Fuel Cycle Facilities
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86-3 Rupture of Uranium Hexafluoride Cylinder and Release of Gases (Sequoyah Fuels Corporation;
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Gore, Oklahoma) - Further information is I
I provided and the item remains open.
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Other NRC Licensees i
84-13 Contaminated Radiopharmaceuticals Used in Diagnos-l tic Administrations (Nuclear Pharmacy, Inc., of l
Chicago, Illinois; Syncor International, Inc. of Blue Ash, Ohio) - This previously closed A0 is, reopened to provide information on significant enforcement action being taken. The item is then reclosed.
86-6 Breakdown of Management Controls at an Irradiator Facility (Radiation Technology',-Incorporated; Rockaway, New Jersey) - Further information is provided and.the item remains open.
The draf t report contains six items for Appendix C ("Other Events of Interest").
The items are:
1.
Reactor Vessel Indications at Oconee Unit 1.
2.
NRC Augmented Inspection Team Sent to Pilgrim.
3.
Construction Problems at Comanche Peak.
n The Commissioners 4
i 4.
Sabotage of Offsite Power Lines to Palo Verde.
S.
Fire in Charcoal Filter Tanks at Perry Unit 1.
6.
Water Level Instrumentation Problem at Lacrosse.
When Commission approval is received, the report will be up-dated, if necessary, before it is published. The report will be designated as NUREG-0090, Vol. 9, No. 2.
l There are no items for Enclosure 3 (i.e., items which were candidates for inclusion as A0s, but which in the staff's judgement did not meet the' criteria for A0 reporting after further study) to this Commission Paper..Any such items were either upgraded to proposed A0s or Appendix C items.
Recommendations:
That the Commission:
1.
Approve the contents of the proposed Second Quarter CY 1986 Abnormal Occurrence Report to Congress, and l
2.
Note that upon approval and publication, forwarding letters to the Speaker of the House and the President of the Senate will be provided to the Chairman for signature. Congressional Affairs will then arrange for appropriate distribution to Congress. A Federal i
l Register notice will be issued to announce the avail-i ability of the quarterly report.
In addition, a sepa-rate Federal Register notice (describing details of the l
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-terly reports as soon as reasonably possible.
It is ex-pected that Commission action within two weeks of receipt of the draft would permit publication and dissemination 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 Second Quarter CY 1986,,
Abnormal Occurrence Report to Congress l
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Page 1 of 2 I
DRAFT (Enclosure 1 of Commission Paper)
The Honorable Thomas P. O'Neill, Jr.
Speaker of the United States-
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House of Representatives Washington, DC 20515
Dear Mr. Speaker:
Enclosed is the NRC report on abnormal occurrences at licensed nuclear facilities, as required by Section 208 of the Energy Reorganization Act of 1974 (PL 93-438), for the second 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 public health or safety. The report states that for this report period, there l
were four abnormal occurrences at the nuclear power plants licensed to operate.
The events were (1) decay heat removal problems at U.S. pressurized water re-actors, (2) out-of-sequence control rod withdrawal at Peach Bottom Unit 3, (3) boiling water reactor emergency core cooling system design deficiency, and (4) continuing multiple equipment failures at Palisades. There were five ab-normal occurrences at the other NRC licensees. Two involved willful failure to report diagnostic medical misadministration to the NRC; one involved a thera-peutic medical misadministration; and two involved diagnostic medical misadministration. There were two abnormal occurrences reported by the
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Agreement States. One involved an uncontrolled release of krypton-85 to an l
unrestricted area; the other involved a contaminated radiopharmaceutical used in diagnostic administrations.
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The report also contains information updating some previously reported abnormal
.i occurrences.
1 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.
l Sincerely, Lando W. Zech, Jr.
Chairman l
Enclosure:
Report to Congress on Abnormal Occurrences (NUREG-0090, Vol. 9 No. 2)
s Page 2 of 2 i
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DRAFT l
(Enclosure 1 of Commission Paper)
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The Honorable George H.
W. Bush j
President of the Senate i
Washington, DC 20510
Dear Mr. President:
Enclosed is the NRC report on abnormal occurrences at licensed nuclear facilities, as required by Section 208 of the Energy Reorganization Act of 1974 (PL 93-438), for the second calendar quarter of 1986.
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In the context of the Act, an abnormal occurrence is an unscheduled incident or I
event which the Commission determines is significant from the standpoint of
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public health or safety. The report states that for this report period, there were four abnormal occurrences at the nuclear power. plants licensed to operate.
The events were (1) decay heat removal problems at U.S. pressurized water re-j actors, (2) out-of-sequence control rod withdrawal at Peach Bottom Unit 3, (3) boiling water reactor emergency core cooling system design deficiency, and j
(4) continuing multiple equipment failures at Palisades. There were five ab-normal occurrences at the other NRC licensees. Two involved willful failure to report diagnostic medical misadministration to the NRC; one involved a thera-l
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peutic medical misadministration; and two involved diagnostic medical I
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misadministration. There were two abnormal occurrences reported by the Agreement States. One involved an uncontrolled release of krypton-85 to an unrestricted area; the other involved a contaminated radiopharmaceutical used in diagnostic administrations.
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. 2)
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NUREG-0090 Vol. 9 No. 2 DRAFT (Enclosure 2 of Commission Paper) j 1
REPORT TO CONGRESS ON l
l ABNORMAL OCCURRENCES I
APRIL - JUNE 1986 Status as of July 31, 1986 Date Published:
November 1986
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Office for Analysis and Evaluation of Operational Data j
United States Nuclear Regulatory Commission j
Washington, D.C.
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k ABSTRACT Section 208 of the Energy Reorganization Act of 1974 identifies an abnormal l
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.
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This report covers the period f rom April 1 to June 30, 1986.
The report states that for this reporting period, there were four abnormal oc-currences at the nuclear power plants licensed to operate. The events were 1
(1) decay heat removal problems at U.S. pressurized water reactors, (2) out-of-sequence control rod withdrswal at Peach Bottom Unit 3, (3) boiling water reactoa emergency core cooling system design deficiency, and (4) continuing multiple equipment failures at Palisades.
There were five abnormal occurrences at the other NRC licensees. Two involved willful failure to report diagnostic i
medical misadministration to the NRC; one involved a therapeutic medical I
misadministration; and two involved diagnostic medical misadministration.
There were two abnormal occurrences reported by the Agreement States. One j
involved an uncontrolled release of krypton-85 to an unrestricted area; the I
other involved a contaminated radiopharmaceutical used in diagnostic administrations.
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The report also contains information updating some previously reported abnormal occurrences.
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i CONTENTS i
Page ABSTRACT iii PREFACE vii
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INTRODUCTION.
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THE REGULATORY SYSTEM................
vii REPORTABLE OCCURRENCES.........
viii AGREEMENT STATES.
ix FOREIGN INFORMATION x
REPORT TO CONGRESS ON ABNORMAL OCCURRENCES, JANUARY-NARCH 1986..
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NUCLEAR POWER PLANTS........
1 86-8 Decay Heat Removal Problems at U.S. Pressurized Water Reactors...........................
1 86-9 Out-of-Sequence Control Rod Withdrawal at Peach Bottom Unit 3........
7 86-10 Boiling Water Reactor Emergency Core Cooling System Design Deficiency..
9 86-11 Continuing Multiple Equipment Failures at
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Palisades............................
11 FUEL CYCLE FACILITIES (Other than Nuclear Power Plants)...
14 OTHER NRC LICENSEES (Industrial Radiographer, Medical Institutions, Industrial Users, etc.).........
14 86-12 Willful Failure to Report a Diagnostic Medical Misadministration.
14 86-13 Therapeutic Medical Misadministration.
15 86-14 Willful Failure to Report Diagnostic j
Medical Misadministration....
17 86-15 Diagnostic Medical Misadministration..
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86-16 Diagnostic Medical Misadministration..
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AGREEMENT STATE LICENSEES................................
20 ASS 6-5 Uncontrolled Release of Krypton-85 to an Un re s t ri c ted Are a.................................... 20 AS86-6 Contaminated Radiopharmaceutical Used in i
Diagnostic Administrations.........
.. 21 REFERENCES 23 APPENDIX A - ABNORMAL OCCURRENCE CRITERIA.......................
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CONTENTS (continued)
Page APPENDIX B - UPDATE OF PREVIOUSLY REPORTED ABNORMAL OCCURRENCES.
27 NUCLEAR POWER PLANTS 27 79-3 Nuclear Accident at Three Mile Island..
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85-1 Premature Criticality During Startup 29 l
85-7 Loss of Main and Auxiliary Feedwater Systems 32 85-14 Management Deficiencies at Tennessee Valley Authority........
33 85-20 Management Deficiencies at Fermi Nuclear Power Station......
36 86-1 Loss of Power and Water Hammer Event......
37 86-2 Loss of Integrated Control System Power and 37 Overcooling Transient i
FUEL CYCLE FACILITIES.
38 86-3 Rupture of Uranium Hexafluoride Cylinder and Release 38 of Gases OTHER NRC LICENSEES 39 84-13 Contaminated Radiopharmaceuticals Used in Diagnostic
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Administrations.
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86-6 Breakdown of Management Controls at an Irradiator Facility 40 APPENDIX C - OTHER EVENTS OF INTEREST 41 l
REFERENCES (FOR APPENDICES) 50 l
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PREFACE INTRODUCTION The Nuclear Regulatory Commission reports to the Congress each quarter under provisions of Section 203 of the Energy Reorganization Act of 1974 on any abnormal occurrences involving facilities and activities regulated by the NRC.
An abnormal occurrence is defined in Section 208 as an unscheduled incident or event which the Commission determines is significant from the standpoint of public health or safety.
Events are currently identified as abnormal occurrences for this report by the NRC using the criteria delineated in Appendix A.
These criteria were promulgated in an NRC policy statement which was published in the Federal Register on February 24, 1977 (Vol. 42, No. 37, pages 10950-10952).
In I
order to provide wide dissemination of information to the public, a Federal Register notice is issued on each abnormal occurrence with copies distributed to the NRC Public Document Room and all Local Public Document Rooms. 0; a minimum, each such notice contains the date and place of the occurrence and describes its nature and probable consequences.
The NRC has reviewed Licensee Event Reports, licensing and enforcement actions
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(e.g., notices of violations, civil penalties, license modifications, etc. ),
generic issues, significant inventory differences involving special nuclear matericl, and other categories of informatio'n available to the NRC. The NRC has determined that only those events, including those submitted by the Agree-ment States, described in this report meet the criteria for abnormal occurrence reporting. This report covers the period from April 1 to June 30, 1986.
Information reported on each events 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 NRC 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 i
regulatory process.
t 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 i
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be achieved and maintained through regulations which specify requirements which l
will assure the safe use of nuclear materials. The regulations include design and quality assurance criteria appropriate for the various activities licensed l
by NRC. An inspection and enforcement program helps assure compliance with the I
regulations.
Most NRC licensee employees who work with or in the vicinity of radioactive f
materials are required to utilize personnel monitoring devices such as film badges or TLD (thermoluminescent dosimeter) badges. These baoges are processed i
periodically and the exposure results normally serve as the official and legal record of the extent of personnel exposure to radiation during the period the badge was worn.
If an individual's past exposure history is known and has been sufficiently low, NRC regulations permit an individual in a restricted area to receive up to three rems of whole body exposure in a calendar quarter. Higher j
values are permitted to the extremities or skin of the whole body.
For unre-I 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 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 I
by the nuclear power industry for the detailed review of operating experience to help identify safety concerns early, to improve dissemination of such infor-mation, and to feed back the experience into licensing, regulations, and operations.
In addition, the NRC and the nuclear power industry have ongoing efforts to improve the operational data system which include not only the type, and quality, 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 Nuclear Plant Reliability Data (NPRD) system. The former system is under the control of the NRC while the latter system is a voluntary, industry-supported system operated by the Institute of Nuclear Power Operations (INPO), a nuclear utility organization.
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
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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 1973 a computer-based data file, with data extracted from licensee reports dating from 1969. Periodically, changes were made to improve both the' effectiveness of data processing and the quality of reports required to be submitted by the licensees.
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Effective January 1,1984, major changes were made to the requirements to report to the NRC. A revised Licensee Event Report System (10 CFR S 50.73) was established by Commission rulemaking which modified and codified the former LER system. The purpose was to standardize the reporting requirements for all nuclear power plant licensees and eliminate reporting of events which were of low individual significance, while requiring more thorough documentation and analyses by the licensees of any events required to be reported. All such reports are to be submitted within 30 days of discovery. The revised system l
also permits licensees to use the LER procedures for various other reports l
required under specific sections of 10 CFR Part 20 and Part 50.
The amendment to the Commission's regulations was published in the Federal Register (48 FR 33850) on July 26, 1983, and is described in NUREG-1022, "Licen.
vent Report System," and Supplements 1 and 2 to NUREG-1022.
Also effective January 1, 1984, the NRC amended its immediate notificatu,n re-quirements of significant events at operating nuclear power reactors (10 CFR S 50.72).
This was published in the Federal Register (48 FR 39039) on August 29, 1983, with corrections (48 FR 40882) published on September 12, 1983. Among the changes made were the use of terminology, phrasing, and reporting thresholds that are similar to those of 10 CFR S 50.73.
Therefore, most events reported under 10 CFR S 50.72 will also require an in-depth I
follow-up report under 10 CFR S 50.73.
The NPRD system is a voluntary program for the reporting of reliability data by j
nuclear power plant licensees. Both engineering and failure data are to be 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
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the Commission considered it may be necessary to make participation mandatory in order to male the system a viable tool in analyzing operating experience.
However, on July 8,1981, INP0 announced that because of its role as an active I
user of NPRD system data, it would assume responsibility for management and funding of the NPRD system.
INP0 reports that significant improvements in licensee participation are being made. The Commission considers the NPRO system to be a vital adjunct to the LER system for the collection, review, and feedback of operational experience; therefore, the Commission periodically monitors the progress made on improving the NPRD system.
i Information concerning reportable occurrences at facilities licensed or other-I wise regulated by the NRC is routinely disseminated by the NRC to the nuclear industry, the public, and other interested groups as these events occur.
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 United States and to the NRC Public Document Room in Washington, D.C.
The Congress is routinely kept informed of reportable events occurring in licensed facilities.
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 l
e.
l States assume regulatory authority over byproduct, source and special nuclear materials (in quantities not capable of sustaining a chain reaction).
Comparable and compatible programs are the basit for agreements.
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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.
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i In early 1977, the Commission determined that abnormal occurrences happening at
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facilities of Agreement State licensees should be included in the quarterly reports to Congress. The abnormal occurrence criteria included in Appendix A
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is applied uniformly to events at NRC and Agreement State licensee facilities.
1 Procedures have been developed and implemented and abnormal occurrences j
reported by the Agreement States to the NRC are included in these quarterly reports to Congress.
FOREIGN INFORMATION The NRC participates in an exchange of information with various foreign govern-ments which have nuclear facilities.
This foreign information is reviewed and considered in the NRC's assessment of operating experience and in its research and regulatory activities.
Reference to foreign information may occasionally be made in these quarterly abnormal occurrence reports to Congress; however, only domestic abnormal occurrences are reported.
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X
e REPORT TO CONGRESS ON ABNORMAL OCCURRENCES APRIL-JUNE 1986 NUCLEAR POWER PLANTS The NRC is reviewing events reported at the nuclear power plants licensed to operate during the second calendar quarter of 1986. As of the date of this report, the NRC had determined that the following events were abnormal occurrences.
86-8 Decay Heat Removal Problems at U.S. Pressurized Water Reactors The following information pertaining to this event is also being reported con-currently in the Federal Register. Appendix A (see Example 12 of "For All Licensees") of this report notes that a series of events (where individual events are not of major importance) which create major safety concern can be cor.sidered an abnormal occurrence.
Date and Place - During December 1985, the NRC Office for Analysis and Evaluation of Operational Data (AE00) issued a case study report which analyzed U.S. pres-surized water reactor (PWR) experience involving loss of operating decay heat removal (DHR) systems. The report covered the period from 1976 through 1983, and part of 1984.
Several additional events have occurred through May 1986. One of the more sig-nificant events occurred on March 26, 1986, at San Onofre Unit 2, a ComLustion
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Engineering-designed plant operated by Southern California Edison Company and located in San Diego County, California.
Nature and Probable Consequences - As described in the AEOD case study report
[AE00/C503 (Ref. 1)], the DHR system [also referred to at various plants as the residual heat removal (RHR) system, and shutdown cooling (SDC) system] is de-signed to remove fission product decay heat from the reactor core. The safety function of the DHR system is to remove heat from the. primary system at a rate that will enable operators to bring the plant from hot shutdown conditions to cold shutdown or refueling conditions, and to maintain the plant in such shutdown conditions for extended periods of time.
For the transition phase associated with cooling the plant from operating pressures and temperatures after a reactor trip, for example, to hot shutdown, the steam generators and the auxiliary feed-water system are used to remove heat from the primary system. Upon reaching the reduced pressures and temperatures associated with the hot shutdown condi-tion, the DHR system is activated. Most DHR systems operate at temperatures of 350 F or less, and at pressures less than 600 psig.
Though there are design dif ferences among the various plants, a typical DHR system is composed of two redundant 100% capacity trains, usually located outside containment. Water is taken from a reactor coolant system (RCS) hot leg, flows progressively through suction isolation valves, the DHR pump, the DHR heat exchanger and back to the reactor vessel. During accident conditions, most DHR systems can be aligned to provide low pressure emergency core coolant functions.
Depending upon plant and reactor conditions, an extended loss of the DHR function could lead to bulk boiling, core uncovery, and possible fuel damage unless timely operator actions are taken either to restore the DHR function or to provide 1
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some alternate means of cooling water to the core. The time available for restoring cooling water prior to core urrovery can be short (i.e., an hour or less) or long (i.e., up to many hours or days) depending on the conditions at the time the DHR capability is lost.
The time margin available for restoring the DHR system, or establishing alternate methods of heat removal, depends upon the RCS temperature, the decay heat rate (which is dependent upon time interval elapsed from reactor trip to DHR system failure and core power operating history), and the amount of RCS inventory.
During some shutdown operations, the RCS may be partially drained (e.g., to per-form steam generator inspections or repairs). Decreased primary system inventory can significantly reduce the time available to recover the DHR function prior to bulk boiling and core uncovery.
The AE00 case study was initiated because of the large number of loss-of-DHR events which had occurred at t h Babcock and Wilcox-designed Davis-Besse plant.
During 1980, the plant experienced nine such events.
The NRC reported the events to Congress as Abnormal Occurrence No. 80-5 in NUREG-0090, Vol, 3, No. 2:
" Report to Congress on Abnormal Occurrences: April-June 1980." The basis for reporting was a serious deficiency in management or procedural controls in major areas.
Subsequent analysis showed that DHR system losses were not unique to Davis-Besse or other Babcock and Wilcox-designed plants.
The AE0D case study found that from 1976 through 1983, at least 130 loss-of-DHR system events had been reported to the NRC by five licensees. There also were others which were not reportable under the NRC reporting requirements in effect
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prior to January 1, 1984. At the time of preparing the AE00 report, the 1984 data was incomplete; however, there were at least 10 events in 1984 which were l
considered significant.
For all cases, plant personnel were able to restore the DHR function prior to reaching an unsafe condition (i.e., core uncovery). However, under slightly dif ferent circumstances (e.g., extended delay in restoring the DHR function),
some of the loss of DHR events could have -led to fuel failure and more serious radiological consequences.
While none of the recorded DHR failures affected the health and safety of the l
public, some of the events caused significant plant disruptions, extended down-
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time, and expensive cleanup and recovery.
ause or Causes - The AE00 study concluded that the underlying or root causes of most of the loss-of-DHR system events are human factors deficiencies involving procedural inadequacies and personnel error. Most of the errors were committed during maintenance, testing, and repair operations.
l The categorization of the 130 DHR system failures that occurred at PWRs during the years 1976-1983 is shown in Table 1.
Events involving problems with the suction / isolation valves and the DHR pumps accounted for about 71% of the DHR system failures. Automatic closure of the suction / isolation valves accounted for about 28% of the events.
(These valves are located in the flow path from the reactor coolant system hot leg (s) to the suction side of the DHR pumps.) The underlying or root causes of most of these events were human errors.
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A Table 1 Categories of 130 Reported Total DHR System Failures When Required to Operate (Loss of Function) at U.S. PWRs 1976-1983 No. of Events
(% of Events)
Automatic Closure of Suction /
37 (28)
Isolation Valves Loss of Inventory 36 (28)
Inadequate RCS Inventory Resulting In Loss of DHR Pump Suction (26)
Loss of RCS Inventory Through DHR System Necessitating Shutdown l
of DHR System (10)
Component Failures 29 (22)
Shutdown or failure of DHR Pump (21)
Inability to Open Suction / Isolation Valve (8)
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Others 28 (22)
Total 130 (100.0)
About 28% of all reported DHR system losses involved loss of RCS inventory.
Twenty-six of the loss of inventory events resulted in inadequate' pump suction, cavitation, or air binding. Many events of this type were significant because of their long recovery times.
Recovery required refilling the RCS and bleeding off the air or vapor bound pump (s).
About 22% of the reported DHR losses involved DHR system component failure.
Twenty-one events involved shutdown or random failure of an operating DHR pump when the other pump or train was inoperable. Eight events involved previously closed suction / isolation valves that could not be opened.
Actions Taken to Prevent Recurrence J
Licensees - Corrective actions have generally consisted of various combinations of equipment modifications, changes in procedures and maintenance, and training of personnel, For some plants, the corrective actions taken have not signifi-cantly reduced the occurrence of DHR losses, or the duration of the losses.
Other plants, such as Crystal River and Davis-Besse, have shown considerable improvement starting about the same time that these licensees implemented actions to improve their planning, coordination, and management of outage and maintenance activities.
NRC - The problems of DHR failures have been under review for some time by the NRf Office of Nuclear Reactor Regulation (NRR). Under Unresolved Safety Issue l
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(U5I) A-45, the review is directed toward determining the need for improvements in current DHR systems. Under Generic lasue No. 99, the review is specifically j
directed to loss of the RHR system during cold shutdown or refueling. Both 05f A-45 and Generic I Jue No. 99 have been assigned high priority for resolution.
The AE00 case study includes several recommendations based upon the potential safety significance of loss of-DHR events for significantly improving DHR system l
reliability and availability.
The recommendations include:
improving human factors by upgrading coordination, planning, and administrative control of sur-veillance, maintenance, and testing operations which are performed during shut-down; providing operator aids to assist in determining time available for DHR recovery;and to assist operators' in trending parameters during loss-of-DHR events; upgrading the training and qualification requirements for operations and maintenance staf f; requiring the ute of reliable, well-analyzed methods for measuring reactor vessel level during shutdown modes; modifying plant design to remove autoclosure interlocks and/or power to the DHR suction / isolation valves l
during periods which do not require valve motion; and clarifying plant technical specifications to eliminate' ambiguities associated with operating mode definitions.
NRR has included the AE00 recommendations in their review under Generic Issue No. 99.
i This item is considered closed for the purposes of this report.
Annex k
Further review of Licensee Event Reports submitted to the NRC since the end of 1983 shows that loss of DHR events continue to be a problem area. There were about 15 events during 1984, 18 during 1985, and 7 as of the end of May 1986.
l One of the more significant events during this period occurred on March 26, 1986, at San Onofre Unit 2 caring a ref ueling outage.
Even though the plant had been shut down for 11 days, there still was substantial fission product decay heat.
The reactor water heated up faster than plant personnel expected (the operators believed boiling would not begin before one hour). During the event, the DHR l
system was unavailable for only 49 minutes. Nevertheless, the reactor water was heated to boiling within about 40 minutes after the total loss of DHR flow. [This time agrees well with that shown in Figure 4, "DHR Recovery Time Margin" in the AE00 Case Study Report (Ref. 1).] Boiling continued for about l
l 7 minutes until DBR flow was reestablished. Core uncovery coulo have begun within about another hour. Radioactivity levels in the fuel handling building increased slightly. The levels did not reach hazardous-levels, however, the event did represent an unanticipated risk to plant personnel.
In addition. the event might not have occurred had a problem, which occurred earlier the same day, been properly analyzed and corrected. Details on the event are discussed below. Description of the event also serves to illustrate one of the major l
insulting in loss of DHR pump suction).
1 The plant was shut down on March 15, 1986, for a refueling outage. At 9:50 p.m.
(PST) on March 26, 1986, the DHR pump became inoperable due to entrapment of air. The entrapped air had been pulled into the pump from the DHR suction line s
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which connects to the bottom of the RCS hot leg discharge piping a few feet from the reactor vessel. At ^he time of the event, the RCS was depressurized and vented to atmosphere and the reactor vessel water level had been purposely drained to a level of the middle of the hot leg.
In this condition the horizontal reactor coolant system piping hot leg is half full of water and half full of air.
This condition provides adequate flow path and suction head for the DHR pumps, with water levels sufficiently low to permit removal of steam generator primary side access covers for tube inspections. Slightly lower levels are known, from previous experience, to be susceptible to creating an air / water vortex whir:h can pull air into the DHR suction line at the bottom of the reactor coolant hot leg, possiM y binding the pump with air. Because of this susceptibility, the reactor w.e level and the DHR flow rate need to be carefully controlled when in this condition.
At the time, the operators were unaware that the reactor vessel water leal in-dication was faulty; the level was, in fact,10 inches lower than they believed.
l Therefore, when the operators lowered the water level 12 more inches to support a maintenance request, the motor amperage on the DHR pump began to take large swings indicating air entrapment. The reactor coolant temperature was 115 F at this time. The operators waited three minutes and restarted the pump.
In a few minutes, tne same amperage swings recurred and the pump was stopped. The standby pump was started with similar results.
The operators then followed their abnormal operating instruction and sent equip-ment operators to vent the OHR piping outside the containment penetration area.
Forty-nine minutes after the total loss of DHR flow, the venting was completed and the main DHR pump was restarted. The DHR hot leg temperature increased momen-tarily, during this period, to 210 F, and then dropped to below 200 F.
- However, the temperature indicators are located above the middle of the hot leg; there -
fore, they were not immersed in water and may not have responded in a timely way.
At this time, the ventilation system in the. fuel handling building automatically l
isolated due to high count rate on radiation monitors.
This was due to noble j
gas being released into the containment via the reactor vessel head instrumen.t
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nozzles, and being drawn through the fuel transfer tube by the lower pressure in the fuel handling building. The noble gas came from previously damaged fuel elements (there was no evidence of any fuel damage due to the March 26, 1986 event) and was released from the RCS due to the vessel water temperature increase, and boiling, during the event.
There were no significant effects on the health or safety of the public due to the event. RCS water addition could have been made at any time during the loss of SDC event. A high pressure safety injection pump was used to maintain RCS water level after RCS flow was reestablished. Additional flow paths of RCS makeup water were also available including gravity drain from the refueling water storage tank and safety injection tanks (which were still filled). There was no fuel damage due to the event. Radioactivity released was only about 2 curies (primarily xenon-135), which would result in a total exposure of an individual at the Exclusion Area Boundary of less than 0.002 mrem.
The primary causes of this event were inadequatr uvel indication equipment installation, calibration, and operation; inade at shift turnover; insufficient 5
l operator sensitivity to the importance of reactor water level control; inadequate procedures; and inadequate training. Additionally, a contributing factor was i
the relative ease of inducing vortexing and subsequent false level indications l
inherent in the unit's design, i
The licensee had installed a new reactor vessel water level indication system for this refueling. The system consisted of two electrical level indicators, a wide range and a narrow range indicator. Apparent problems with the system I
were noted on March 19, 1986, during initial draining of the pressurizer.
The problems required recalibration, apparently due to design data errors. Other level accuracy problems were observed during the following few days and on March 22, 1986, the licensee installed the level indication system used during i
the previous refueling, a tygon tube standpipe.
Subsequent to the event, it was determined that the tygon tube had been mounted j
on an old structural stanchion which had reactor water levels marked with a felt tip pen rather than on an engineered scale installed for that purpose.
i Secondly, the tygon tube was installed with an air bubble. The scale used was in error by two inches.
The air bubble's ef fect on accuracy varied dependent upon the bubble's location. The operators were not aware that the tygon tube would indicate water levels that were significantly in error; therefore, they relied on it to verify proper water level in the reactor vessel.
l Despite these problems, the licensee had an opportunity to discover the improper l
water level indication and take corrective actions to preclude the total loss i
of RHR system event. Earlier in the day of the event, the day shift had begun drain-down, noted that the tygon and elect rical level indications were not in i
agreement, and had noticed the RHR pump amperage start to oscillate (indicating i
vortexing). The day-shift immediately added water until indications of air entrapment ceased. These occurrences were not communicated to the oncoming shift, the shif t that later experienced the loss of shutdown cooling, because the out-going shift did not believe them to be significant in relation to plant conditions at the time of shift turnover.
Actions Taken to Prevent Recurrence Licensee - The licensee has taken, or will take actions to establish a more I
reliable indication of water level. This is to be done by classifying the sys-tems as safety-related and applying the more stringent associated administrative controls; by controlling the configuration of the level indicating systems with more detailed design documentation; by inproving procedures for installation; and by establishing criteria for verification of operability. Additionally, the licensee is reconsidering a design change to make the RHR pump self-venting.
Consideration is being given to developing a correlation of shutdown cooling flow and reactor water level to identify vortexing regions. Further, consideration is being given to extensive operator training concerning RHR operation and to operational procedure improvements.
Implementation of several of these improve-ments and a significant increase in management and operator control were observed during a subsequent drain-down to reestablish mid-loop conditions on April 22, 1986.
NRC - Upon being notified of the event, the NRC Resident Inspectors began an
~examination of the event, including the root causes.
The inspection report, together with a Notice of Violction, were sent to the licensee on June 20, 1986 (Ref. 2).
NRC Region V will continue to review the licensee's corrective actions.
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t Unless new significant information becomes available, this item is considered closed for the purposes of this report.
A A A A A A A A A A 86-9 Out-of-Sequence Control Rod Withdrawal at Peach Bottom Unit-3 The following information pertaining to this event is also being reported concurrently in the Federal Register. Appendix A (see Example 11 of "For All Licensees") notes that serious deficiency in management or procedural controls in major areas can be considered an abnormal occurrence.
Date and Place - On March 18, 1986, during a startup of Peach Bottom Unit 3, numerous personnel errors by four licensed operators resulted in a control rod being withdrawn out-of-sequence without being detected by these operators. The next operating shift detected the error and manually scrammed the unit. Peach Bottom Unit 3 is a General Electric-designed boilina. water reactor (BWR) operated by Philadelphia Electric Company (the licensee) and located in York County, Pennsylvania.
Nature and Probable Consequences - To limit reactivity additions during startup and low power operations, BWRs utilize a variety of procedural controls and monitoring systems. The reactor operator follows a control rod withdrawal sequence procedure developed by reactor engineers. A rod worth minimizer (RWM) computer system monitors this sequence and blocks control rod movement upon significant deviation from the prescribed sequence. A rod sequence control
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system (RSCS) complements and backs up the RWM in restricting control rod move-ment. A major purpose of these controls is to limit the consequences of a control rod drop event during the startup.
On March 17, 1986, a startup was begun on Peach Bottom Unit 3.
The RWM was bypassed due to a computer hardwere iault. As allowed by technical specifica-tions, a second licensed reactor operator ;as assigned to monitor another licensed reactor operator as the latter withdrew control rods in accordance with the sequence prescribed by procedure. At 1:28 a.m. on March 18, 1986, the operator withdrew cor. trol rod 10-23 out of sequence instead of rod 02-23.
The second operator monitoring the rod withdrawals failed to notice the error.
Later, at its prescribed place in the procedure, both operators signed off the withdrawal of control rod 10-23.
Shortly afterward, the reactor attained cri-ticality. At 2:30 a.m., withdrawal of additional control rods in an attempt to l
increase reactor power was blocked by the RSCS since rod 02-23 was not in its prescribed position. The Shift Superintendent and Shift Supervisor then bypassed the RSCS rod 02-23 full out logic with a keylock switch without verifying the rod position and conformance to the rod withdrawal sequence as required by the procedure for bypassing RSCS logic.
Rod withdrawal and startup continued with rod 02-23 fully inserted instead of being fully withdrawn as required.
Af ter 7:00 a.m., the oncoming shif t requested that the RWM be returned to service. This was ' accomplished at 7:38 a.m.; the operators noted an insert error for rod 02-23. The rod was confirmed to be out of position for the sequence.
The Shif t Supervisor returned the RSCS bypass for control rod 02-23 to normal.
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Two control rods were inserted and then the reactor was manually scrammed from approximately 3% power at 8:55 a.m.
The NRC Senior Resident inspector and Duty Of ficer were notified of the scram and the out-of-sequence rod shortly af terward, j
l The licensee presented an analysis of potential consequences of a rod drop event I
for various rod pairs for the March 18 event. The peak enthalpy deposition in a I
fuel pin, had a rod drop occurred with rod 02-23 inserted, for the worst case was calculated to be 118 cal /gm. This is less than the peak enthalpy deposition of 215 cal /gm from the reload analysis for the current fuel cycle and the 280 cal /gm l
design criterion.
Cause or Causes - The out-of sequence control rod withdrawal resulted from numerous personnel errors by four licensed operators. One licensed reactor j
operator withdrew the wrong control rod from the core. The RWM, designed to i
detect such an occurrence, was inoperable. A second licensed operator was assigned to independently verify the correct rod withdrawal sequence as required by Peach Bottom Technical Specifications; he did not identify the error. When the point in the sequence to withdraw the rod already incorrectly withdrawn was reached, neither reactor operator identified the previius error. The Shift Supervisor and Shif t Superintendent who were overseeing the startup activities failed to note the error.
Further, they bypassed the RSCS without assuring that the bypassed control rod was in its correct position, as required by the procedure f or use of the bypass keys.
These personnel errors by four licensed individuals showed inattention to detail, failure to adhere to procedural requirements, and complacent attitudes.
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Actions Taken to Prevent Recurrence Licensee - The four individuals involved in this event were disciplined.
Plant staff management meetings were held with all operations personnel to discuss the event and their individual responsibilities.
Procedural controls were strengthened to, among other things, use best ef forts to place the RWM in ser-vice, dedicate a second operator to sequence verification if RWM is bypassed, generate rod position maps at specific withdrawal points and compare with prepared
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rod position maps and require positive rod position verification prior to RSCS l
bypass.
NRC - After notification of the out-of-sequence control rod withdrawal by the IIcensee on March 18, a special safety inspection into the event was conducted at Peach Bottom Atomic Power Station on March 18-21, 1986. The inspection re-sults were forwarded to the licensee in a letter dated March 25, 1986 (Ref. 3).
An enforcement conference was held at NRC Region I on March 27, 1986, between NRC and licensee personnel to discuss the causes of the event and the corrective actions taken and planncd. A Notice of Violation and Proposed Imposition of Civil Penalty in the amount of $200,000, dated June 9, 1986, was sent to the licensee (Ref. 4) describing the violations resulting from numerous personnel errors by several licensed personnel.
The enforcement letter stated that these personnel errors indicate that a pattern of inattention to detail, failure to adhere to procedural requirements, and a generally complacent attitude of staff toward performance of their duties con-tinues to exist at Peach Bottom.
Since 1983, the licensee has been cited three times, and civil penalties imposed, for violations pertaining to personnel not 8
following procedures. The'iatest incident demonstrates that the actions taken to correct this pattern have not been effective. Such problems are indicative of a lack of management involvement in and attention to station activities to assure that the station personnel respect, understand the need for, and adhere to licensee policies :nd procedures for the safe operation of the facility.
The proposed civil penalty of $200,000 represented a 100% escalation because:
(1) in each case, an opportunity existed for a licensed individual to detect and correct the rod pull error, but the error was not recognized, and (2) the enforcement history at Peach Bottom regarding personnel adherence to procedures l
has been poor.
Unless new significant information become available, this item is considered closed for the purposes'of this report.
86-10 Boiling Water Reactor Emergency Core Cooling System Design Deficiency The following information pertaining to this event is also being reported con-currently in the Federal Register. Appendix A (see the third general criterion) of this report notes that major deficiencies in design, construction, use of, or management controls for licensed facilities or material can be considered an abnormal occurrence. In addition, Example 12 of "For All Licensees" of Appen-dix A notes that incidents with implications for similar facilities (generic incidents) which create major safety concern can be considered an abnormal occurrence.
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Date and Place - On May 19, 1986, the Boston Edison Company (BECO) notified the NRC that a significant design deficiency in the residual heat removal (RHR) system minimum flow protection logic at the Pilgrim Nuclea-Power Station (PNPS) had been discovered. Pilgrim is a General Electric (GE)-designed boiling water reactor (BWR) located in Plymouth County, Massachusetts.
l Later, it was found that some other GE-designed BWRs also contained the same design deficiency.
l Background - The RHR system, which operates at low pressure, functions in dif-f erent modes to remove reactor decay heat under normal and emergency situations (e.g., loss of coolant accident, LOCA).
For normal situations, the RHR system can be operated in the shutdown cooling / head spray mode and the steam condensing i
mode. For emergency situations, the RHR system functions as part of the emer-gency core cooling system (ECCS) operating in the containment spray / cooling mode and the low pressure coolant system (LPCI) mode.
At Pilgrim, to prevent the RHR pumps from running dead headed, each pair of pumps is provided with a minimum flow bypass capability. The minimum flow bypass consists of an orifice flow bypass which allows a flow of approximately 10 per-cent of rated flow to bypass the reactor vessel and be returned to the suction source. The minimum flow bypass line for each pair of RHR pumps is connected to a single line and controlled by a single minimum flow bypass valve. The minimum flow bypass valves are normally open. The valves will close upon sensing flow in either of the RHR loops.
Prior to the deficiency found at Pilgrim, the NRC had issued on December 13, 1985, Inspection and Enforcement Information Notice No. 85-94, " Potent'ial for Loss of
t Minimum flow Paths Leading to ECCS Pump Da.. age During a LOCA" (Ref. 5) to all nuclear power reactor facilities holding an operating license or a construction permit. The Notice alerted licensees of recent instances at various BWR and pressurized water reactor (PWR) plants where it was discovered that minimum flow requirements might not or could not be met for some ECCS pumps under small-break LOCA conditions.
The Notice suggested that licensees review the informa-tion for applicability to their facilities and consider actions, if appropriate,
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to preclude similar problems at their facilities.
Nature and Probable Consequences - While reviewing Information Notice No. 85-94 i
f or applicability to Pilgrim, BEC0 discovered that a single failure under cer-tain accident sequences could result in all RHR minimum flow bypass valves being signaled to close while all other pump discharge valves also closed. The current logic configuration for Pilgrim minimum flow bypass valves is that a high flow signal from either the A or B RHR loops will close both A and B bypass valves.
Thus, a postulated single failure of a flow sensing instrument may result in all RHR pumps running without flow. This condition would lead to the pumps run-ning dead headed with potential for pump damage in a few min?.es.
The design deficiency is significant to public health or safety because if this single failure occurred in conjunction with an automatic start of the RHR system, RHR pump damage may occur if unrecognized by the operator.
This event could disable RHR functions including LPCI, head spray, drywell spray, shutdown
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cooling, torus spray and suppression pool cooling. As a result of the loss of suppression pool cooling over a long period of time, core spray pumps could ultimately lose net positive suction head and also be unavailable. Thus,
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systems designed to remove decay heat under both normal and accident condi-tiors could be severely degraded.
j Cause or Causes - The deficiency is attributed to error in design of the logic for RHR mini-flow control. The NRC Staff believes that the error was most i
likely the result of inadequate consideration of the effects of mini-flow iso-lation during design of logic for determining the proper (intact) coolant loop into which emergency core coolant would be injected following a loss of coolant accident.
Actions Taken to Prevent Recurrence i
Licensee - Following discovery of the deficiency, all operating shifts at Pilgrim were briefed on the problem.
In addition to this, the licensee performed a design review of other ECCS systems to determine if they had a similar defi-ciency. The results of the review confirmed that only the RHR minimum flow l
protection logic was deficient.
Since the plant was in the cold shutdown l
condition at the time of the discovery of the deficiency, no other immediate mitigative actions were deemed necessary. The licensee is currently evaluating I
short-term and long-term modifications to correct the problem.
In a May 30, 1986, letter to the NRC Staff, the licensee made a commitment to implement the short-term modifications prior to restart from the current outage.
NRC - The NRC Staff was first notified of the deficiency by the licensee in a report made to the NRC Headquarters Operation Center per 10 CFR S50.72 on May 19, 1986. A followup letter, pursuant to 10 CFR 921.21, was submitted on May 23, 1986. Followup discussions between the staff, licensee, and GE indicated j
a high likelihood that other GE BWRs in operation also had the problem.
10 l
I FUEL CYCLE FACILITIES (Other Than Nuclear Power Plants) l The NRC is reviewing events reported by these licensees during the second calendar quarter of 1986. As of the date of this report, the NRC had not determined that any events were abnormal occurrences.
1 A A A A R A A R A A l
OTHER NRC LICENSEES l
(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, industrial, and academic fields.
Incidents were reported in this category from licensees such as radiographer, medical institutions, and byproduct material users.
The NRC is reviewing events reported by these licensees during the second calen-dar quarter of 1986. As of the date of this report, the NRC had determined that j
the following events were abnormal occurrences.
86-12 Willful Failure to Report a Diagnostic Medical Misadministration
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The following information pertaining to this event is also being reported concurrently in the Federal Register. Appendix A (see the general criterion) of this report notes that an event involving a moderate or more Severe impact on public health or safety can be considered an abnormal occurrence.
In addition, Example 11 of "For All Licensees" of Appendix A notes that serious deficiencies in management or procedural controls can be considered an abnormal occurrence.
Date and Place - On May 8, 1985, a patient at Mercy Hospital, Wilkes-Barre, Pennsylvania, received an injection of a radiopharmaceutical (a diagnostic dose of technetium-99m) intended for another patient. The misadministration was willfully not reported to the NRC as required by 10 CFR S35.43.
l Nature and Probable Consequences - An anonymous allegation was received by NRC Region 1 on May 8, 1985.
The alleger stated that a misadministration had occurred that morning at Mercy Hospital when the Chief Nuclear Medicine Techni-l cian mistakenly injected the wrong patient with a radiopharmaceutical. Further, i
the alleger stated that the misadministration would not be reported to the NRC.
The required report of the misadministration was due to the NRC by July 10, 1985.
On July 17, 1985, two NRC Region I inspectors performed a routine unannounced inspection and followup of the allegation at the licensee's facility. During the inspection, the Chief Nuclear Medicine Technician stated that no misadmini-strations had occurred since the one reported to the NRC in 1984. However, the inspectors noted that records showed one patient had received two radiopharma-l ceutical injections in a one hour period on May 8, 1985. The Chief Nuclear Medicine Technician stated that this was not because of a misadministration.
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i On August 7, 1985, an investigator.from the NRC's Office of Investigations (01) went to Mercy Hospital. During an interview with the Chief Nuclear Medicine Technician, she admitted she had lied to the NRC on July 17, 1985.
The Chief Nuclear Medicine Technician also stated she was told that the Medical Director of Radiology, who is also the licensee's Radiation Safety Of ficer (RS0), did not want the misadministration reported. The RSO stated during an interview with the 01 investigator on August 7, 1985, that he had informed some of his staff not to report the misadministration.
The consequences of the licensee's actions in this incident are that (1) it decreases the NRC's confidence that this licensee will report incidents as required by regulM.:n and (2) it delays implementation of procedures to prevent further vi', administrations of a similar nature.
The ef fects on the patient, mistakenly receiving the radiopharmaceutical, would be expected to be small due to the relatively low levels of exposure invo.ved. However, it did represent an unnecessary exposure.
Cause or Causes - The cause is due to the deliberate failure of the RSO to follow the required procedure for repcrting misadministration and instructing the hospital staff not to report this particular misadministration.
Actions Taken to Prevent Recurrence Licensee - The licensee, as well as another licensee in which the RS0 is involved, requested an extension to respond to the NRC enforcement actions
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described below.
NRC - On June 17, 1986, the NRC forwarded to Mercy Hospital (1) an Order Tequiring the licensee to show cause why the Chief Nuclear Medicine Technician and the RSO should not be prohibited from the performance or supervision of any licensed activities, and (2) a Notice of Violation and Proposed Imposition of Civil Penalty in the amount of $5,000 (Ref. 13).
The RSO at Mercy Hospital is also listed as an authorized user of NRC licensed material on the license of Valley Radiology Associates, Inc., Kingston, Pennsylvania. Therefore, on June 17, 1986, the NRC issued a similar Order to his licensee (Ref. 14).
1 Future reports will be made as appropriate.
- aaaa*****
i 86-13 Therapeutic Medical Misadministration
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The following information pertaining to this event is also being reported concurrently in the Federal Register. Appendix A (see the general criterion) of this report notes that an event involving a moderate or more severe impact l
on public health or safety can be considered an abnormal occurrence.
Date and Place - On April 9, 1986, at Maryview Hospital, Portsmouth, Virginia, a patient received a therapy dose in a chemical form other than that intended.
This resulted in an unintended dose of several hundred rads to the patient's bone marrow.
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i Nature and Probable Consequences - A physician asked the Nuclear Medicine i
Department to order a dose of phosph,rus-32 as colloidal chromic phosphate on April 7,1986, for administration to 0 kidney carcimoma patient for abdominal ascites reduction on April 9, 1986. Tt.is verbal order was relayed to Nuclear Medicine through a third party, and the chemical form of phosphorus-32 was not made clear. Nuclear Medicine proceeded to order 15 mi7iicur ies of phosphorus-32 as sodium phosphate because this chemical form was used more frequently than the colloidal form. The order was processed ano received in the rospital in the normal manner.
On April 9, 1986, the physicist drew up the dose in a syringe, assived it in i
the dose calibrator, and then put it aside. Shortly thereafter, a physician j
(other than the physician who ordered the dose) administered the dose ;.ntra-peritoneally to the patient.
Later the same day, the Chief Nuclear Medicine
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Technologist, while discussing this particular patient with a nurse, discovered that the dose was administered intraperitoneally rather than intravenously, as intended by the physician who ordered the dose. This information was relayed to several physicians and was also reported later that day to the NRC.
l On April 10, 1986, the patient was administered stable phosphorous to accelerate excretion of the phosphorous-32. Blood counts for leucocytes, red blood cells, l
hematocrits and platelets showed no significant depression as of April 21, 1986.
The consequences of the misadministration was a significant unintended dose to i
the patient's bone marrow. The licensee estimated the dose to be at least 150 rads. However, the NRC's medical consultant believes the dose could have been as much as 700-800 rads to the patient's bone marrow with an increased I
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chance of the patient contracting leukemia.
The misadministration constituted a significant failure to comply with NRC regulatory requirements. The patient was subjected to a procedure unrelated to the authorized uses of phosphorus-32 as sodium phosphate.
Cause or Causes - The root cause was the lack of written prescriptions for i
ordering therapeutic doses.
Actions Taken to Prevent Recurrence Licensee - The licensee established written procedures and forms to provide for written prescriptions and therapeutic radionuclides procedures. The licensee's agreement to establish procedures for ordering and administering therapy doses had been previously documented in an NRC Confirmation of Action Letter, dated April 10, 1986.
NRC - In addition to engaging a medical consultant and issuing the Confirmation of Action Letter, the NRC Region 11 conducted a special inspection at the hos-pital on April 11, 1986. An Enforcement Conference with the licensee was held on May 2, 1986, to discuss NRC concerns regarding the inspection findings. At the conference, the licensee presented the previously mentioned written proce-i dures and forms.
On August 7, 1986, the NRC issued to the licensee (1) a Confirmatory Order Modifying License, and, (2) a Notice of Violation and Proposed Imposition of Civil Penalty in the amount of $2,500 (Ref. 15). The Order, effective 16
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i Actions Taken to Prevent Recurrence i
Licensee - All concerned personnel have been retrained on the policy of not ad--
j ministering radioisotopes without a written requisition and of the requirement i
to obtain the specific consent of a radiologist for all cases requiring the ad-
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ministration of' greater than 300 microcuries of iodine-131.
NRC - The incident is being reviewed by an NRC medical consultant. NRC Region I I
pTans to conduct an inspection after the consultant's review is complete.
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future reports will be made as appropriate.
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l 86-16 Diagnostic Medical Misadministration l
The following information pertaining to this event is also being. reported con-I 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 June 17, 1986, at the Tripler Army Medical Center, Tripler i
AMC, Hawaii, a patient received a dose of 3.09 mci of I-131 instead of a pre-l scribed dose of 50 uci for a thyroid imaging procedure.
l Nature and Probable Consequences - A 54 year old female patient was given a
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3.09 mei dose of I-131 by mistake. The patient was scheduled for a thyroid l
imaging procedure which utilizes only 50 uti,of I-131, The radiation exposure s
received by the patient due to the 3.09 mci 1-131 dose is estimated to be 2472 rad to the thyroid, 0.43 rad to the ovaries, and 1.45 rad to the whole body.
Contact with the licensee was made on July 9,1986, regarding any possible clini.
cal symptoms or adverse health effects due to the 3.09 mci 1-131 dose. The licensee stated that tne patient had been hospitalized for observation. On July 6,1986, the patient was discharged due to the lack of any clinical symptoms.
The patient has been scheduled for 90 days interval checkups at her duty station on Guam. An annual medical workup has also been scheduled. The high exposure i
to the thyroid may result in some degree of impairment in its function.
Cause or Causes - Th.; misadministration was the result of an isolated incident of misreading the consultation sheet.
l Actions Taken to Prevent Recurrence Licensee - Effective immediately, the dispensing procedure for radioactive iodine is as follows:
(a) In.all cases', the final dispensing and checking of the dose will be done by a staf f physician or radiology resident assigned to Nuclear Medicine Service.
1 (b) The identification of the patient as well as the' final amount dispensed will be co-signed by the physician involved in that particular procedure.
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(c) The quality assurance manual for Nuclear Medicine Service is being updated to stipulate the new review procedures.
NRC - The circumstances of the misadministration were discussed in detail with LIIe licensee on July 3, 1986 by a member of the NRC Region V management staff.
The licensee's corrective actions appear to be acceptable.
The NRC will not issue any further requirements in this matter at this time. The matter will be i
reviewed again during the next inspection.
This item is considered closed for the purposes of this report.
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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 j
report the events to the NRC for inclusion in this report. During the second
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calendar quarter of 1986, the Agreement States reported the following abnormal occurrences to the NRC.
i AS86-5 Uncontrolled Release of Krypton-85 to an Unrestricted Area l
Appendix A (see the first general subcriteria) of this report notes that moder-ate exposure to, or release of, radioactive material can be considered an ab-normal occurrence.
In addition, Example 3 of "For All Licensees" of Appendix A 4
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of this report notes that the release of radioactive material to an unrestricted l
l area in concentrations which, if averaged over a period of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, exceed l
500 times the.egulatory limit of Appendix B, Table II, 10 CFR Part 20 (10 CFR l
1 S 20.403(b)), can be considered an abnormal occurrence.
Date and Place - On May 8, 1985, during routine operation of a Trio-Tech
" Tracer-Flo" system at Micro-Rel Division, Medtronic, Incorporated, of Tempe, Arizona, a malfunction occurred which caused approximately 11.2 curies of l
radioactive krypton-85 to be vented into the atmosphere.
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Nature and Probable Consequences
" Tracer-Flo" systems are used to test sealed objects, such as electronic components, to determine whether they are, in fact, j
sealed. Tested objects are placed in a test chamber which is evacuated and a mixture of nitrogen and radioactive krypton-85 is introduced. This m!xture is then removed and replaced by air. The objects are then tested for residual l
radioactivity.
If none is detected, this would indicate that the objects have been properly sealed.
l On May 8,1985, during routine operation of the " Tracer-Flo" fine leak system, the unit " locked" into the first cycle of operation. The unit then began to run thruugh the other cycles while maintaining the mechanical cor.ditions of the first cycle. This situation resulted in the krypton-85 being released to an unrestricted area, rather than being retained within the system. There was no evidence that any overexposure occurred.
Cause or Causes - A thorough inspection of the machine was made and all mechani-cal systems were found to function properly. The failure was attributed to the l
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i machine's logic board. This was concluded by a step-by-step replacement of in-l tegrated circuits on the logic P.C. board until control panel indications were normal. The unit was then cycled a number of times and found to work properly.
Actions Taken to Prevent Recurrence Licensee - Even though the licensee has an exemplary maintenance program, it would not have prevented this type of release. The P.C. board logic failure can only be rectified by design changes by the manufacturer.
State Agency - The Agency monitored the licensee's response to this event and confirmed completion of the actions described above.
The Agency performed an inspection of the circumstances associated with the event and the licensee was assessed a civil penalty in the amount of $3,000. Due to the licensee's good l
past history and cooperation with the Agency, the civil penalty was mitigated l
to $1,500, which was imposed upon and paid by the licensee.
This item is considered closed for the purposes of this report.
A A A A A A A A A A l
AS86-6 Contaminated Radiopharmaceutical Used in Diagnostic Administrations 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 l
considered an abnormal occurrence.
k Date and Place - On May 9,1985, a breakthrough of molybdenum-99 (a radioactive contaminant) occurred in a molybdenum-99/ technetium-99m generator at Scripps Memorial Hospital of Encinitas, California.
The breakthrough went unrecognized and the contaminated technetium-99m radiopharmaceutical was administered to four patients scheduled for diagnostic medical tests.
iherefore, these patients received exposures higher than necessary.
Nature and Probable Consequences - Technetium-99m is a radiopharmaceutical which is widely used in hospitals and doctors' of fices for diagr,using a variety of diseases.
It has a short halflife of 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> (i.e., it loses half of its radio-activity every E hours).
It is a product of the decay of another radioactive material, molybdenum-99.
The technetium-99m producing devices, called generators, contain molybdenum-99.
Technetium-99m, the short-lived product, is removed from the generator as needed by using a saline solution which combines with the technetium-99m, but leaves most of the molybdenum-99 in place. Molybdenum-99 has no medical application and is considered a contaminant; NRC requirements permit no more than 5 micro-curies molybdenum-99 contaminant in a dose of technetium-99m.
The generator was purchased by New England Nuclear by Nuclear Pharmacy, Inc.
and used to process unit doses in the San Diego, California area.
Later the generator was sold to Scripps Hospite.1 (a State of California licensee). After a few days of use of the generator. the licensee's nuclear medicine scanning equipment developed anomalies which made the scanning results useless (no image, but with indications of a high energy background).
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i After determining that the scanning equipment was not at fault, the licensee suspected molybdenum-99 breakthrough. A physician at the San Diego Veterans Administration Hospital confirmed the presence of the contaminant. He estimated
-liver doses to the patients ranging from 130 rads 10 260 rads. As discussed fur-ther below, it is believed that DTPA was inadvertently used, rather than the saline solution, for removing technetium-99m from the generator. Therefore, due to possible rapid clearance of the DTPA from the body, the actual doses may have been less than those estimated.
Blood test results of the patients were reported to be normal, perhaps because the material may not have deposited in the vascular compartment.
The whole body dose for each patient was estimated to be a few mrad. The nuclear medicine physician at the hospital reported in January 1986 that "no adverse effects have been identified in any of the four patients."
l Cause or Causes - Af ter many milkings of the generator with normal eluants, it l
appears that DlPA, a chelating agent, was inadvertently used in place of the usual saline solution (the vials were almost identical). This DTPA removed a
. substantial amount of the molybdenum-99 from the column. After the fact tests estimate that as much as 1.0 mci /cc may have toded up in the doses. Secondly, although molybdenum-99 breakthrough testing was routinely performed, it appears that the nuclear medicine technologist observing the dose calibrator readings had come to ignore which indicator light was lit,. i.e., millicurie or microcurie and to simply record the digital readout assuming it was microcurie.
There is a practical certainty that the calibrator was indicating millicuries which should have been noticed by the technologist.
Actions Taken to Prevent Recurrence Licensee - Upon suggestion of Mo99 breakthrough, the generator was taken out of service and affected patients were identified. The dose calibrator which had been independently checked and calibrated only one month earlier was reapproved by the licensee's consultant. All succeeding molybdenum-99 and aluminum break-through safety checks were confirmed by either a second nuclear medicine techni-
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cian or nuclear medicine physician.
Later, the hospital discontinued the use of generators and began using bulk technetium-99m. But tests for molybdenum-99 breakthrough were continued as a precautionary measure.
State Agency - The event was investigated during an onsite visit by the Agency.
The licensee was cited under one of its license conditions for failure to per-form adequate molybdenum-99 breakthrough tests on the generator eluant.
This report by the Agency was considerably delayed because the Agency's medical consultant, who was asked to evaluate the patients' doses, provided vastly dif-ferent (and lower) estimates than the hospital physician but did not provide further inf ormation to explain the discrepancies. Having received no response from the consultant to its ir.quiries, the Agency has accepted the hospital physician's dose estimates.
This item is considered closed for the purposes of this report. - -
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REFERENCES 1.
Case Study Report, " Decay Heat Removal Problems at U.S. Pressurized Water Reactors," AE00/C503, prepared by the NRC Office for Analysis and Evaluation of Operational Data, December 1985.*
2.
Letter from A. E. Chaffee, Chief, Reactor Projects Branch, NRC Region V, to Kenneth P. Baskin, Vice President, Nuclear Engineering Safety and Licensing, Southern California Edison Company, forwarding (1) a Notice of Violation and (2) Inspection Report Nos. 50-361/86-11 and 50-362/86-11, Docket Nos. 50-361 and 50-362, June 20, 1986.*
3.
Letter from Richard W. Starostecki, Director, Division of Reactor Projects, NRC Region I to Mr. S. L. Daltroff, Vice President, Electric Production, Philadelphia Electric Co., forwarding Inspection Report No. 50-278/86-09, Docket No. 50-278, March 25, 1986.*
4.
Letter from Thomas E. Murley, Regional Administrator, NRC Region I, to Mr. S. Daltroff, Vice President, Electric Production, Philadelphia Electric Co., forwarding a Notice of Violation and Proposed Imposition of Civil Penalty, Docket No. 50-278, June 9,1986.
- 5.
U.S. Nuclear Regulatory Commission, Inspection and Enforcement Information Notice No. 85-94, " Potential for Loss of Minimum Flow Paths Leading to ECCS Pump Damage During a LOCA," December 13, 1985.*
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6.
U.S. Nuclear Regulatory Commission, Inspection and Enforcement Compliance Bulletin No. 86-01, " Minimum Flow Logic. Problems that Could Disable RHR Pumps," May 23, 1986.*
7.
Letter from James G. Keppler, Regional Administrator, NRC Region III, to Dr. F. W. Buckman, Vice President, Nuclear Operations, Consumers Power Company, enclosing the " Systematic Assessment of Licensee Performance Report for the Palisades Nuclear Generating Station," Docket No. 50-255, Februa ry 12, 1986.*
8.
Confirmatory Action Letter (CAL)-RIII-85-15 from James G. Keppler, Regional Administrator, NRC Region III, to R. B. DeWitt, Vice President, Nuclear Operations, Consumers Power Company, Docket No. 50-255, October 30, 1985.*
9.
Letter from James G. Keppler, Regional Administrator, NRC Region III, to Dr. F. W. Buckman, Vice President, Nuclear Operations, Consumers Power Company enclosing Task Force Report dated May 1, 1986, for Palisades Nuclear Generating Station, Docket No. 50-255, May 16, 1986.*
10.
Letter from James G. Keppler, Regional Administrator, NRC Region III, to Dr. F. W. Buckman, Vice President, Nuclear Operations, Consumers Power Company, forwarding (1) a Notice of Violation and Proposed Imposition of Civil Penalty and (2) Inspection Report No. 50-255/86008 (DRP), Docket No. 50-255, July 1, 1986.*
- Available in NRC Public Document Room, 1717 H Street, NW, Washington, DC 20555, for inspection and copying (for a fee).
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11.
Confirmatory Action Letter (CAL)-RIII-86-032 from James G. Keppler,
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Regional Administrator, NRC Region III, to Dr. F. W. Buckman, Vice President, Nuclear Operations, Consumers Power Company, Docket I
No. 50-255, May 21, 1986.*
12.
Letter from Charles E. Norelius, Director, Division of Reactor Projects, NRC Region III, to Dr. F. W. Buckman, Vice President, Nuclear Operations, Consumers Power Company, forwarding Augmented Investigation Team Report No. 50-255/86017 (DRP), Docket No. 50-255, June 30, 1986.*
13.
Letter from James M. Taylor, Director., NRC Office of Inspection and Enforcement, 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.*
1 14.
Letter from James M. Taylor, Director, NRC Office of Inspection and l
Enforcement, to Salvatore M. Imperiale, M.D., Director of Nuclear j
Medicine, Valley Radiology Associates, Inc., forwarding an Order to Show Cause Why the License Should Not Be Modified, Docket No. 30-15110, l
June 17, 1986.*
15.
Letter from James M. Taylor, Director, NRC Of fice of Inspection and Enforcement, to J. Bland Burkhardt, Administrator, Maryview Hospital,
[ forwarding (1) Confirmatory Order Modifying License, (2) Notice of 1
Violation and Proposed Imposition of Civil Penalty, (3) Letter to all
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Medical Licensees from V. L. Miller (NRC) dated May 2, 1979, and (4) NRC Inspection Report No. 45-10831-02/86-01], Docket No. 030-03347, August 7, 1986.*
16.
Letter from James M. Taylor, Director, NRC Office of Inspection and Enforcement, to Roland E. Kohr, President, Bloomington Hospital, forwarding an Order to Show Cause Why License should Not Be Suspended and i
Modified (Effective Immediately), Docket No. 030-01644, April 22, 1986.*
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APPENDIX A 1
ABNORMAL OCCURRENCE CRITERIA The following criteria for this report's abnormal occurrence determinations were set forth in an NRC policy statement published in the Federal Register on February 24, 1977 (Vol. 42, No. 37, pages 10950-10952).
An event will be considered an abnormal occurrence if it involves a major reduc-tion in the degree of protection of the public health or safety. Such an event
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would involve a moderate or more severe impact on the public health or safety and could include but neeu not be limited to:
1.
Moderate exposure to, or release of, radioactive material licensed by or otherwise regulated by the Commission; l
2.
Major degradation of essential safety-related equipment; or 3.
Major deficiencies in design, construction, use of, or management controls for licensed facilities or material.
Examples of the types of events that are evaluated in detail using these criteria 1
are; I
l For All Licensees
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1.
Exposure of the whole body of any individual to 25 rems or more of radia-tion; exposure of the skin of the whole body of any individual to 150 rems or more of radiation; or exposure of the feet, ankles, hands or forearms of any individual to 375 rems or more of radiation (10 CFR S20.403(a)(1)), or equivalent exposures from internal sources.
2.
An exposure to an individual in an unrestricted area such that the whole-body dose received exceeds 0.5 rem in one calendar year (10 CFR S20.105(a)).
l 3.
The telease of radioactive material to an unrestricted area in concentra-tions which, if averaged over a period of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, exceed 500 times the regulatory limit of Appendix B. Table II, 10 CFR Part 20 (10 CFR S20.403(b)).
l 4.
Radiation or contamination levels in excess of design values on packages, or l
loss of confinement of radioactive material such as (a) a radiation dose rate of 1,000 mrem per hour three feet from the surface of a package con-taining the radioactive material, or (b) release of radioactive material from a package in amounts greater than the regulatory limit.
5.
Any loss of licensed material in such quantities and under such circum-stances that substantial hazard may result to persons in unrestricted areas.
6.
A substantiated case of actual or attempted theft or diversion of licensed material or sabotage of a facility.
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5 7.
Any substantiated loss of special nuclear material or any substantiated in-ventory discrepancy which is judged to be significant relative to normally expected performance and which is judged to be caused by theft or diversion or by substantial breakdown of the accountability system.
8.
Any substantial breakdown of physical security or material control (i.e.,
access control, containment, or accountability systems) that significantly weakened the protection against theft, diversion, or sabotage.
9.
An accidental criticality (10 CFR S70.52(a)).
10.
A major deficiency in design, construction, or operation having safety impli-cations requiring immediate remedial action.
11.
Serious deficiency in management or procedural controls in major areas.
12.
Series of events (where individual events are not of major importance), re-curring incidents, and incidents with implications for similar facilities (generic incidents), which create major safety concern.
For Commercial Nuclear Power Plants 1
1.
Exceeding a safety limit of license technical specifications l
(10 CFR S50.36(c)).
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Major degradation of fuel integrity, primary coolant pressure boundary, or l
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primary containment boundary.
1 3.
Loss of plant capability to perform essential safety functions such that a potential release of radioactivity in excess of 10 CFR Part 100 guidelines could result from a postulated transient or accident (e.g., loss of emer-j I
gency core cooling system, loss of control rod system).
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4.
Discovery of a major condition not specifically considered in the safety analysis report (SAR) or technical specifications that requires immediate remedial action.
l S.
Personnel error or procedural deficiencies which result in loss of plant 1
l capability to perform essential safety functions such that a potential re-lease of radioactivity in excess of 10 CFR Part 100 guidelines could result from a postulated transient or accident (e.g., loss of emergency core cool-ing system, loss of control rod system).
_Fo,r Fuel Cycle Licensees l
1.
A safety limit of license technical specifications is exceeded and a plant shutdown is required (10 CFR S50.36(c)).
2.
A major condition not specifically considered in the safety analysis report or technical specifications that requires immediate remedial action.
3.
An event which seriously compromised the ability of a confinement system to perform its designated function.
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APPENDIX B UPDATE OF PREVIOUSLY REPORTED ABNORMAL OCCURRENCES 1
During the April through June 1986 period, the NRC, NRC licensees, Agreement i
States, Agreement State Licensees, and other involved parties, such as reactor vendors and architects and engineers, continued with the implementation of ac-tions necessary to prevent recurrence of previously reported abnormal occurrences.
The referenced Congressional abnormal occurrence reports below provide the ini-tial and any updating information on the abnormal occurrences discussed. Those occurrences not now considered closed will be discussed in subsequent reports in the series.
NUCLEAR POWER PLANTS 79-3 Nuclear Accident at Three Mile Island This abnormal occurrence was originally reported in NUREG-0090, Vol. 2, Nol. 1,
" Report to Congress on Abnormal Occurrences: January-March 1979," and updated in each subsequent report in this series, i.e., NUREG-0090, Vol. 2, No. 2 through Vol. 9, No. 1.
It is further updated as follows.
Reactor Building Entries During the second calendar quarter of 1986, 91 entries were made into the TMI-2
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reactor building, bringing the total number of entries since the March 1979 acci-dent to 956.
Reactor building activities during this period centered around (1) controlling microorganism populations in the reactor coolant system (RCS) and thereby improving water clarity, (2) pick and place defueling, and (3) the use of a drilling rig to obtain stratified samples of the damaged core. Addi-tional reactor building entries were made to remove two of the six internal
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vent valves and perform maintenance on the,various systems needed for the cleanup effort.
Reactor Vessel Defueling Operations As discussed in the previous update report, visibility in the reactor vessel progressively decreased during the first quarter of calendar year 1986 to a point at which defueling activities were being adversely affected. Microorganisms (algae, fungi, bacteria, and anaerobes) were initially identified as the cause of filter plugging in the Defueling Water Cleanup System which rapidly made i
this system become ineffective in maintaining water clarity. The licensee I
conducted a three phase program of chemical treatment, high pressure flushing ard filtration of the RCS. At the end of April 1986 the Temporary Reactor Vessel Filtration System, using a large diatomaceous earth filter, was put into operation. High pressure hydrolancing of the RCS began in early May 1986.
Also in May, the licensee began injecting borated water treated with hydrogen peroxide into'the RCS via the Standby Pressure Control System. The licensee established an initial 200 ppm residual concentration of peroxide as a biocide to kill the microorganisms. As a result, water visibility was improved-to over 27 i
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.i one foot. The licensee is continuing the program on a periodic basis to control the populations of microorganisms and maintain water clarity.
Pick and place defueling (using remote tools to move damaged fuel and core debris into defueling canisters), halted during the microorganism control prcgram, resumed on May 23, 1986. Near the end of June, pick and place _ defueling was
- again halted to ready the work platform for installation of the drilling rig to obtain stratified core samples. One additional canister was transferred to the i
fuel pool at this time bringing the total number of canisters transferred out of the reactor building to 43.
The total weight of debris removed from the
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reactor vessel thus far is 51,670 lbs., representing nearly 17% of the 308,000 lbs, of debris estimated to be in the vessel.
Core boring to obtain stratified samples of the core began on July 3, 1986.
This effort was designed to obtain full length samples of the reactor core from 3
the surface of the debris bed to within inches of the inner surface of the~ lower head of the reactor vessel.
Information gained from drilling regarding the hardness, ductility, and friability of the core material will be used in plan-ning future defueling activities. The core samples (approximately 2 1/2" in diameter and 8 ft. long) will be shipped in canisters to the Idaho National Engineering Laboratory (INEL) for examination.
Dose rates associated with defueling activities have continued to remain low.
Dose rates on the defueling work platform average approximately 8 mrem /hr, and the highest measured dose rates during canister transfer from the reactor vessel to the storage racks have been less than 40 mrem /hr.
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EPICOR-II Submerged Demineralized System (SDS) Processing Approximately 212,474 gallons of water were processed through the 505 during the reporting period. Approximately 148,468 gallons were processed through the EPICOR-II system during the quarter.
Cask and Liner Shipments There were no of fsite shipments of EPICOR-II or SDS liners. At'the end of ' June, seven defueling canisters stored in the fuel storage pool were loaded into the first rail-mounted shipping cask. The total weight of the debris contained in the seven canisters is 2400 lbs. The first shipment is expected to occur some-time in July 1986. Additional shipments are expected over the next two and one-half years.
Auxiliary and Fuel Handling Building Decontamination activities continued during the second quarter of 1986. Activ-ities included; scabbling and painting of the floors in the neutralizer tank rooms and the mini decay heat pump room; steam-cleaning of the tendon access gallery; scabbling in the reclaimed boric acid tank room; vacuum cleaning of the 281' elevation; and water flushing of the Westinghouse valve room.
f 28
The ACRS Ad Hoc Subcommittee on TVA met with TVA and the Staff on June 12-13, 1986, to discuss the Corporate Nuclear Performance Plan and receive a briefing on the root causes of the current situation and corrective actions underway.
The full ACRS met with TVA and the Staff on July 10, 1986. ACRS concerns include training of TVA managers, the transition from contractor managers back to TVA employee managers, and the span of control of the Manager of Nuclear Power.
TVA has not yet responded to the Commission's evaluation of intimidation and harassment (I&H) at TVA which was forwarded to TVA on June 2, 1986. The Staff is reviewing all TVA 1&H issues that were received prior to March 1986 to deter-mine if they are being tracked and followed up as appropriate by the TVA General Counsel, Inspector General, or Commission staff. The TVA Inspector General (TVA IG) has expressed a need to contact concerned individuals directly, as part of his investigative process.
The Staff (including the NRC Office of Investigations) is developing a method to contact selected concerned individuals in an attempt to establish communications between these individuals and the TVA IG.
Equipment Qualification The TVA certification of completion of environmental qualification actions at Sequoyah has not been received. As TVA has completed portions of the equipment qualification (EQ) ef fort, the Staff continues to conduct onsite reviews.
The most recent was conducted in late June 1986 and a follow-up inspection is planned for August 1986.
Discrepancies identified by both the Staff and TVA at Sequoyah remain to be
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corrected; however, the TVA EQ Program appears to be sound. The Staff believes that the major EQ discrepancies which need to be resolved by TVA before they will be able to certify that Sequoyah is in ' ompliance with the EQ rule involve c
Raychem electrical cable splices and the environmental effects of certain high energy line breaks.
TVA and the Staff will face additional effort to assure EQ compliance at Browns Ferry and Watts Bar.
l Employee Concerns i
Nearly 5,000 TVA employee concerns have been raised at the Watts Bar facility.
Some of these involve safety-related equipment or intimidation, harassment, or l
wrongdoing issues. About 400 of these concerns apply to the Sequoyah facility.
TVA has established an internal program for evaluating employee concerns to replace the former program managed by Quality Technology Company (Q1C). TVA l
committed to provide the NRC with a new submittal to describe in detail their l
new program for dealing with employee concerns, but the submittal date is un-known. The Staff will evaluate (on a sampling basis) the TVA resolution to these concerns during the coming months (beginning in August 1986).
l TVA and QTC terminated their contract in April 1986 at which time the Staff ob-l tained copies of all of the QTC generated employee concerns records. The Statf has nearly completed the screenir.g and expurgation effort on the QTC employee concerns files to identify safety-related information considered by NRC as needed by TVA to resolve individual concerns. This information is being trans-mitted to the TVA IG for further screening prior to release to the TVA line i
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organizations. To date, the Staff screening has identified about 100 poten-tially new safety-related issues.
TVA has regrouped the Watts Bar employee concerns to facilitate evaluation and resolution.
In early June, the Staff inspected TVA progress.
Based on limited review, the Staff remains concerned about the details and depth of the TVA evaluation process. The Staff has developed review guidelines and made person-nel assignments to support NRC technical review of the resolution process.
i Welding Staff review of welding at Sequoyah is nearing completion and TVA's actions appear to be acceptable. However, TVA needs to respond to the Staff's questions and approve the related employee concerns investigation reports in order for the Staff to complete its review.
TVA continues to reinspect welds at Watts Bar.
In late June, the NRC Staff met with TVA to discuss their program. While the Staff considers the logic of the ov.erall program to be basically sound, the docketed program is deficient in that it contains insufficient detailed information to permit an adequate tech-nical review.
Some of the key information needed includes sample size and acceptance criteria, applicability of portions of the ASME Code, certain quality assurance aspects related to welding, and detailed project procedures. NRC inspection of this activity is ongoing and will focus on:
Independent examinations of hardware EG&G inspector qualification / certification records
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1 TVA's engineering calculations I
Implementation of EG&G's reinspection efforts Design Control TVA is developing their program for evaluating the adequacy of design control at Sequoyah.
The design control program will be submitted as part of the forthcoming revision to the Sequoyah Nuclear Performance Plan.
Inspection of TVA progress in this area was conducted in June 1986 and periodic inspections y
are planned in the coming months. This may be a pacing item for Sequoyah restart.
Quality Assurance i
l Quality assurance (QA) issues involve, primarily, TVA instruction activities at Watts Bar. Numerous allegations regarding the QA program have been raised and are under review by TVA and the Staff. The Staff is reviewing a May 1, 1986 revision to the TVA QA Topical Report and met with TVA on this report in early July.
Licensed Operator Requalification l
In response to the unsatisfactory performance of Browns Ferry operators on the NRC-administered requalification examination given in November 1985, TVA has implemented an extensiYe program to upgrade Browns Ferry licensed operator knowledge and capabilities. The first group (of three groups) completed training in June. The new retraining program appears to be acceptable to the Staff and implementation of the program is being evaluated.
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future reports will be made as appropriate.
A A R R A A A A A A 86-1 Loss of Power and Water Hammer Event This abnormal occurrence, which occurred at San Onofre Unit 1 on November 21, 1985, was originally reported in NUREG-0090, Vol. 9, No. 1, " Report to Congress on Abnormal Occurrences: January-March 1986."
It is updated as follows.
As discussed in the report of this abnormal occurrence, the licensee's program for addressing issues stemming from the November 21, 1985 incident was origi-nally presented to the NRC Staff on April 8, 1986. Additional information was submitted on May 1 and May 5, 1986. The Staff has completed their technical review of the licensee submittals and prepared safety evaluation reports.
The licensee and NRC Staff met with the Commissioners on June 12, 1986. The Commissioners reviewed the licensee's and Staff's actions taken to analyze and correct the deficiencies that led to the event.
Principal among these actions was the development and implementation of a Material Condition Review Program.
The program was designed to define a suitable material standard for systems and components in an older plant and to ensure that the material condition of those items was maintained.
On June 25, 1986, the Region V Administrator visited the site to assess the material condition of San Onofre Unit 1.
Although minor deficiencies were l
identified, the Unit condition was found to be greatly improved from what it had been prior to the November 21, 1985 event.
l Based on the evaluations by the NRC Office of Nuclear Reactor Regulation, the Of fice of Inspection and Enforcement, and the NRC Region V Office, the Staff concluded that the licensee's plant improvements for resolution of the water hammer event were adequate for facility restart subject to certain procedural changes and additional commitments to which the licensee agreed.
On July 13, 1986, the Region V Administrator concurred with the restart of Unit 1.
The reactor was taken critical on July 15, 1986, for low power physics testing in preparation for return of the plant to service.
I NRC Inspection and Enforcement Information Notice No. 86-49 was issued on June 16, 1986, to all nuclear power reactor facilities holding an operating license or a construction permit (Ref. B-9).
The Notice described age / environment failures of electrical cables at San Onofre Unit 1 (which contributed to the November 21, 1985 event). The Notice also suggested actions which licensees could take to improve in-service cable reliability.
This item is considered closed for the purposes of this repo'rt.
j maa*a aaaa=
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86-2 Loss of Integrated Control System Power and Overcooling Transient This abnormal occurrence, which occurred at Rancho Seco on December 26, 1985, was originally reported in NUREG-0090, Vol. 9, No. 1, " Report to Congress on Abnormal Occurrences, January-March 1986."
It is updated as follows.
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t On June 12, 1986, Sacramento Municipal Utility District (the licensee) met with the NRC Staff to discuss the scope and content of the preliminary action plan for performance improvement.
The intent was to obtain NRC comments so that they could be incorporated into the final report. As a result of the meeting, an NRC team visited the Rancho Seco site on June 17-18, 1986, to review in detail the method being utilized by the licensee to define its system review and test program. The team found that the licensee had a structured program underway to find and correct problems with hardware, procedures, maintenance, etc. However, work on defining the system review and test program had just started and little information was available on program scope or content.
The licensee submitted its action plan for performance improvement on July 3, 1986, and the Staff has initiated its review. The Staff plans to have a working meeting with the licensee to discuss the action plan in August. The critical path to restart decision is the Staff's review of the action plan and issuance of the Staff's safety evaluation report.
Future reports will be made as appropriate.
A A A A A A A A f A FUEL CYCLE FACILITIES j
J 86-3 Rupture of Uranium Hexafluoride Cylinder and Release of Gases
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This abnormal occurrence, involvin'g Sequoyah fuels Corporation, Gore, Oklahoma,
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was originally reported in NUREG-0090, Vol. 9, No.1, " Report to Congress on Abnormal Occurrences: January-March 1986." It is updated as follows.
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By letter dated May 8, 1986, Sequoyah Fuels Corporation requested authorization l
to restart production of UFs. However, this request was not accompanied by com-mitments that could be incorporated into the license. Accordingly, by letter l
dated May 23, 1986, the NRC Staff requested that additional information be pro-vided. A response to this request was received by letter dated June 25, 1986, and the review of the information is in progress. NRC has also held two public meetings in Gore, Oklahoma, to solicit the concerns of local individuals so that these concerns can be taken into consideration during the restart review.
The NRC Staff estimates that action will be taken on the restart request during the latter part of the third calendar quarter of 1986.
As discussed in NUREG-0090, Vol. 9, No.1, a Lessons-Learned Group was reviewing the event as well as regulatory practices in general regarding such fuel facili-l ties. This Group was formed on February 20, 1986, by the NRC Acting Executive Director for Operations to prepare a report based on experience gained from i
this event. The goal of the Group was to identify actions NRC might reasonably i
take from a licensing and inspection standpoint to prevent similar incidents, as well as to clarify NRC's regulatory role regarding facilities of this type.
A further goal was ta assess the adequacy of the NRC response to the incident, as well as the follow-on activities.
The Group's report was issued during June 1986 as NUREG-1198 (Ref. 8-10). The report presents discussions and recommendations on the process, operation and design of the facility, as well as on the responses of the licensee, NRC, and i
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I APPENDIX C OTHER EVENTS OF INTEREST The following items are described below because they may possibly be perceived 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 safety; therefore, they are not reportable as abnormal occurrences.
1.
Reactor Vessel Indications at Oconee Unit 1
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During March 1986, Duke Power Company (the licensee) notified the NRC that
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unacceptable indications had been found during an inservice inspection (ISI) at Oconee Unit 1.
The plant is a Babcock & Wilcox (B&W)-designed pressurized water reactor located in Oconee County, South Carolina. The pressure vessel was manufactured by B&W.
During the 151, the reactor vessel to flange shell weld was ultrasonically 3
examined from the flange face.
Twenty-two indications were recorded, all from j
the clad side of the flange. Only one of the indications extended onto the unclad side of the flange.
The indications were located from 24 to 32 inches from the flange.
Previous examinations conducted from the vessel inner and outer diameter surfaces as well as the unclad side of the flange face did not detect these same indications.
The licensee stated that with the exception of one geometric reflector recorded in 1979, there were no recordable indications en this weld. The 1986 examination, however, was the first examination
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conducted from the clad side of the flange face.
j Based on the reported size, number and ultrasonic amplitude of the flaw j
indications in a reactor vessel, other licensees have performed additional i
inspections to further confirm and characterize flaw indications.
Duke Power Company, however, elected to complete the ISI with the available examination data, install the vessel closure head and perform a fracture mechanics evaluation of the flaw indications using the measured dimensions.
Initially, the licensee evaluated the data as though the indications represented real flaws since there was no evidence to prove otherwise. Nine of these indications were acceptable when compared to the ASME Section XI acceptance standard IWB-3510. The remaining 13 indications were considered acceptable by the licensee by analytical methods permitted by IWB-3600 in accordance with Appendix A (Section XI) procedures.
Because the licensee decided to resolve the issue by this analytical method, i.e., Linear Elastic Fracture Mechanics Analysis, NRC Region 11 requested that the NRC Office of Nuclear Reactor Regulation (NRR) review the licensee's fracture mechanics evaluation.
Two meetings were held with the licensee in the NRC Region 11 office on April 8 and April 21, 1986 during which the licensee presented their inspection results, fracture mechanics evaluation and their proposed action plan. At the April 8, 1986 meeting, the licensee informed the NRC Staff that some of the reflectors may be geometric in origin. This preliminary conclusion was based on a limited qualitative laboratory test on a reactor vessel mock up at the licensee's Mt. Vernon, Indiana facility. However, the licensee. intended to 41
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consider the flaw indications as actual flaws and to complete the fracture mechanics evaluation based on the requirements of ASME Section XI
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Subarticle IWB-3600.
The Staff reviewed the licensee's submittals, along with additional information requested during the April 21, 1986 meeting, and concluded that Oconee Unit 1 may restart and operate for eight weeks on the condition that the licensee submit, before June 16, 1986, results on the following work effort:
Perform a comparison of Units 1 and 3 vessel flange geometry, material (s) and cladding and determine why Unit 3 ISI examination of the same weld did not produce similar results, Conduct an ultrasonic examination study on the Mt. Vernon mock-up reactor vessel, and 1
Review the original weld design and fabrication history including l
nondestructive examination records of the weld in question.
In addition, the Staff stated that the NRC reserves the right to re-evaluate its position on the above, in the event that a significant transient occurred on the Unit 1 vessel.
In a letter dated April 24, 1986, the licensee submitted their fracture mechanics evaluation and a summary of the ultrasonic testing results which
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located and sized the flaw indications.
NRR has concluded that Oconee Unit 1 can be safely returned to full power and operated with actual flaws of the size and circumferential locations described in the safety evaluation report (SER). This conclusion was based on the Staff's review of the fracture mechanics analysis evaluation performed by the licensee, I
the pressure temperature limits, and Pressurizer Code Safety Valve Set Point contained in the Oconee Unit 1 Technical Specifications. This conclusion was supplemented with the following conditions:
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The licensee will submit prior to June 16, 1986, a technical report summarizing their on going ultrasonic testing program on the Mt. Vernon mock-up vessel, The Staff will review and determine whether the conclusion that the subject flaw indications are enveloped by the dimensions measured by the licensee is still conservative, Because the Staff considers several of the flaws as conditionally acceptable per IWB-3122.4, three augmented ISIS based on 10 CFR S50.55.a(g)(4) will be required, and At least six months before the next scheduled refueling outage, the licensee will provide a report describing detailed plans for the above l
augmented ISIS.
l On June 13, 1986 the licensee submitted to NRR a report with the requested information. This report stated that preliminary results, from.the data 42
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obtained at Mt. Vernon, point to a high probability that the data recorded during the Oconee Unit 1 examination indicated geometric conditions.
The report provided a detailed description of the examination of the Mt. Vernon vessel mock-up using the actual calibration block and similar examination
-1 equipment to that used during the Oconee Unit 1 examination, and results from
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those examinations.
Specific recommendations for conducting ultrasonic 1
examination (s) from the flange face of the reactor vessel in future ISIS were provided. These will be employed in the upcoming Oconee Unit 2 ISI to be performed in August 1986.
The licensee's report is currently undergoing review by the Staff, i
The licensee has returned the plant to power operation. This event had no impact on the public health or safety; therefore, it is not considered reportable as an abnormal occurrence.
- A A A A A A A 2.
NRC Augmented Inspection Team Sent to Pilgrim On April 12, 1986, an NRC Augmented Inspection Team (AIT) was sent to the Pilgrim Nuclear Power Plant after events occurring on April 4 and 12, 1986, raised some concern regarding safe operation of the plant. Pilgrim, operated by Boston Edison Company (the licensee), utilizes a General Electric-designed boiling water reactor. The plant is located in Plymouth County, Massachusetts.
l On April 4 and 12, 1986, the reactor scrammed from low power during routine 1
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reactor shutdowns. Both scrams were caused by unexpected Group I primary con-tainment isolations.
In both cases the isolation signal was promptly reset, but the four outboard main steam isolation valves (MSIVs) could not be promptly reopened. As a result, the main condenser was not available as a heat sink during a portion of each event and the high pressure coolant injection (HPCI) system was operated in the test mode to' control most of the subsequent reactor cooldown.
In regard to the April 4 scram, a routine reactor shutdown had been initiated earlier on the same day af ter oil leakage was detected in the main turbine control air system.
In regard to the April 12 scram, a routine reactor shutdown had been initiated on April 11 af ter periodic residual heat removal (RHR) system high pressure alarms were received on both April 10 and 11 indicat-l ing that the RHR system was being pressurized by reactor coolant leakage past various valves. An Unusual Event was declared because of the RHR valve leakage.
Following the second scram, NRC Region 1 issued Confirmatory Action Letter (CAL) No. 86-10 to the licensee on April 12, 1986, which required that all affected equipment be maintained in its as found condition (except as necessary to maintain the plant in a safe shutdown condition) in order to preserve any evidence which would be needed to inspect or reconstruct the events (Ref. C-1).
The AIT conducted a special safety inspection from April 12 through April 25, 1986, to review the circumstances associated with the areas of concern, i.e.,
(1) the spurious Group 1 primary containment isolations, (2) the failure of the MSIVs to promptly reopen after the isolations, and (3) the recurring i
pressurization of the RHR system due to valve leakage. Based on the AIT j
l 43 I
inspection, the team noted that the licensee's problem solving approaches were carefully structured and appeared thorough.
In addition, the team drew the following conclusions for the three areas of concern:
No root causes for the spurious primary containment isolations on April 4 and 12, 1986, were identified during the inspection period, despite considerable licensee effort. The team did not identify any weaknesses in the licensee's problem solving approach.
The failure of the MSIVs to re-open following the containment isolations on April 4 and 12 was caused by partial or complete mechanical separation of the valve pilot poppets from the MSIV valve stem assemblies. Pilot poppet set screws did not prevent the poppets from unscrewing from the i
stem assemblies.
The RHR pressurization events reflect slow leakage (about 0.5 gpm) past a I
check valve and two motor operated injection valves in the "B" RHR loop.
Lack of RHR pressure instrumentation and the lack of periodic tests of the RHR injection check valves inhibit a more thorough diagnosis. No apparent RHR valve failure mechanism has been identified as the reason for this leakage.
I The licensee's conduct of the reactor shutdown on April 11 and 12, 1986, l
was prudent in light of the recurring RHR pressurization events.
1 l
The licensee's root cause evaluations were not completed and corrective
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actions were not finalized during the AIT inspection. NRC review of these
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actions should be conducted prior to startup from this outage.
1 The above AIT findings are documented in Inspection Report No. 50-293/86-17 which was forwarded to the licensee on May 16, 1986 (Ref. C-2).
The previously mentioned NRC Region I CAL also required that the licensee submit a written evaluation to the NRC of the events noted above prior to restart and to seek Regional Administrator authorization for restart. The plant remained shutdown for the entire second quarter of 1986.
The actual safety consequences of the events were minimal as the isolation system and MSIV failures were in the conservative direction and there was no indication that the potential existed for a sudden overpressurization and failure of the RHR piping. Therefore, based on the information developed to date, the item is considered below the threshold for abnormal occurrence reporting.
It is being reported here because of (1) the item may be perceived by the public to be of public health or safety significance, (2) a special NRC j
AIT was organized and sent to the plant to review the events and licensee actions, (3) the considerable ef fort expended by the licensee and the.NRC in resolving the issues, and (4) the extensive shutdown being experienced by the plant.
In addition, the leakage past valves forming the isolation barrier between the i
high pressure reactor coolant system and the low pressure piping of the RHR system has been a recurrent problem in BWRs. This problem was addressed in Abnormal Occurrence No. 84-8 in NUREG-0090, Vol. 7, No. 3 (" Report to Congress on Abnormal Occurrences: July September 1984") and updated in Appendix B of NUREG-0090, Vol. 8, No. 2 (" Report to Congress on Abnormal Occurrences:
44 l
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i April-June 1985"); an NRC Office for Analysis and Evaluation of Operational l
Data (AE00) Case Study (AE0D/C502) issued in September 1985 (Ref. C-3); and Inspection and Enforcement Information Notice Nos. 84-74 and 84-81 issued on September 28, 1984, and November 16, 1984, respectively (Refs. C-4 and C-5).
The february 1986 problem at Pilgrim was included in Inspection and Enforcement Information Notice No. 86-40 which was issued on June 5, 1986 (Ref. C-6).
- A A A A A A A A A 3.
Construction Problems at Comanche Peak l
On May 2, 1986, the NRC Director of Inspection and Enforcement transmitted to Texas Utilities Electric Company (the licensee) two separate Notices of Viola-tion and Proposed Imposition of Civil Penalties in the amount of $250,000 as a result of multiple failures to meet requirements pertaining to the construction and quality assurance program at Comanche Peak Units 1 and 2 (Ref. C-7), and in the amount of $120,000 as the result of failure in three instances to assure quality control inspectors the organizational freedom to identify problems (Ref. C-8).
These issues are discussed separately below.
The Comanche Peak Steam Electric Station (CPSES) Units 1 and 2 are Westinghouse-desigried pressurized water reactors and are located in Sommervelle County, Texas.
Construction and Quality Assurance
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The NRC has devoted substantial resources to evaluating the adequacy of con-struction at the CPSES facility.
In addition to the routine and special inspec-tions conducted by NRC Region IV, a Construction Appraisal Team inspection was conducted by the Office of Inspection and Enforcement on January 24 - February 4,1952, and February 14 - March 3,1983.
From April 13 - 18, 1984, a review by the Special Revies Team (SRT) was conducted by representatives of NRC Region
- 11. Subsequently, the NRC Technical Review Team (TRT) was assembled which con-sisted of approximately 50 specialists from NRC headquarters, NRC regional offices, and consultants, to evaluate and reselve technical issues and issues identified as a result of allegations. The results of the review of the issues by the TRT are documented in Safety Evaluation Report (SER) NUREG-0797, Supple-ments 7, 8, 10, and 11 (Ref. C-9).
These issues include failure to ensure that quality control inspectors were properly qualified and certified, ineffective interactions between the various engineering and construction groups, deficien-cies in the quality control inspection program, and failure to properly imple-ment the site's corrective action program.
In response to these issues, the licensee has issued a " Comanche Peak Response Team Program Plan and Issue-Specific Actions Plan" to describe planned corrective actions (Ref. C-10).
Two separate special inspections were conducted on November 18 - December 18, 1985, and January 1 - March 4, 1986, respectively, by Region IV, concerning the Unit 1 as-built cable tray inspection program (Ref. C-11) and the procurement and installation of electrical containment penetration assemblies (Ref. C-12) furnished by the Bunker Ramo Corporation. These two inspections identified failures to properly inspect the Unit I cable trays and. deficiencies in the procurement and installation of electrical containment penetration assemblies.
45
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On June 25, 1986, the licensee was granted an extension in responding to the TRT issues and Region IV special inspections (Ref. C-13). The licensee response (Ref. C-14) is currently under review by Region IV personnel.
Organizational ~ Freedom I
As a result of numerous allegations of intimidation, harannent, and discrimin-
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ation, and the relevance of this issue to the contentions in the ongoing operat-ing license hearing, the NRC undertook a comprehensive review and taaluation of the allegations of intimidation, harassment, and discrimination at COSES. A report prepared by an NRC Comanche Peak Intimidation Panel (Panel) aided by a Study Team of consultants was issued on November 4, 1985 (Ref. C-15).
The NRC staf f has reviewed the Panel report, the completed Department of Labor discrimi-nation cases regarding CPSES, the NRC Office of Investigation reports, and the licensee's responses regarding intimidation at CPSES. Three incidents were identified as violations by the NRC staff as a result of this review.
On June 2, 1986, the licensee responded to the violations concerning intimida-tion, harassment and discrimination and requested further NRC staff considera-i tion with regard to two of the violations (Ref. C-16).
The response is under l
review by the NRC staff.
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1 Because the above issuas occurred during facility construction and there was no fuel in the reactors, there were no effects on public health or safety.
Therefore, the issues are not reportable as an abnormal occurrence. They are being reported here because (1) they may be perceived by the public to be of
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public health or safety significance, (2) considerable resources were expended in evaluating the issues, and (3) the considerable civil penalties which have been imposed.
- A A A A A A A A A 4.
Sabotage of Offsite Power Lines to Palo Verde j
i Palo Verde is a three unit nuclear power station operated by Arizona Public j
Service Company and located in Maricopa County, Arizona. Each unit is'a Combustion Engineering-designed pressurized water reactor.
On May 14, 1986, beginning at approximately 8:59 p.m., MST, of fsite power on 3 of 4 transmission lines was lost due to phase-to ground faults on each line.
The power remained faulted through the evening until the afternoon of May 15, 1986. During the morning hours of May 15, the licensee found that a rope or strap had been slung over the transmission lines causing a phase-to ground fault. The fault location on each lina was about 35 miles from the plant.
The transmission lines converge on the site from four different directions.
The ground straps were far enough apart to require a coordinated effort by several individuals to accomplish the power loss within a few minutes. Unit 1 was in Mode 5 (cold shutdown) for a maintenance outage. Unit 2 remained in Mode 3 (hot standby) pending completion of the investigation of the event.
Unit 3 was in the preoperational test phase, with no fuel in the reactor.
The licensee notified local law enforcement agencies and the FBI.
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REFERENCES FUR APPENDICES l
B-2 Confirmatory Action Letter from James C. Keppler, Regional Administrator, NRC Region III, to Wayne H. Jens, Vice President-Nuclear Operations, Detroit Edison Company, Docket No. 50-341, July 16, 1985.*
B-2 Letter from Charles E. Norelius, Director, Division of Reactor Projects, NRC Region III, to Wayne H. Jens, Vice President-Nuclear Operations, Detroit Edison Company, forwarding Inspection Report No. 50-341/85040 (DRP), Docket No. 50-341, January 7, 1986.*
B-3 Letter from James M. Taylor, Director, NRC Office of Inspection and Enforcement, to Walter J. McCarthy, Jr., Chairman of the Board and Chief Executive Officer, Detroit Ediscn Company, forwarding an Order 4
Modifying the License Effective Immediately and a Notice of Violation and Proposed Imposition of Civil Penalty, Docket No. 50-341, July 3, 1986.*
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B-4 Letter from A. E. Chaffee, Chief, Reactor Operations Branch, NRC Region V, to Kenneth P. Baskin, Vice President, Nuclear Engineering Safety and Licensing, Southern California Edison Company, forwarding a Notice of Violation.and Inspection Reports Nos. 50-361/86-11 and 50-362/86-11, Docket Nos. 50-361 and 50-362, June 20, 1986.*
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B-5 U.S. Nuclear Regulatory Commission, " Loss of Main and Auxiliary Feedwater Event at the Covis-Besse Plant on June 9, 1985," USNRC Report i
l NUREG-1154, published July 1985.**
B-6 10 CFR S 50.54(f) letter from Harold R. Denton, Director, NRC Office of Nuclear Reactor Regulation, to Joe Williams, Jr., Senior Vice President, Nuclear, Toledo Edison Company, Docket No. 50-346, August 14, 1985.*
l B-7 Letter from John P. Williamson, Chairman and Chief Executive Officer, Toledo Edison Company, to Harold R. Denton, Director, NRC Office of Nuclear Reactor Regulation, Docket No. 50-346, September 10, 1985.*
B-8 U.5. Nuclear Regulatory Commission " Safety Evaluation Report related to the restart of Davis-Besse Nuclear Power Station, Unit 1, following the event of June 9, 1985," USNRC Report NUREG-1177, published June 1986.**
- Available in NRC Public Document Room, 1717 H Street, NW, Washington, DC 20555, for inspection and copying (for a fee).
- 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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B-9 U.S. Nudeer Regulatory Commission, Inspection and Enforcement I
Information Notice No. 86-49, " Age / Environment Induced Electrical Cable Failures," June 16, 1986.*
l B-10 U.S. Nuclear Regulatory Commission, " Release of UFs from a Ruptured Model 48Y Cylinder at Sequoyah fuels Corporation facility: Lessons -
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,ed Report," USNRC. Report NUREG-1198, published June 1986.**
B-11 Letter from James M. Taylor, Deputy Director, NRC Office of Inspection and Enforcement, to Mark T. Hebner, President and Chief Executive Officer, Syncor International Corporation, forwarding an Order Imposing a Civil Monetary Penalty, License No. 34-18309-01MD, January 2, 1985.*
B-12 Letter from James M. Taylor, Deputy Director, NRC Office of Inspection and Enforcement, to Nuclear Pharmacy, Inc., forwarding an Order Modifying Licenses (effective immediately), License Nos. 12-18044-01MD, 14-19990-01MD, 20-21227-01MD, 37-18461-01MD, 37-19586-01MD, 37-21322-01, and 48-174C5-01MD, October 26, 1984.*
B-13 Letter from James M. Taylor, Director, NRC Office of Inspection and Enforcement, to Monty Fu, Chairman of the Board and Chief Executive Officer, Nuclear Pharmacy, Inc., forwarding a Notice of Violation and Proposed Imposition of Civil Penalties, License Nos. 37-21322-01, 12-18044-01MD,14-19990-01MD, 20-21227-01MD, 37-18461-01MD, 37-19586-01MD, and 48-17466-01MD, April 10, 1986.*
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B-14 U.S. Nuclear Regulatory Commission, Inspection and Enforcement Information Notice No. 86-54, " Criminal Prosecution of a Former Radiation Safety Officer Who Willfully Directed an Unqualified Individual to Perform Radiography," June 27, 1986.*
C-1 Confirmatory Action 8.etter No. 86-10 f rom Thoma:, C Murley, Regional Administrator, NRC Pegion I, to William D. Harrington, Senior Vice President, Nuclear, Boston Edison Company M/C-Nuclear, Docket No. 50-293, April 12, 1986.*
C-2 Letter from Richard Starostecki, Director, Division of Reactor Projects, NRC Region I, to William D. Harrington, Senior Vice President, Nuclear, Boston Edison Company M/C Nuclear, forwarding Augmented Incident Response Team Report (Inspection Report No. 50-293/86-17), Docket No. 50-293, May 16, 1986.*
C-3 U.S. Nuclear Regulatory Commission, Case Study Report, "Overpressurization of Emergency Core Cooling Systems in Boiling Water Reactors," prepared by NRC Office for Analysis and Evaluation of Operational Data.
Preliminary report, dated February 1985 was. issued March 7, 1985
- Final report (designated as AEOD/C502) was issued in September 1985.*
C-4 U.S. Nuclear Regulatory Commission, Inspection and Enforcement Information Notice No. 84-74, " Isolation of Reactor Coolant System from Low-Pressure Systems Outside Containment," September 28, 1984.*
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C-5 U.S. Nuclear Regulatory Commission, Inspection and Enforcement Information Nntice No. 84-81, " Inadvertent Reduction in Primary Coolant Inventory in Boiling Water Reactors During Shutdown and Startup,"
November 16, 1984.*
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C-6 U.S. Nuclear Regulatory Commission, Inspection and Enforcement Information Notice No. 86-40,'" Degraded Ability to Isolate the Reactor Coolant System from Low-Pressure Coolant Systems in BWRs," June 5, 1986.*
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C-7 Letter from James M. Taylor, Director, NRC Office of Inspection and Enforcement, to W. G. Counsil, Executive Vice President, Texas Utilities Electric Company, forwarding two Notices of Violation and Proposed Impostion of Civil Penalties, Enforcement Action (EA) 86-09, Docket Nos. 50-445 and 50-446, May 2, 1986.*
C-8 Letter from James M. Taylor, Director, NRC Office of Inspection and Enforcement, to William G. Counsil, Executive Vice President, Texas Utilities Electric Company, forwarding a Notice of Violation and Proposed Imposition of Civil Penalties, Enforcement Action (EA) 86-63, Docket Nos. 50-445 and 50-446, May 2, 1986.*
C-9 U.S. Nuclear Regulatory Commission, " Safety Evaluation Report related to the operation of Comanche Peak Steam Electric Station Units 1 and 2," Docket Nos. 50-445 and 50-446, USNRC Report NUREG-0797, published July 1981. ** Pertinent Supplements published as follows:
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Supplement No. 7 published January 1985.**
l Supplement No. 6 published February 1985.**
l Supplement No.10 published April 1985.**
Supplement No. 11 published May 1985.**
C-10 Letter from W. G. Counsil, Executive Vice President, Texas Utilities Generating Company, to Vincent S. Noonan, Director, Comanche Peak Project, NRC Office of Nuclear Reactor Regulation, forwarding a report
(" Comanche Peak Steam Electric Station, Units 1 and 2,. Comanche Peak Response Team Program Plan and Issue-Specific Action Plans, Revision 3"), Docket Nos. 50-445 and 50-446, January 27, 19E6.*
C-11 Letter f rom E. H. Johnson, Director, Division of Reactor Safety and Projects, NRC Region IV, to W. G. Counsil, Executive Vice President, Texas Utilities Generating Company, forwarding Inspection Report No. 50-445/85-19, Docket No. 50-445, March 28, 1986.*
C-12 Letter from E. H. Johnson, Director, Division of Reactor Safety and Projects, NRC Region IV, to W. G. Counsil, Executive Vice President, Texas Utilities Generating Company, forwarding Inspection Report Nos. 50-445/86-04 and 50-446/86-03, Docket Nos. 50-445 and 50-446, March 27, 1986.*
C-13 Letter from James M. Taylor, Director, NRC Office of Inspection and Enforcement, to W. G. Counsil, Executive Vice President, Texas Utilities Electric Company, Docket Nos. 50-445 and 50-446, June 25, 1986.*
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C-14 Letter from W. G. Counsil, Executive Vice Presioent, Texas dtilities Generating Company, to James M. Taylor, Director, NRC Office of Inspection and Enforcement, Docket Nos. 50-445 and 50-446, August 4, 1986.*
C-15 Letter from Vincent S. Noonan, Director, Comanche Peak Project, NRC Of fice of Nuclear Reactor Regulation, to W. G. Counsil, Executive Vice l
President, Texas Utilities Generating Company, forwarding a report
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(" Report of the Review and Evaluation of Allegations of Intimidation and Harassment of Employees at Cotanche Peak Steam Electric Station U,its 1 and 2," October 1985), Docket Nos. 50-445 and 50-446, November 4, 1985.*
I C-16 Letter f rom W. G. Counsil, Executive Vice President, Texas Utilities j
I Generating Company, to James M. Taylor, Director, NRC Office of Inspection and Enforcement, Docket Nos. 50-445 and 50-446, June 2, 1986.*
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UNITED STATES
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NUCLEAR REGULATORY COMMISSION
- .g j.j REGION V 0, '-
,o'f 1450 MARIA LANE, SUITE 210
%,..U WALNUT CREEK. CALIFORNIA 94596 OCT C i 1995 l
l MEMORANDUM FOR:
C. J. Heltemes, Jr., Director, Office for Analysis and Evaluation of Operational Data FROM:
J. B. Martin, Regional Administrator I
SUBJECT:
ABNORMAL OCCURRENCE REPORT TC CONGRESS FOR SECOND QUARTER.
This is in reply to your memorandum, subject as above, dated September 18, 1986.
The attached comments (Enclosure.1) provide update and/or clarifications to the subject report for Region V facilities.
With the comments provided, we concur in the issuance of the report, er ! 0- :
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i.artin Regional Administtstor Attachment a/s cc: P. Bobe, AEOD
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-StsTo7o(94 F/u hff
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l ENCLOSURE 1 l
COMMENTS / UPDATE INFORMATION FOR A0 REPORT TO CONGRESS - SECOND QUARTER 1986 l
1.
Page 4. A0 86 Add the following as indicated:
In middle of next to last paragraph, the sentence which ct.rently reads " Core uncovery could have begun within about another hour."
l should be revised to read, " Core uncovery could have begun within j
abou. another hour if no water addition was made."
... In middle of next to last paragraph, the sentence which currently reads, "In addition, the event might not have occurred had a problem, which occurred earlier the same day, been properly analyzed I
and corrected." should be revised to read, "In addition, the event might not have occurred had a level indication' problem, which I
occurred earlier the same day, been properly analyzed and corrected."
l 2.
Page 37, A0 86-1.
Add the following between last two paragraphs; i
1 On September 17, 1986, the NRC forwarded to the licensee a Notice of Violation and Proposed Imposition of Civil Penalties in the amount of $180,000 (Ref. B-9).
l 3.
Page 38, A0 86-2.
We recommend this update be revised to read as follows:
On August 18 and September 4, 1986, Sacramento Municipal Utility District (the licensee) met with the NRC Staff to discuss the scope and content of the action plan for performance improvement. The licensee submitted this action plan for performance improvement on July 3, 1986, and the Staff has initiated its review. The critical path to restart decision is the Staff's review of the action plan and issuance of the Staff's safety evaluation report.
Future reports will be made as appropriate.
4.
Page 50, References for Appendices. Add the following reference:
B-9 Letter from John B. Martin, Regional Administrator, NRC, Region V to D. J. Fogarty, Executive Vice President, Southern California Edison Company, forwarding a Notice of Violation and Proposed Issuance of Civil Penalties, Docket No. 50-206 September 17, 1986.
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