ML19294C612
| ML19294C612 | |
| Person / Time | |
|---|---|
| Issue date: | 06/30/1980 |
| From: | NRC OFFICE OF MANAGEMENT AND PROGRAM ANALYSIS (MPA) |
| To: | |
| References | |
| NUREG-0090, NUREG-0090-V03-N01, NUREG-90, NUREG-90-V3-N1, NUDOCS 8009170056 | |
| Download: ML19294C612 (38) | |
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NUREG-0090 Vol. 3 No. 1 REPORT TO CONGRESS ON ABNORMAL OCCURRENCES JANUARY-MARCH 1980 Status as of May[23,]1CSO Date Published:
June 1980 Office of Management and Pr, gram Analysis United States Nuclear Regula: cry Commission Washing:cn, 9.C.
20555 Of6 8009170
e t
iii ABSTRACT Section 208 of tne Energy Reorganization Act of 1974 identifies an abnormal.
occurrence as an unscneculed incident or event which the Nuclear Regulatory Commission determines to be significant from the stancpoint of public health or safety and requires a quarterly recort of such events to be made to Congress.
This report, the twentieth in the series, covers the period from January 1 to March 31, 1980.
The following incidents or events, including any submitted by the Agreement States, were determined by the Commission to be significant and reportaole:
1.
ThereIwere three'labnormal occurrences at the nuclear power plants licensed to operate. @ne~involvedexposurestobetaradiationinexcessofregula-tory limits.
One] incident (a generic concern) involved a transient -
initiated by partial loss of pcwer. [Thetnirdinvolvedthecompleteloss of a safety system.]
2.
There was one abnormal occurrence at the fuel cycle facilities (other than nuclear power plants).
The incicent involved a loss of confinement system which resulted in internal deposition of plutonium in an employee.
2.
There was one abnormal occurrence at other licensee facilities.
The incident involved overexposure to individuals in unrestricted areas.
4.
There were no aonormal occurrences recorted by the Agreement States.
This report also contains information updating some previously reported abnormal occurrecces.
iv TABLE OF CONTENTS PAGE A B 5 T Tt t. C T..........................................................
iii PREFACE..................
v INTR 000CTION..................................................
v TH E R EGU LATO RY SY ST EM.........................................
vi REPORTABLE OCCURRENCES......................................
viii AGREEMENT STATES...........
ix REPORT TO CONGRESS ON ASNORMAL CCCURRENCES, J a n u a ry-Ma r c h 19 8 0..............................................
1 N UC L EA R P CW E R P LANTS...........................................
1 80-1 Occupational Overexposures to Skin and Extremities..
1 80-2 Transient Initiated by Partial Loss of ?cwer........
3 80-3 Failure of Salt Water Ccoling System......
8 FUEL CYCLE FACILITIES (Other than Nuclear Power Plants)......
10 80-4 Loss of Confinement System Resulting in Plutonium Deposition in an Employee..........
11 OTHER NRC LICENSEES (Industrial Radiographers, Medical Institutions, Incustrial Users, Etc.).............
13 S0-5 Overexposure to Individuals in Unrestricted Areas...
13 AGREEMENT STATE LICENSEES.....................................
15 APPENDIX A - ABNORMAL CCCURRENCE CRITERIA..........................
15 APPENCIX 5 - UPDATE OF PREVIOUSLY REPORTED ABNORMAL CCCURRENCES....
19 NUC LEAR POWER P LANTS......................
19 nunt:M.N, a Ai:
., Nas-2.....
ui :
- e c
c4 APPENDIX C - OTHER EVENTS OF INTEREST..........................
25
v PREFACE INTRODUCTION The Nuclear Regulatory Commission reports to the Congress each cuarter under provisions of Section 208 of the Energy Reorganization Act of 1974 on any abnon.Tal.cccurrences 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 frem the standccint of public health or safety.
Events are currently identified as a: normal occurrences for this re:crt by tne NRC using One criteria delineated in Appendix A.
These criteria were pr0mul-gatec in an NRC policy statement wnich was puolished in the Federal Recister (42 FR 10950) on Fe ruary 24, 1977.
!n order to provide wice cissemination of information to the public, a Federal Recister notice i.; issuec en each abnormal occurrence with ccoies distri:utec to ne NRC Public Occument Room and all local pu:lic document roces.
At a miniraum, each sucn notice contains the date anc piace of the occurrence and describes its nature and probable consequences.
The NRC has reviewed Licensee Event Reports, licensing and enforcement actions (e.g., violations, infractions, ceficiencies, civil penalties, license modifi-cations, etc.), generic issues, significant inventcry differences involving s:ecial nuclear material, and other categories of information available to the NRC.
The NRC nas cetermined that only those events, inclucing these submitted cy tne Agreement States, described in this report meet tne criteria for annormal Occurrence re:crting.
This report, the twentietn in the series, covers the period cetween January 1 - March 31,1980.
Information reocrted en each event includes:
date and place; nature anc proba:le consequences; cause er causes; and actions taken to prevent ecurrence.
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vi THE REGULATORY SYSTEM The system of licensing and regulation by wnicn NRC carries out its responsi-bilities is implemented through rules and regulatiens in Title 10 of the Code of Federal Regulations.
To accomolish its cojectives, NRC regularly c:nducts 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, witn the NRC establishing standards and rules; issuing licenses and permits; inspecting for ccepliance; enforcing license requirements; anc carrying on continuing evaluations, studies and research projects to improve botn the regulatory pr0 cess and tne protection of the puolic nealth and safety.
Public particioation is an element of the regulatory process.
In the licensing and regulation of nuclear power plants, tne NRC follows the philosophy that the health and safety of the puolic are best assured through the establishment of multiple levels of protection.
These multiple levels can be acnieved and maintained through regulatiens wnich soecify requirements wnien will assure tne safe use of nuclear materials.
The regulations incluce design and quality assurance criteria a:propriate for the various activities licensec by NRC.
An inscection and enforcement program helps assure comoli-ance with the regulations.
Recuirements for reporting incicents or events exist which help identify deficiencies early and aid in assuring that c0rrective action is taken to prevent their recurrence.
After the accident at Three Mile Island in Marcn 1979, the NRC and other groups (a Presicential Ccemission, Congressional anc NRC special incuiries, industry, special interests, etc.) spent substantial efforts to analyze tne accicent and its implications for the safety of ccerating reactors and to identify the changes needed to improve safety.
Scme deficiencies in design, operation and regulation were identified that required actions to upgrade safety as applied to nuclear pcwer plants.
Their included modifying clant har ware, imoroving emergency preparedness, and '- reasing considerably the empnasis on human factors such as expanding the nu:cer, training, and cuali-fications of the react 0c operating staff anc upgrading plant management and technical sucport staffs' capabilities.
In accition, each plant has installed dedicated telepnene lines to the NRC for raoid ccmmunication in the event of any incident.
Dedicated groucs have been formed botn by the NRC and oy the incustry for tne cetailed review of operating ex:erience to help identify safety concerns early, to imoreve dissemination of sucn information, and to feed ack tne experience into the licensing and regulatica process.
Most NRC licensee emolayees wne work witn racioactive materials are recuired t0 utilize personnel menitoring devices such as film badges or TLD (tnermo-luminescent dosimeter) bacges.
These adges are crocessed periodically and tne exposure results ncrmally serve as the official anc legal record of tne extent Of personnel exocsure to radiation curing tne pericd the badge was
vii THE REGULATORY SYSTEM (Continued) orn.
If an individual's past exposure history is known and has been suffi-ciently low, NRC regulations permit an individual in a restrictec area to receive up to three rems of whole bcdy exposure in a calendar quarter.
Higher values are permitted to the extremities or skin of tne wnole body.
For unre-stricted areas, permissible levels of radiation are considerably smaller.
c'ermissible doses for restricted areas and unrestrictec 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.
viii REPORTABLE OCCURRENCES Since the NRC is responsible for assuring that regulated nuclear activities are concucted safely, the nuclear industry is required to report incidents or events which involve a variance frem the regulations, such as personnel over-ex csures, radioactive material releases above prescribed limits, and malfunc-tiens of safety-related equipment.
Thus, a reportable occurrence is any incident or event occurring at a licensed facility or related to licensed activities which NRC licensees are recuired to repcrt to tne NRC.
The NRC evaluates each reportable occurrence tc determine the safety imolications involved.
Because of tne broad sc ce of regulation and the conservative attitude toward safety, there are a large number of events reported to One NRC.
The information crovided in these reports is used by the NRC and the industry in their continuing evaluation and improvement cf nuclear safety.
Some of the reports describe events that have real or potential safety implications; however, most of the recorts received frem licensed nuclear power facilities descri:e events that did not airectly involve the nuclear reacter itself, but invcived equipment and ccm:cnents which are peripheral aspects of the nuclear steam supply system, and are minor in nature witn respect to impact en puolic health and safety.
Many are discovered during routine inspection and surveillance testing and are correc:ec upon discove.ry.
Typically, they concern single malfunctions of c mpenents er parts of systems, with redundant Opera:le components or systems continuing to be availacle to perform the design function.
Information concerning reportable occurrences at facilities licensed or other-wise regulated by the NRC is routinely disseminated y NRC to the nuclear incustry, the public, and other interested groups as these events occur Dissemination includes deposit of incident reports in tne NRC's puclic document recms, special notifications to l'censees and Other affected or interested groucs, and public announcements.
In addition, a biweekly computer printcut containing information on reportable events received feca NRC licensees is sent to the NRC's more than 120 loca' puolic document recms throughcut the United States and to the NRC Puolic Document Rocm in Washington, D.C.
The Congress is routinely kept informed of reportable events occurring at licensed facilities.
iX AGREEMENT STATES Section 274 of the Atomic Energy Act, as amended, authorizes the Commission to enter into agreements with States whereby the Commission relinquishes and the States assume regulatory authority over byproduct, source and special nuclear materials (in quantities not capable of sustaining a chain reaction).
Com-parable and compatible programs are the basis for agreements.
Presently, information on reportable occurrences in Agreement State licensed activities is puolicly available at the State level.
Certain information is also provided to the NRC under exchange of information previsions in the agreements.
NRC prepares a semiannual summary of tais and other infornation in a document entitled, " Licensing Statistics and Otner Data," which is puolicly available.
In early 1977, the Commission determined that abnormal occurrences happening at facilities of Agreement State licensees should be included in the quarterly report to Congress.
The abnormal occurrence criteria incluced in Accendix A is applied uniformly te events at NRC and Agreement State licensee facilities.
Procecures have been developed and imolemented and any abnormal occurrences recorted :y tne Agreement States to the NRC are included in these quarterly reports :: Congrass.
REPORT TO CONGRESS ON ABNORMAL OCCURRENCES JANUARY-MARCH 1980 tUCLEAR PCWER PLANTS I e I4RC is reviewing events reported at the nuclear r cwer plants licensed to operate during the first quarter of 1980.
As of the date of this report, the NRC had determined that the following events were abnormal occurrences.
80-1 Occucational Overexoosures to Skin and Extremities During the preparation of this report, the following item was determined reportable using the criteria given in Appendix A of this report.Y Feceral Recister noticing is being made in conjunction with the noticing of 1ssuance of tmis report.
Date aad Place - On August 29, 1979, the Metropolitan Edison Company (Met-Ed) reportea to tne NRC the overexposures to six workers on August 28, 1979, at the Unit 2 Three Mile Island'(TMI) Nuclear Station located near Middletewn, Penr.sylvania.
Nature and Probable Consecuences - On August 28, 1979, six individuals, one a contractor Healtn Physics foreman, entered the north makeup valve room in the TMI Unit 2 Fuel Handling Building to inspect and tighten leaking valves in preparation for roca decontamination.
Reactor coolant system water leaking from these valves was highly contaminated as a result of the March 28, 1979 accident.
The stay time limit for this work was computed based on a radiological survey using an Eberline Instrument Company "Teletector" portable survey instrument.
The survey identified gamma radiation dose rates generally of 10 to 15 rem per hour in the room with one small area at 25 rem per hour.
Occupancy for each individual during the work was limited to a four-minute stay time and to areas not exceeding 15 rem per hour of gamma radiation.
However, the survey instrument used is not designed to measure beta radiation dose rates on range selections in excess of 2 rem per hour.
It was later determined that beta radiation exposure rates in the area were as high as 2500 rads / hour, or about 150 times the gamma exposure rates.
At about 1600 on August 29, 1979, it was determined from thermoluminescent
{
dosimeters worn by the six individuals that they had received beta raciation l
doses much greater thcn their gamma radiation doses.
The beta doses were in j
excess of the limits specified in 10 CFR 20.10 for exposures to the skin of the whole body or to an extremity.
These exposures were reported to the NRC on August 30, 1979.
Example 1 ("For All Licensees") notes. that exposure of the skin of the whole body of any #
individual to 150 rems or more of radiation can be considered an abnormal occurrence.
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2 A Met-Ed medical consultant examined five of tne six incividuals on t7e after-noon of August 30, 1979, and reported that there was no indication of a medically significant clinical effect from exposure to the skin for any of these individuals.
Since the radiation exposure of all six indivicuals exceeded NRC's regulatory limits for a calendar quarter the six individuals were restricted from radiation work for the remainder of the calendar quarter.
A follow-up report, submitted to the NRC on September 28, 1979, indicated large extremity (right leg) doses of non penetrating (beta) radiation, two of which were about 150 rem.
A further follow-up report, dated Decemcer 5,1979, gives the results of a reconstruction of the incident of August 2S, 1979, and gives estimates of exposures receivec by tne six workers.
Overexposures were reported for each worker to the skin of the whole body and overexposures for two of the workers to their nands.
NRC's exposure limits for the skin of the whole body are 7-1/2 rems per calendar quarter and 13-3/4 rems per calendar quarter for the hands and forearms, fee
- and ankles.
Summarized below are the exposures to these workers for the third calencar quarter of 1979:
EXPOSURE
SUMMARY
Skin Dose
- Ratio Hand Dose Ratio Individual (Rems)
(Skin Dose / Limit)
(Rems)
(Hand Dose / Limit) a 166 22.1 32 4.4 b
161 21.5 38 2.0 c
40
- 5. 3 8
0.A d
29 3.9 6
0.3 e
26
- 3. 6 16 0.9 f
13
- 1. 7 13 0.7
^$1nce one legs constitute a major portion of the bocy, leg skin exposures are considered whole body skin exposures.
Cause or Causes - These overexposures resulted from inadequate review and planning in preparation for the work and an inadequate radiation survey.
The radiation survey instrument selected for the work was not capable of measuring the actual radiation dose rate; the Health Physics representative was not adequately trained, or alerted, to survey for beta radiation in this work; and the radiation work permit thus did not require protective clothing acequate to prevent these overexposures.
Actions Taken to Prevent Recurrences Licensee - In those areas contributing to the occurrence, Met-Ed has reviewed ano strengthened tha Health Physics program.
Health ?hysics personnel nave
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3 been retrained in the use and limitation of their radiation survey instruments and in the proper planning and preparation for jobs.
Improved survey instruments have been obtained.
Personnel dosimetry practices have been upgraded and appropriate protective clothing requirements have been specified for areas where there are significant ceta radiation dose rates.
NRC - The services of a consultant have been obtained to furnish an independent case assessment of the six individuais.
NRC has reviewed Met-Ed's Health Physics program and directed that improvements De made on each of the items identified as a cause of the occurrence.
Onsite NRC inspectors are reviewing and observing licensee activities on every operating shift daily to ensure that necessary corrective action is taken.
NRC also has the generic aspects of the event under review to assure that a safety concern does not exist cue to possible inadequacies of present practices and regulatory requirements for occupational radiation monitoring in post-accident plant environments.
There are limitations on the ability of standard survey instruments to mea;ure.
radiation fields of different quality which must be understood and users must have availacle the necessary resoonse curves for instrument calibration, particularly for beta radiation.]
80-2 Transient Initiated by Partial Loss of Power During precaration of this report, the following item was determined recortable using the criteria given in Appendix A of this report.
Examole 12 ("For All Licensees") notes that a series of events (where individual events are not of major imoortance), recurring incidents, and incidents with imolications for similar facilities (generic incicents), which create major safety concern, can be considered an abnormal occurrence.
Federal Recister noticing is being made in conjunction with the noticing of issuance of tnis report.
Date and Place - On February 25, 1980, the Crystal River Unit 3 nuclear gener-ating piant (CR-3) located in Citrus County, Florida, experienced an incident involving an electrical malfunction in an instrumentation and control system.
This resulted in:
a reactor and turbine trip; the opening of the pressurizer power operated relief valve (PORV), the pressurizer spray valve, and a code safety valve; decreased feedwater flow to the steam generators; actuation of the engineered safety features (ESF) systems; and a cischarge of aporeximately a3,000 gallons of primary coolant into the containment building.
The pressur-i:ed water reactor was designed by the Sabcock & Wilcox Comoany (3&W).
Certain of the events were similar in some respects to those that took place during the accicent at Three Mile Island Unit 2 (TMI-2) on Marcn 2S, 1979 anc to those tnat have occurred at other B&W plants in recent years.
Nature and Probable Consecuences - A portion of the non-nuclear instrumentation (NNI) +24 voit power supply was lost due to a short to ground.
This loss affected automatic plant control systems and about 7C% of NNI control board indicators (such as reactor coolant system temperature, pressure, and flow; steam generator pressure and level; and pressuri:er level).
It caused the pressuri:er power coerated relief valve (PORV) ard the pressurizer spray valve to open.
The failure also caused false control signals to oe sent to tne
4 Integrated Control System (ICS), the most sicnificant of which caused a reduction infeedwaterflowtothesteamgenerators.[Also,thef,alseT signal caused ave the ICS to withoraw the control rods to increase power.J The reduction in feecwater flow recuced the reactor heat removal rate below the reactor heat generation rata which had increased because of rod withdrawal yielding a reactor coolant system temperature and pressure increase in spite of the open PORV and spray valve.
As a result, the reactor tripped on high pressure and then was subsequently partially depressurited.
The ocerators secured the reactor coolant pumps as required by the emergency procedures.
Hign Pressure Injection (HPI) automatically initiated as a result of Reactor Coolant' System (RCS) depressuri:ation due to loss of coolant inventory through the open FORV and the cooling effects associated with the reactor trip.
Shortly after receipt of a high reactor coolant drain tank level alarm, the PORV block valve was closed and, with approximately 7C% of NNI inoperable or inaccurate, the operator correctly decided that there was insufficient information available to justify terminating HPI.
Therefore, the RCS and pressuri:er were filled solid, causing RCS pressure to increase to the point wnere one safety valve lifted, and flow through the safety valve was spilling Reactor Coolant System water into the containment through tne Reactor Coolant Drain Tank rupture disk.
Power was restored to the NNI's about 20 mirutes after the start of the transient.
Plant conditions that existed then incluce:
pressuri:er filled solid with water; reactor coolant pressure 2400 psig; reactor coolant outiet temperature 556 F; steam generator "A" cry; the core being cooled by water flow from the high pressure injection system out the open safety valve and by natural circulation through steam generator "S".
Three minutes after the transient began, an operator, specifically assigned to purge containment to the atmosphere, terminated the purge.
Tnen, because water was still being discnarged from the RCS into the reactor building, the reactor building pressure increased to the point (4 psig) where automatic building isolation occurred.
After the restoration of power to the instrumentation, the ocerators throttled high pressure injection to reduce the flow of water through the ocen safety valve and into the reactor building.
The operators also re-established the water level in steam generator "A".
About 41 minutes after the transient began, the licensee declared a Class "B" Emergency based on the fact tnat coolant was being dischargec through the open safety valve and high pressure injection had been automatically initiated.
All non-essential site personnel were evacuated and off-site agencies notified.
About 1-L'2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after the beginning of the transient, tne operators estab-lished RCS pressure control using normal makeup and letecwn flows.
Because RCS temoerature and pressure were well under control and the core was being adecuately coolec by natural circulation, hign pressure :colant injection was shut off The decision was then made to heat up the pressurizer in preparation
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5 for establishing a steam bubble in the pressuri:er.
About 3-3/4 hours after the incident began, the pressurizer was heated sufficiently and a steam bubble was established in the pressurizer.
Forced flow in the RCS was established about 6-3/4 hours after the beginning of the incident by starting two reactor coolant pumps.
During the incident, about 43,000 gallons of RCS water was spilled into the containment.
The radiation level within cor tainment reached 50 rems per hour early in the event and decayed rapidly.
This was the result of short lived gasecus activity (i.e., xenon-138, krypton-89) wnien were released frca the reactor coolant.
Radioactivity released to the environment was within regula-tory limits, and the spilled RCS water was reprocessed for in plant use.
No core damage occurred.
Although there was no impact on the general public or olant emoloyees as a result of this incident, it is significant that failures in the NNI system, wnich is considered non-safety related, can have such an impact on reactor system cperation, control, and safe shutdown.
The design of this system received limited regulatory review and it is not subjected to rigorous separation, power reliability, or quality assurance criteria.
A particular area of concern in the B&W design deals with an apparent lack of sufficient design interface requirements between tne nuclear steam suoply system (NSSS) and the balance-of plant (BOF); that is, interfaces between the safety requirements for the auxiliary feedwater system and the cperating requirements for the ence-tnrough steam generators (OTSC) and the ICS.
Cause or Causes - The immediate cause of the C ystal River Unit 3 event has oeen tentatively identified as a +24 volt power short to ground between knife edge connectors of a Sailey Control Company Voltage Buffer card.
The voltage buffer card was misaligned in its receptacle, and adjacent connectors caraying
+2?V and " common" were bent such that they contacted one r ther.
This shcr circuit cleared itself during subsequent re-energi:ing of we power sucoly by burning through the foil on a printed circuit card.
(Subsequent review by Clorida Power Corporation identified a second voltage buffer card whicn was also misaligned but had not caused a short circuit.)
In addition t2 the transient caused oy this failure, a number cf similar transients have occurred in S&W plants involving power supply failures to non nuclear instrumentation (NNI) or the integrated control system (ICS) that caused reactor trips, PORV actuations, and feecwater transients.
Since Dece=cer 1974, a total of 29 NNI/ICS power failures have been identified on S&W plants.
Twenty-one of these events caused reactor trips, 17 caused PORV actuation, and a resulted in engineered safeguards (high pressure infection) actuations.
A steam dt.mo valve stuck open in one event, and feecwater transients occurred in 19 of tnese events.
A pressuri:er safety vaive lifted in one event, and a PORV stuck or failed open in three events.
Three ICS power failures tna occurred while the reactor was at power did not cause reactor trips.
In five ocwer failures, the reactor was in a shutdown condition when the events cccurred.
6 Based on these data, NNI/!CS pcwer failure perturbations have been severe enough, considering the S&W-integrated plant response characteristics, to cause reactor trip in most of the events (note:
in most of these instances, the reactor trip was the result of a consequential feedwater transient).
Acproximately 18 percent of all cbserved B&W feedwater transients have becn caused my NNI/ICS power failures.
This indicates that many other initiators of feecwater transients and reactor trips exist.
The data a: pear to show that, given an NN!/ICS power failure, it is very likely to result in a severe feecwater transient that will trip the reactor on high pressure.
Actions Taken to Prevent Recurrences Licensee - The CR-3 plant was shut down for resolution of the problem and imclementation of corrective actions.
These actions include:
Complete testing and inspection cf the NNI system for similar failures.
Installed new redundant channels of indication for twenty-three key plant parameters to provide more reliable information to the coerator.
Conduct comoreher ;ive operator training in resconse actions for NNI and ICS failures.
Install positive positien indication en the pcwer operated relief valva and the two cod 2 safety valves.
Modify ne NNI power supply to provide more reliable power.
Evaluate NNI pcwer supply reliability in response to IE Sulletin 79-27 (Loss of Non-Class 1E Instrumentation and Control Power System Sus During Cparation).
Modify the control circuitry for the FORV and pressurizer spray valves so that the valves will not open in the event of loss of NNI power.
NRC - Immediata NRC response was taken when the transient occurred.
Regional anc Headquarters emergency response centers were activated and regional teams dispatched to the site to assess plant conditions and evaluate the significanca of the event.
B&W licensecs were contacted regarding the Crystal River 3 NNI pcuer loss and possible operational consequences.
A meeting was held with B&W licensees at NRC Heacquarters on Maren 5. 1980, to review the event at Crystal River and discuss prccesed corrective actions.
IE Information Notice No. 50-10 (Partial Loss of Non-Nuclear Instrument System Power Supply Ouring Operation) was issued to all reactor licensees on March 7,1930.
In addition, faced with the CR-3 incident and the accarent high frequency of near similar types of transients in other S&W plants, a special task force (i.e., B&W Reactor Transient Response Task Ferce) was establisnea by tne
7 Directcr, Office of Nuclear Reactor Regulation (NRR), on March 12, 1980, to assess the generic aspects of coerating experiences of the B&W plants.
This assessment would include consideration of the apcarent sensitivity of the B&W plants to transients involving overcooling and undercooling conditions, small break loss-of coolant accidents (LCCAs), and the consequences of malfunctions and failures of the ICS and NNI.
The study would be made in conjunction with all the actions already taken or proposed in response to the TMI-2 accident.
The sensitivity of B&W reactor designs to transients and accidents has been discussed previously with the Commission.
Of particular concern is tne plant recovery from certain anticipated transients that can lead to frequent chal-lenges to the engineered safety features.
Some preliminary findings and conclusions were presented in the April 25, 1979 NRR Status Report on Feecwater Transients in S&W Plants which served, in part, as the casis for the confirma-tory shutdown ordars issued in May 1979 to all B&W cperating plant lionsees.
A more cc=olete exposition on this sucject is found in recent staff reports NUREG-0560, NUREG-0565, and NUREG-0645.
Scme of the cesign changes already acceeplished have helped to reduce the sensitivity of tre S&W design to certain transients (e.g., the addition of anticipatory reactor trips for icss of feedwater and turbine trip).
However, the operating ex,:erience obtained recently frem CR-3 requires furtner consideration of the NRC position on the B&W plants.
As stated in a memorandum dated October 25, 1979, frem the Director, NRR, to the Ccemissioners, the staff is continuing its review of the sensitivity issue and has initiated with the Office of Nuclear Regulatory Research a detailed stucy of risk assessment, on a relative basis, of the S&W design.
The present requirements being imposed en B&W plants are these developed mainly by the Lessons Learned Task Force and the Sulletins and Orders Task Force within NRR.
Further actions are being considered in an overall, inte-grated NRC Action Plan now under develcpment.
The plan will incorcorate the reccmmendations of the President's Cc= mission on the Accident at Three Mile Island as well as those of the NRC's Special Inquiry Group, the Acvisory Committee on Reactor Safeguards (ACRS), and other investigatory groups within the NRC and industry.
It is intended that the recommendations of the B&W Reacter Transient Response Task Force will be included in tne Task Action Plan for imclementation felicwing approval of the actions.
Other efforts dealing with the question of sensitivity of the B&W plants to transient response and failures of the instrumentation and control systems en a generic basis include the review of the CR-3 event and the responses to IE Bulletin 79-27.
IE Bulletin 79-27 ceals with an incicent tnat occurred at Oc: nee Unit 3 en Novem er 10, 1979, wherein a loss of NNI resulted in a cartial less of indication in the control recm.
In both of these matters, tne staff is reviewing resconses from the S&W licensees as weil as the joint report of tne Institute of Nuclear Power Operations and the Nuclear Safety Analysis Center (INP0/NSAC), cated March 11, 1980, regarding the CR-3 incicent.
Staff re crts will be issued following completion of these reviews.
8 Sy letter dated March 6,1980, the S&W ccerating plant licensees were asked a number of questions regarding the effect of the CR-3 event and actions being taken against consequential failure acdes in the NNI for their plants.
In adcition, Florida Power Corporation (FPC, licensee for CR-3) was recuested to specifically discuss the impact of the TMI-2 Lessons Learned and Bulletins and Orders modifications on its f acility with respect to the February 25, 1980 event.
FPC prcvided its resconse to this request by letter dated March 12, 1980.
Responses frem other S&W licensees were presented curing the week of Marcn 17, 1980.
On April la,1980, the Director, Office of Nuclear Reactor Regulation !ssued a Confirmatory Orcer cased on FPC commitments stated in their March 12, 1980 letter.
These licensee generated actions should reduce the prcbability of a similar power loss causing unexpected plant responses and allow the operator to better cepe with losses of instrumentation and control functions.
Similar orders were issued to all the Operating S&W nuclear plant licensees.
Although a detailed review of the CR-3 response has net teen ccmpleted, it can be concluded that no known effects resulting frcm the event would lead the staff to suspend implementation of the Lessons Learned er Eulletins and Orcers requirements.
However, further censideration of variations in the CR-3 event sequence is warranted to perhaps better assess the overall aspects of the lessens learned actions.
Even following the implementation of all the required and intended actions on tne B&W plants, there will ce ne guarantee that transients and accidents, similar to those that occurred at the TMI-2 and CR-3 facilities, can ce ccm-pletely prevented.
Mcwever, it is believed that the occurrence of such events would be less fraquent and of less consequence.
Thus, the effect of the actions wculd 5e to provide an increased margin of assurance to the health and safety of ths public.
Further recorts will be made as appropriate.
[30-3 Failure of Salt Water Cooline System During preparation of this report the folicwing item was determined reportable using ne criteria given in Appendix A of this report.
Example 12 (For All Licensees) notes that a series of events (where individual events are not of major boortance), recurring incidents, and incidents with implications for similar facilities (generic incidents), which create major safety cencern, can ce considered an abnormal occurrence.
Federal Register noticing is ceing made in conjunction with the noticing of issuance of nis recort.
Cate and lace - On March 10, 1980, San Oncfre Unit 1, a 456 MWe pressuri:ed water reactor that is operated by Southern California Edison Comoany in San Diego County, experienced the failure of the Salt Water Cooling System (SWCS).
Nature and Precacle Consecuences - The Salt Water Cooling System is a safety-relatec system na. scoplies cooling water to safety ecuipment such as neat
9 exchangers and necessary auxiiiary safety-related equipment.
The SWCS is recuired for normal operation and safe shutdcwn of the plant under certain transient and accident conditions.
If the system is inoperable, the reactor is required to be shut down, since without adequate cooling certain safety-related equipment wculd eventually fail.
This event involved the loss of two recuncant safety grade pumps; loss of a third pump which, thcugh not safety grade, was a backue pump in the Salt Water Cooling System; and the failure of the plant staff to shut dcwn the plant as required by the plant technical specifications and the plant operating procedures.
During this event, there was no asscciatac accident or radioactivity releases.
The temcerature in the component cooling water system, which is cooled by the Salt Water Cooling System, increased 16*F in 10 minutes, but was within normal operating parameters.
The sequence of events was as follows: While tne plant was operating at 10C".
power, the snaft on the south salt water cooling pump failed and the pump stocced pumping.
The redundant ncrth pumo aut:matically started, but did not succly salt water cooling to the cceponent cooling system since its isolation valve failed to autcmatically open.
Subsequently, the plant operators attempted to start the auxiliary salt water coeling pump, but there was insufficient priming and the pump was stepped.
Af ter aceut fif teen minutes, the operators -
sucolied a source of water for component cooling via aligning the screen wasta pumas to the salt water cooling system, a manual operation.
About twenty minutes later, the auxiliary salt water pumo was adecuately primed and placed in service, and about two and one-half hours later one of the main salt water c0 cling pumos was made operable when its isolati0n valve was opened.
Althcugh a plan shutdcwn was initiated accut 45 minutes after the initial loss of the sait water cooling system, it was terminated after a slight power reduction of about three megawatts after offsite plant management was consulted.
Other f acters likely associated with this event were deficiencies previously
,cted in the licensee's preventive maintenance program and the existence cf scoe desiccant in the plant's service air lines.
The desiccant problem had contributed to at least one previcusly reported valve failure to operata.
Altnough the plant operators followed the acpropriate procedures to ensure some cooling was sucplied for component cooling purposes and that the event was of a short duration (about 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> before a safety grace pump was made cperable), the NRC considers the ccmcination of the equipment failures, the failure to shut down the plant as required, and the deficiencies in tne preventive maintenance to be a major safety concern.
Cause er Causes - The equipment failures were randem failures - one SWCS pume snaft fa lec cue to a: parent excessive vibration; the other SWCS pum isolation valve f ailed to ocen cue to a deteriorated C-ring in a scienoid valve On tne valve acerator; and the auxiliary SWCS pump f ailed to prime due to an air leak into the priming system (this pump also had experienced priming prebiems curing periods of low tide).
The olant staff did nct continue the recuired snutcewn due to apparent misunderstandings of tne Technical Specification requirements and the technical basis for the requirements.
Other contri uting factors may have been desiccant in the plant service air system contribu-ing
10 to the SWCS pump isolation valve failure and preventive maintenance and surveil-lance program deficiencies centributing to the other andem failures.
Actions Taken to Prevent Recurrence Licensee - The licensee has taken, or will take, action as folicws:
(1) The equipment that failed was, Or will be, repaired and returned to service.
System redesign and changes to the preventive maintenance program will be implemented to improve system reliaoility.
(2) Desiccant is being flushed from the plant service air system during a refueling outage.
(3) Management has taken action to improve the plant staff's knowledge of the Techical Specifications and their basis.
Tighter controls were also placed on the administrative process for changing procedures.
Clari fica-tions are also being made to the Technical Specifications.
(4) The preventive maintenance program is being reviewed for imcrovements.
(5) An engineering review is being concucted of the reliability of the SWC system and its backups.
A review is also being made of the capability of the plant to withstand costulated accidents if the SWC5 is unavailacle.
NRC - The NRC has concucted special irspecticns of the facility related to tais event and through i s routine inspection and enforcement process has t
inspected the adecuacy of management and administrative controls, including the preventive maintenance program.
Based on the inspection of this event, tne licensee nas been cited with infractions of NRC reculations for failure to
~
stu: the plant cown when coth salt water cooling pumes and tne auxiliary salt water cooling pumos were inoperable.
As a result of a Fecruary 1979 inspection, the licensee was cited in January 1980 for non-ccmpliance with requirements for precedures for pump testing and for in-service testing of pumps and valves and a number of deficiencies related to the preventive maintenance progr71..
The NRC is requesting the licensee to further assess the imolications of a loss of SWCS during postulated accidents.
It i,s also continuing to review the acequacy of the licensee's corrective actions.J FUEL CYCLE FACILITIES (Other than Nuclear Power Plants)
The NRC is reviewing events reported by these licensees curing the first cuarter of 1980.
As of the cate of this report, the NRC had determinec that the fellowing event was an abnormal occurrence.
11 80-4 Loss of Confinement System Resultinc in Plutenium Decosition in an imotoyee Preliminary information pertaining to this incident was reported in the Federal Recister (
).
Appendix A (Example 3 of "For Fuel Cycle Licensees") or tnis report notes that an event which seriously ccmcrcmised tne ability of a confinement system to perform its cesignated function can be considered an acnormal occurrence.
In addition, exposure to an ccerator during the event appears to have ceen in excess of the limits of Section 20.103, 10 CFR Part 20,
" Exposure of Individuals to Concentrations of Radioactive Materials in Air in Restricted Areas."
Date and Place - The Bacccck and Wilcox Ccmpany, Nuclear Materials and Manufac-turing Division, Apollo, Pennsylvania, reported that en Novemcer 15, 1979, an operato' working in the Parkc Tcwnsnip Plutonium Facility received an apparent internal ceposition of plutonium when the integrity of the glovecox, in wnich ne was working, was comprcmised.
Nature and Probable Ccnsecuences - On November 15, 1979, at about 8:00 a.m., a tecnnician was reca1 ring a pcwcer blender in a plutnnium contaminated glovecox.
At this time he was pounding on a shaft removed trem the bim.f ar wita a hammer using the ficer of the glovecox as a base while attemoting to remove a frc:en bearing frca the shaft.
Attacned to the bottom of tais glovebox was a "well" that was used to hold containers to collect discharges frca the blending unit.
At accut 3:30 a.m. a second technician, wnile monitoring himself for contami-nation prior to leaving the area, discovered elevated levels of alpna radiation accarently frem plutonium contamination.
The technician doing the repair work immediately monitored his shoes and work clothing and found elevated contamina-tion levels on bis work clothing in *,he area of his upper bccy.
Other emolcyees
,.cr<ing in the area were requested to monitor themselves for contamination, and contaminated shoes were found on several individuals.
The area was then evacuated.
Fifteen individuals were working in the area; 12 of the 15 shcwed positive results on nose smears; only the individual working en the blender had exces-sively hign nose smear centamination results.
All 12 indivicuals were placed on a bioassay program and were whole bcdy counted.
This incident involved occucational hazards due to the loss of confinement in the glovebox.
There was no release of contamination to the off-site environ-ment, and the building confin;..ent systams were cperable.
Bioassay and wncle-Occy counting results indicated that 11 of the 12 individuals were not sucjected
- internal deccsition of radioactive material.
The 12th incividual, according to the licensee's evaluation wnich has been reviewed by NRC inscectors, was originally believed to have been subjected to an internal deccsition in tne lung of between 40 :: 50 nanocuries (nCi) of plutonium, as determined by in-vivo lung counting at tne University of Pittsburgn.
This incivicual was the person working at the glovebox at the time of the incident.
12 Recent in-vivo lung counting was performed on this individual at Los Alames Scientific Laboratory (LASL) using a new, sophisticatec pnantom (human bocy icck up).
The evaluation by LASL indicates a lung burden of 10 to 15 nCi of plutonium plus about 5 nCi of americium-241.
The new phantem is being ruuted to national laboratories and certain other facilities for calibration purposes, but it is not expected to be availaole to the University of Pittsburgn for 3 to a months.
It will therefore be several months before it is Known whetner the apparent discrepancy in the counting results is cue to calibraticn pre-cedures.
Isotepic analysis of an air sample collectec at the time of the incident appears to support tne LASL lung count.
Calculations have :een performed b= sed on the International Ccemission en Radiological Protection (ICRP) Tasn Croup Lung Mcdel and ICRP parameters for transfer of the inhaled material to tne bone.
For the upper value of 50 nCi, total dose to the lung wculd be about 100 rems, 95 percent of wnich would occur within aoout 6 years after the intake.
The total bone dose in 50 years would be about 320 rems or 6 tc 7 rems per year.
Using ICRP risk weighting factors for partial bcdy excesures, the effective whole body dose equivalent in the first year wculd be about 6 rems (compared to an average annual limit for wnole bocy cese of 5 rems).
Review cf available records indicates that a lung burcen of 50 nCi of plutonium represents one of the three largest Pu burdens to an :ccupational worker at licensed facilities.
Further, the lack of audible alarms incicates tnat the potential was present for ex osure t0 additional perscnnel.
Cause or Causes - The cause of the airborne release was due to equipment / material f aliure.
ine seal located between the gloveccx and the giovecox "well" located under the ficer of tne glovebcx had apparently failed and/cr bolts hoicing the "well" in place loosened.
Actions Taken to Prevent Recurrence Licensee - Operations in the area were suspended pending decontamination of cae rocm.
The licensee has repaired the original seal, and has installed a secondary seal between the glovecox floor and the well.
Consultant medical attention was furnished to the employee directly invcived.
The employee involved was removed frcm work in potentially contaminated areas indefinitely.
Audible alarm, continuous air monitors were installed in the area as accitional safety warning devices to supplement existing continuous air samolers, and to provide mere comolete coverage of the work area.
The licensee will continue to evaluate the empicyer's medical condition and deposition plutonium bccy burden; however, it may te some time before cetter cetermination of the ceposition is available and before the actual nalf-life in tne lung (ICRP estimates 500 days) can be estimated.
NRC - An investigation was conducted in which several items (not event related) of nonccmpliance with regulatory recuirements were identified.
These itas were corrected by the licensee.
Subsecuent action incluced a letter to tie licensee confirming that a recuest for a license amencment will te su:mit;ed
13 tc the NRC wnich specifies the criteria for use and location of the continuous audible alarm, continuous air scnitoring devices.
The NRC has also reviewed license c:.mmitments of other clutenium processing licensees to assure that the audible alarm, continuous air monitors are used in their operations.
This incident is closed for purposes of this report.
OTHER NRC LICENSEES (Industrial Radiographers, Medical Institutions, Incustrial Users, etc.)
There are currently more than 8,000 NRC nuclear material licenses in effect in the United States, principally fc use of radioisotooes in the medical, industrial, and academic fields.
Incidents were reported in this category frem licensees such as radiographers, medical institutions, and byproduct material users.
The NRC is reviewing events reported by these licensees during the first cuarter of 1980. As of :ne date of this report, tne NRC had cetermined that tne following events were acncrmal occurrences.
9 S0-5 Overexcesure to Individuals in Unrestricted Areas Duri preparation of this recert, the following item was determined reportable using ne criteria given in Jppencix A of this report.
Example 2 ("Fer All Licensees") notes that an exposure to an individual in an unrestricted area such that the whole bccy dose received exceeds 0.5 rem in one calendar year (10 CFR Part 20.105(a)) can be considered an abnormal cccurrence.
Feceral Recister noticing is being mace in con unction with :ne noticing of 1ssua.'ce oT "31s report.
Date and Place - A.nerican X-Ray and Inspection, Incorporated (AXI), of Farmington niiis, Micnigan, a field radiographer, exacsed 10 individuals in unrestricted areas to whole bocy doses in excess of 0.5 rem in a calendar year during the years 1977 and 1978.
Nature and Probable Consecuences - The licensee performed industrial radiog-racny services at various rieia sites for custcmers.
Incustrial radiogracny uses a radiation source to make X-ray-like pictures of heavy metal cojects.
In accition to field work at construction sites, AXI performed radiograony in a garage that was part of its facility in Farmington Hills, between August 10, 1977, and June 12, 1979.
The principal work in the garage during tnis period was raciograohing sample welds, called " coupons," wnicn are prepared by welders as part of tneir qualification tests.
The licensee's facility is located in an office-industrial cceplex.
The
- ortien of the garage area where the radiograpny work was performed shares a cc mon wall with two other businesses.
la The radiography was performed with iridium-192 sources ranging in strength from a to 100 curies.
During radiography exposures, radiation from the iridium-192 source was transmitted through the garage wall into the adjoining cusinesses which were not controlled by the licensee.
When the radiation source was not in use it was housed in a lead shielded container.
The licensee did not perform radiation surveys or conduct any surveillance in the unrestricted areas adjoining tne AXI facility.
The owners anc emoloyees of adjacent businesses were not nctified when the radiograohy was being performed.
Independent radiation surveys were performed by inspectors from the Nuclear Regulatory Commission's Region III (Chicago) office during an investigation January 8 - March 11,1980.
These surveys were conductec to determine the radiation levels created in the unrestrictec areas adjoining the licensee's garage during radiograpnic exposures.
Based on these survey results and on the radiographic crocecures, the NRC inspectors cetermined that, if an individual were continuously present in the unrestricted area adjacent to the licensee's garage, he could have received a maximum radiation exposure as high as 1200 millirems in any one hour.
NRC regulations limit the radiation levels in unrestricted areas such that exposures should not exceed 2 millirems in any one hour, nor 100 millirems in any seven consecutive days.
Based on the measured radiation exposure levels, when correlated with source sizes, work records, and time and attendance recorcs of employees at the adjoining businesses, it was estimated that the maximum exposure received by any employee of the adjacent businesses was 3 rems curing the 23-month period. Ten indivicuals in unrestricted areas were estimated to have received radiation exposures in excess of 0.5 rem in a calendar ysar, a level used in approving licensed activities.
The 'RC investigation determined that 36 other employees raceived less than 0.5 rems per year of unnecessary radiation exposure.
These levels of radiation excosure would not be expected to produce any medically discernible effects.
For comparative purposes, employees of NRC licensees who are exposed to raciation as part of their work are permitted to receive up to 5 rems per year.
Caust or Causes - The principal causes of the incident were the licensee's faliure:
(1) to make radiation surveys in the unrestricted areas; (2) to provice a shielded facility adequate to reduce tne radiation to acceptable levels in :ne unrestricted areas; and (3) to maintain surveillance of the unrestrictec areas so as to prevent entry or otherwise control exacsures to cersonnel.
Actions Taken to Prevent Recurrence Licensee - Radiogra:hy operations in the garage were stopped in July 1979.
NRC - An investigation was conducted from January 3, 1980, to Marca 11, 1980, as a result of allegations of unsafe raciation pro: action practices mace on
15 Novemoer 26, 1979, by a former AXI employee.
The investigation substantiated the allegations and identified several items of noncompliance witn NRC regula-tions.
An Order suspending the company's NRC license was issued on February 29, 1980; the order also recuired AXI to snow cause why its license should not be revoked.
On May 19, 1980, the comoany's NRC license was revoked.
A Cease and Desist Order was issued at the same time to Corsira X-Ray, Inc., a successor company which had takan over the AXI activities at the same location.
Corsira X-Ray did not possess an NRC license cnd was not authorized to possess and use radiation sources for performing industrial raciograpny.
The order also required Corsira X-Ray to transfer the radiation sources to an authorized recipient.
On May 15, 1980, Corsira X-Ray was issued an NRC license to do field raciography after naving met NRC requirements.
The NRC's Region III (Chicago) office met with the emolayees of the adjacent businesses to discuss the findings of the NRC investigation on May 22, 1980.
An NRC mecical consultant reviewed the exposure information and assisted the NRC and any concerned employee in evaluating the exposures.
This incident is closed for purposes of this report.
AGREEMENT STATE LICENSEES Procedures have been developed for tne Agreement States to screen unscheduled incidents or events using the same criteria as tne NRC (see Appendix A) and report tne events to the NRC for inclusion in this report.
Curing the first quarter Of 1980, the Agreement States reported no abnormal occurrences to the NRC.
16 APPENDIX A ASNORMAL OCCURRENCE CRITERIA The following criteria for this report's abnormai occurrence determinations were set forth in an NRC policy statement published in the Federal Recister (42 FR 10950) on February 24, 1977.
Events involving a major recuction in the degree of protection of the public health or safety.
Such an event would involve a moderate or more severe impact on the puolic health or safety and could include but neec not be limited to:
1.
Moderate exposure to, or release of, radioactive material licensed by or otherwise regulated by the Commission; 2.
Major degradation of essential safety-related equipment; or 3.
Major deficiencies in design, construction, use of, or manage-ment controls for licensed facilities or material.
Examples of the types of events that are evaluated in detail using these criteria are:
For All Licensees 1.
Exposure of the whole body of any individual to 25 rems or more of radiation; exposure of tne skin of the whole body of any individual to 150 rems or more of radiation; or excesure of the feet, ankles, hancs or forearms of any individual to 375 rems or more of radiation (10 CFR Part 20.403(a)(1)), or equivalent exposures from internal sources.
2.
An exposure to an individual in an unrestricted area such that the whole body dose received exceeds 0.5 rem in one calendar year (10 CFR Part 20.105(a)).
3.
The release of radioactive material to an unrestricted area in concentrations which, if aversged 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 Appencix E, Table II, 10 CFR Part 20 (lc CFR Part 20.403(b)).
4 Radiation or contamination levels in excess of design values on packages, or loss of confinement of radioactive material sucn as (a) a radiation dosa rate of 1,000 mrem per hour three feet from the surface of a package containing the racicactive material, or (b) release of radioactive material from a package in amounts greater tnan tne regulatory limit (10 CFR Part 71.35(a)).
17 5.
Any loss of licensed material in such quantities and under such circumstances that substantial hazard may result to persons in unrestricted areas.
6.
A substantiated case of actual or attemoted theft or diversion of' licensed material or sabotage of a facility.
7.
Any substantiated loss of special nuclear material or any substantiated inventory discrecancy which is judged to be significant relative to normally expected performance and which is judged to be caused oy theft or diversion or by cebstantial breakdown of the accountability system.
S.
Any substantial breakdown of physical security or material control (i.e., access control, containment, or accountability systems) that significantly weakened the protection against cheft, diversion or sabotage.
9.
An accidental criticality (10 CFR Part 70.52(a)).
10.
A major deficiency in design, construction or coeration having safety implications requiring immediate remedial action.
11.
Serious deficiency in management or procedural controls in major areas.
12.
Series of events (where indivicual events are not of major importance),
recurring incidents, and incidents with imolications for similar facilities (generic incidnits), whicn create major safety concern.
- or Ccmmercial Nuclear Power Plants 1.
Exceeding a safety limit of license Technical Specifications (10 CFR Part 50.36(c)).
2.
Major degradation of fuel integrity, primary coolant pressure boundary, or primary containment boundary.
3.
Loss of plant capability to perform essential safety functions such that a potential release of radioactivity in excess of 10 CFR Part 100 guidelines could result from a postulated transient or accident (e.g., loss of emergency core cooling system, loss of control rod system).
4 Discovery of a major condition not specifically considered in the Safety Analysis Report (SAR) or Tecnnical Specifications that recuires immediate remedial action.
18 5.
Personnel error or procedural deficiencies which result in loss of plant capability to perform essential safety functions such that a potential release of radioactivity in excess of 10 C?R Part 100 guidelines could result from a postulated transient or accident (e.g., loss of emergency core cooling system, loss of control rod cystems).
For Fuel Cycle Licensees 1.
A safety limit of license Technical Specifications is exceeded and a plant shutcown is required (10 CFR Part 50.36(c)).
2.
A major condition not specifically considered in the Safety Analysis Report or Technical Specifications that requires immediate remedial action.
3.
An event which seriously comcromised the ability of a confinement system to perform its desigaated function.
19 APPENDIX S UPDATE OF PREVICUSLY REPORTED ABNORMAL CCCURRENCES During the January through March 1980 period, the NRC, NRC licensees, Agreement States, Agreement State licensees, and other involved parties, such as reactor vendors and architects and engineers, continued with the implementation of actions necessary to prevent recurrence of previously reported abnormal occur-rences.
The referenced Congressional abnormal occurrence reports below provice the initial and any upcating information on the abnormal occurrences discussec.
Those occurrences not now considered closed will be discussed in subsequent recorts in the series.
NUCLEAR POWER PLANTS The following abnormal occurrence was originally reported in NUREG-0090-5,
" Report to Congress on Abnormal Occurrences:
July-September 1976," and uodated in subsequent reports in the series, i.e., NUREG-0090-8, Vol.1, No. 4, Vol. 2, No. 3, and Vol. 2, No. 4 It is further updated as follows:
76-11 Steam Generator Tube Intecrity Since the last general todate of this item (NUREG-0090, Vol. 2, No. a), tne following significant developments related to pressurized water reactor (PWR) steam generator tube integrity have occurred.
Westincnouse Desicned Units Point Beach Unit 1 continued to excerience tube degracation due to a phenomenon tesignated as "ceep crevice cracking." The unit has completed the 60-day operating interval allowed under terms of a Confirmatory Order issued on Novemoer 30, 1979.
The inspection required by the order has indicated that the rate of tuce degradation is somewhat retarded.
The licensee has ordered some long lead time items such as tubesheets and channel heads for potential replacement of the steam generators at this unit.
Trojan is scheduled to remove scme defective tubes for laboratory examinations curing the current refueling outage.
As reported earlier, Trojan previously experienced a tube leak due to a defect tangent to the inner tube row U-bend.
The Westingneuse topical report on their in-situ retubing concept is still under review cy the staff.
Comoustion Encineerine (CEl Desicned Units The staff is continuing their review of a proposed steam generator reclacement program for Palisades.
20 Babecck & Wilcox (B&W) Desicned Units No significant occurrences since the last update.
NRC Actions The generic tasks, reported on previously in NUREG-0090, Vol. 2, No. 4, continue to progress with a scheduled completion of May 1980.
Further reports will be mace as appropriate.
n n a s s s The following abncrmal occurrence was originally reported in NUREG-0090, Vol. 2, Nc.1, " Report to Congress on Abnormal Occurrences:
January-March 1979," and updatad in subsequent reports in this series, i.e., Vol. 2, No. 2 and Vol. 2, No. 4.
It is further utdated as follows:
79-2 Deficiencies in Picine Desien As previcusly reported, the Nuclear Regulatory Commission ordered five plants to snut ccwn on Marcn 13, 1979, until reanalysis and necessary mccifications were made to safety-related piping systems to bring them into conformance with requirements for withstanding earthquakes.
The plants ordered shut down were Beaver Valley Unit 1, James A. Fit:?atrick, Maine Yankee, anc Surry Units 1 and 2.
The recuired reanalysis and necessary modifications were completed for Maine Yankee, and an order was issued on May 24, 1979, terminating the March 13, 1979 Show Cause Order.
For Surry 1, an order was issued on August 22, 1979, to lift the suspension of cperation of the unit required by the Show Cause Order based en partial ccm-pletion of reanalyses and necessary modifications required by the Order.
Currently, all the reanalysis and medi'ications for the design basis earthquake (CBE) leading condition have been ccmpieted.
The required cperating basis earthcuake (08E) reanalysis and possible modifications will be completed oy Septemter 1980.
At that time, the Show Cause Order will be automatically terminated.
In the interim, VEPCO, the licensee, has committed to snut cown and inspect safety-related piping systems in the event of an earthquake with an acceleration greater than 0.01g.
For Surry 2, tne reanalysis and necessary modifica icns required by the Show Cause Order, for both the CBE and DBE loading conditions, nave ei ner ceen completed or are to be completec prior to startue.
An order was issued en Marcn 25, 1980 to terminate the Show Cause Orcer.
Fcr Fit: Patrick, an orcer was issued on August 14, 1979, to lift the suscension of coeration of the unit.
This war iased on partial completion of tne reanalysis
21 and necessary modification of the systems required by ',he Show Cause Order, for both the CBE and DSE loading conditions.
As a part of the 60-day require-ment of the August 14, 1979 order, 37 supports outside containment, all deter-mined to be operable with factors of safety equal to or greater than two, were icentified to require modification.
Of these 37 supports, 32 support modifica-tions have been implemented.
The remaining 5 succorts will be modified prior to a return to power following the current refueling outage.
At that time, the requirements of the August 14, 1979 Order will be completely satisfied.
- inally, for Seaver Valley, an order was issued on August 8,1979, to terminate the Show Cause Order based on partial completion of tne required reanalysis and necessary modification for the DEE loading condition.
After the Novemoer 30, 1979 snutdown for refueling and modifications to the reactor, 7 computer proolems were found to exist for which the SHOCK II run, instead of hand calculations as originally reported in the Show Cause reanalysis effort, is the run of record for some of the pipe supports incluced in these problems.
Among tne 57 supporcs identified in the 7 problems, 5 sucports were found to require modification.
For the DSE loading condition, the reanalysis is complete, and all the necessary modifications will be completed prior to the planned July 1980 restart.
For the CBE loading condition, the reanalysis and the necessary modifications will also be comoleted prior to the planned July 1980 restart.
At that time, the licensee will have met its commitment to snut down and inspect safety related piping systems in the event of an earthquake with an acceleration greater than 0.01g.
Further reports will be made as accrocriate.
a x x x x x The following abnormal occurrence was originally reported in NUREG-0090, Vol. 2, No. 1, " Report to Congress on Abnormal Cccurrences:
Janua n - March 1979," and ucdated in subsequent reports in the series, i.e.,
NUREG-0090, Vol. 2, No. 2, Vol. 2, No. 3, and Vol. 2, No. 4.
I. is further updated as follows:
79-3 Nuclear Accident at Three Mile Island (1)
Primary Coolant Water Leak On February 11, 1980, a small leak develoced from the TMI-2 makeup system.
The maxeup system maintains water inventory in the reactor primary coolant system anc controls reactor system pressure.
Up to 1,000 gallons of tne crimary coolant water leaked from the system to tne TMI-2 auxiliary builcing sumo and a small amount of krypton-35 (less than 0.3 curies) was released curing the incident.
The location and cause of the leak was the pressure switen on the "B" makeup pumo.
The system was isolated and tne standby pressure control system was initiated to maintain makeu: water and control reactor system pressure.
22 (2)
Radioactive Gas Leak On February 12, 1980, a small leak developed in the reactor building atmosphere air sample recirculator line.
The leak continued for approximately 17 hours1.967593e-4 days <br />0.00472 hours <br />2.810847e-5 weeks <br />6.4685e-6 months <br /> prior to octaining an air sample and isolating the line.
The total radioactivity released during the leaking period was calculated by the NRC on-site staff to be 4.1 curies of Kr-85.
The release was well belcw the regulatory limits of 10 CFR 20 and TMI-2 Environmental Technical Specificatiens.
(3)
Environmental Assessment for the Decontamination of the TMI-2 leaccor Builc1nc Acmosonere On March 12, 1980, the NRC/TMI Program Office staff presented to the Commission the draft report, " Environmental Assessment for the Decontamination of the Three Mile Island Unit 2 Reactor Building Atmospnere (NUREG-0E62)." The staff reccmmended that the reactor building atmosphere be decontaminated by controlled purging to the environment (subsequently addenda 1 and 2 to NUREG-0662 were issuec in March and April 1980, respectively).
The public comment period on the report has been extended to May 16, 1980.
(a)
Reactor Buildinc Personnel Airlock Entrv On March 13, 1980, the first entry into the TMI-2 reactor building persennel airlock No. 2 was made.
Prior to the entry, the airlock was curged of a small cuantity of Kr-85 gas (approximately 57 mci).
One purging began on March 10, 19S0 and lasted until March 13, 1950.
Radiation surveys taken in the airlock indicated no surface or airborne contamination.
The average area reading in tne airlock was 50 mr/hr with a maximum reading of 150 mr/hr (localized to a 1 ft square area).
Visual cbservations through the coservation port in the inner airlock decr did not reveal any visual camage.
The area in front of the dcor acceared clean and dry.
(5)
Radiolocical Environmental Monitorine Procram The on going radiological environmental monitoring around the TMI site and nearby communities is being performed by (a) Metropolitan Edison Ccmcany (the licensee), (b) the Ccmmonwealth of Pennsylvania, (c) the U.S. Environmental Protection Agency, and (d) the Nuclear Regulatory Ccmmission.
(a) Licensee's Procram Tne Three Mile Island Nuclear Station Offsite Emergency Radiolegical Environmental Mcnitoring Program consists of caily surface anc crinking water samoles taken at one ucstream raw drinking water station, one u:: stream surface water station, one plant intake station, one plant effluent station, five downstream crinking water stations (at which tnree raw and four finished samples are taken), and two downstream stations located on tributaries flowing into the Susquehanna.
Acditional samoies include weekly milk samcles at five locations, weekly air particulate and air iodine samples at eight locations, monthly crecioitation samples at
23 four locations and quarterly TLDs at 73 locations (of which 20 are also sampled monthly).
Fish, aquatic plants, and aquatic sediments will be samoled in July and Octocer if availaole, and miscellaneous food products will be sampled as available.
(b) Commenwealth of Pennsylvania Radiolacical Monitorinc Precram The Department of Environmental Resources cf the Ccmmonwealth of Pennsylvania operates three continuous air sampling stations; one at the Evangelical Press Building in Harrisburg, one at the TMI Observation Building, and one in Goldsboro near the boatdeck.
Each air samoling station consists of a particulate filter followed by a charcoal cartridge.
The filters and cartridges are changed weekly; the particulate air samples
're gamma scanned and beta counted for reactor-related racienuclides.
The particulate air samples are composited quarterly and analyted for Sr-S9 and Sr-90.
The charcoal samples are gamma scanned for reactor-related radianuclides.
(c)
U.S. Environmental Protection Acency (EPA) Radialecical Monitoring ProCram EPA cperates a network of eignteen continuous air monitoring stations at radial distances ranging frca 0.5 mile to 7 miles frem TMI.
Each station includes an air sampler, a gamma rate recorder, and three thermoluminescent desimeters (TL0s).
Air samoles are collected frca each station and analy:ed typically three times per week.
Gamaa rate recorder measures and recorcs external exposures.
Recorder charts are read on the same schedule usec for air sample collection and tne charts removed weekly for review.
TLDs are placed at each monitoring station and at 0.2 mile intervals along roads immediately parallel to the Susquehanna River near TMI cut to a distance of about 2.5 miles frem tne reactor.
These TLDs are read quarterly.
(d)
U.S. Nuclear Reculatory Ccemission Radiolocical Monitorina Procram The Nuclear Regulatory Commission (NRC) operates one air sampling station that is located in the middle of the reactor complex.
The air samoles are changed weekly and are analyzed by gamma spectrometry.
The NRC has placed two sets of TLDs at 47 locations.
Both sets are read on a monthly basis; however, flexibility exists to read one set at more frequent intervals should conditions warrant.
(6)
Radioactive Solid Radwaste Shicments On Janury 11, 1980, a meeting was held in Olympia, Washington, with Governor Dixie Lee Ray and members of her staff to discuss future shipments of TMI-2 solid racwaste to Richland, Washington, for burial.
The meeting was held at ne request of the licensee for the purpose of obtaining Governor Ray's appr oval for future shipments of scent dewatered resin to the low-level waste disposal site in Richland, Washington, for curial.
The NRC participated at the mecting.
Subsequently, the 1icensee obtained the approval and shipments resumed.
24 In response to increased corcern by Federal and State regulators of cackaged solid radwaste over the quantities of free water in packaged waste and the cevelopment of criteria limiting those quantities, tne licensee initiated a program to stucy various liner dewatering techniques, determine the parameters whicn influence liner free water, and quantify the amount of free water which could be expected in a dewaterec liner upon arrival at the burial site.
The NRC on-site staff has followed and evaluated the licensee's dewatering program since its inception.
The NRC concluded that tne licensee's dewatering procedure is acceptable and results in a quantity of free water which is well below current burial site acceptance criteria.
Further reports will be made as appropriate.
x x x x x =
AGREEMENT STATE LICENSEES The following abnormal occurrence was originally recorted in NUREG-0090, Vol. 2, No. 2, " Report to Congress on Acnormal Occurrences:
April-June 1979."
It is further updated as follows:
AS 79-1 Releases of Tritium and Contamination of Food American Atomics Corporation, the licensee, suspended coerations on June 15, 1979, and subsequently chose not to try to reopen the plant.
The licensee was required by the State to davelop a cecommissioning plan.
NRC proviced techni-cal assistance to the State in evaluating the decommissioning prooosals.
Of major concern was the presence of about 450,000 curies of tritium containec in rejected light sources and other scrap.
In late September 1979, Governor Saboitt ordered the National Guard to take control of tne American Atomics facility and had the rejected light sources and scrap removed to a leased U.S.
Geoartment of Defense facility wnere the material is new stored.
Decommis-sioning efforts at the plant continue under the supervision of the Ari:ena regulatory authorities.
This incident is closed for purposes of this report.
x x x x x =
The following abnormal occurrence was originally reported in NUREG-0090, Vol. 2, No. 2, " Report to Congress on Abnormal Cccurrences:
April-June 1979."
It is further upcated as follows:
AS 79-2 Overex osures from a Radiocrachy Source The California State Ecard of Inquiry has concluded its investigation of the incicent and is preparing a report.
The preliminary findings indicate that the incident was due to the faulty design of the connector assemoly.
In addition, the radiographer failed to follow prescribed procedures by failing to insert tne shipping plug into the radiograony camera after disconnecting tne
25 guide tube and failing to survey the camera.
As a result of the criminal invectigation of the incident, tne radiographer was found guilty of failure to take the necessary precautions to protect health and safety, failure to report tne loss of the source, and failure to report the overexposure of individuals to radiation.
Additional civil suits are pending.
The California State Board of Inquiry has concludec its investigation of the incicent and is preparing a report.
The preliminary findings indicate that the radiographer failed to follow prescribed procedures by failing to insert tne snipping plug into the radiography camera after disconnecting the guice tube and failing to survey tne camera.
In addition, faulty design of the connector assemcly is indicated as a contributing factor leading to this incident.
As a result of the criminal investigation of the incident, the radiogracher was found guilty of failure to take the necessary precautions to protect health and safety, failure to report the loss of the source, and failure to report the overexcesure of individuals to radiation.
Additional civil suits are pending.
This incident is closed for purpcses of this report.
26 APPENDIX C OTHER EVENTS OF INTEREST The following event is described below because it may possibly be perceived by the puolic to be of public health significance.
Tne event did not involve a rafor reduction in the level of protaction provided for public health or safety; therefore, it is not reportable as an abnormal occurrence.
Yankee-Rowe Turbine Tailure At about 9:30 a.m. on February 15, 1980, the Yankee-Rowe turbine experienced multiple turoine disc failures.
The plant had been shut down since January 19, 1980, to implement the Lessons Learned Category A items.
All work was completed and on February 14, 1980, operations to return tne unit to power were underway.
All turoine testing was comoleted showing all conditions to be normal.
The reactor was at 2 to 3 percent power to provide sufficient steam for turbine starter.
The turoine was rolled, brought up to 1500 rpm full speed and put on goverr~r control.
At 11:35 a.m., conditions were achieved to synchroni:e the generator and load the turbine.
Before synchronizing, a thumoN as neard from an acparent severe jarring of the turoine. The turoine coasted down for about 25 :ninutes compared to the normal coastdown time of 45 to 60 minutes.
It is believed that the turbine tripped automatically on tnrust bearing float.
But tne manual trip was also immediately activated, so it is not known which action actually caused the trip.
There are no vibration trip devices on this unit.
There was severe turbine vibration.
The vibration meters on No. 3 and No. 4 bearing read off scale (10 mils upcer scale limit).
The other bearings read between 6 and 7 mils.
Normal vibration is 1 mil or less.
Attemots to ou. the turoine on turning gear failed to turn the *.urbine.
Observation at tne insoection ports indicated no apparent damage to the last stage.
With the cover removed at the 3rd point extraction line, some debris was seen, anc a section of disc about 1 ft long was removed, containing one blade sheared at its dovetail, another blade twisted.
The reactor did not experience a transient as a result of the turbine failure.
There was negligible decay heat in the core because of the long period of snutoown and no condition to cause a reactor trip.
The reactor was left critical for about 1/2 to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> tnen was shut down normally and was placed in a cold snutdown condition.
There was no loss of condenser vacuun.
Conditions in the reactor cooling system and the secondary system appeared normal.
Yankee Atomic Electric Company (the licensee) contracted Westinghouse to perform an investigation of the turoine failure and to do the repairs.
The licensee anticipated a prolonged snutdown of the facility.
The turoine outer and inner casings were removed and an initial visual inspec-tion was performed.
It was found that both first stage discs in the low pressure rotor were completely failed.
They were broken in several majcr p'eces and many smaller fragments.
Major damage was also observed at several
27 adjacent rows of blades and stators.
Preliminary information indicates exten-sive cracking in the bore of the first stage disc at the generator end.
The mechanism of cracking is not known at this time, but is considered most likely to be primarily stress corrosion.
It appears that one large piece of a first stage disc got wedged against the shaft during turoine coastcown anc may have caused significant damage to the shaft. The inner housing was severely cracked and distorted, and cracks were found in the outer housing at the centering pin area.
The disc fragments were shipped to Westinghouse Steam Turcine.)ivision, Lester, PA, on Fecruary 22, for a detailed investigation.
One of the discs, that from the generator end, exhibited extensive bore cracking while the other, that from the governor end, exhibited no obvious bore cracking.
The reason for this disparity is not yet known as the discs are reportec to be essentially identical in metallurgical and mechanical properties.
The investigation is continuing, and Westinghouse has presented preliminary results to the utility.
3-1 OTHER EVENTS CONSIDERED FOR ABNORMAL CCCURRENCE REPORTING The following incidents are samples of incidents seriously considered for abnormal occurrence reporting.
The incidents are briefly discussed and the reasons why they are not being reported are stated.
Tne incidents were judged not to have involved a major reduction in the level of protection proviced for public health or safety. [0ne incident, Item 3 - Multiple Reactor Trips at Browns Ferry, is still uncer review.]
This enclosure is provided to the Commission per Ccmission ccmments on SECY-76-471, the enclosure will not be a part of the publisned report.
1.
Personnel Excesure to Radioactive Materiais at Ovster Creek On March 2,1980, two maintenance workers at the Oyster Creek Nuclear Generat-ing Station sustained extensive radicactive contamination of the skin and experienced an intake of radioactive material.
This occurred while the indivi-duals were wiping a control rod clace tcol during wnich they were not wearing any respiratory protection.
An inspection was conducted on March 13, 1980, to review the event and-the ame#4cy of licensee corrective actions.
The inspec-tion revealed several deficiencies in the implementation and adecuacy of the licensee's radiation protection program.
The licensee made immediate procedural cnanges to prevent recurrence of this type incident, and provided adcitional criteria for the use of respiratory ;rotection ecuipment.
Deficiencies in the licensee's raciation protection program were also revealed during a previous inspection conducted in October and November 1979.
Subse-quently, the licensee made changes in his radiation protection organization, ieclemented a formal training / retraining program, anc estaclished a raciological engineering grouc to perform tecnnical evaluations of tne radiciogical problems associated with work activities prior to the start of work.
The licensee is evaluating the potential intake of transuranics in excess of allowable limits by the two maintenance workers involved with tne event on March 2, 1980.
A preliminary NRC evaluation indicates no intake of radio-active materials in excess of allowable limits; the internal exposures of the two individuals were 8 percent and 23 percent of allowacle quarterly limits.
Based on these exposures and the significance of the licensee's deficiencies, it is considered that the event did not satisfy any of the criteria for acnormal occurrence report *ng.
2.
Receated Reactor Trios at Indian Point Unit 3 A totai of four reactor trips occurred at Incian Point Unit 3 curing the peried of Marcn 4 :nrough March 6, 1980.
The first anc fourth trips were causec ey Icw icw steam generator water level whicn were the result of main feec umo turbine control oil system malfunctions.
The second trio occurred unen a ain steam isolation valve inadvertently closed, causing icw low water level in a steam generator cecause of " shrinkage." The valve closure was
3-2 attributed to a contractor emoloyee " jostling" a valve trip solenoid.
The tnird trip occurred when the main turbine trioced.
The main turoine trip resulted frem actuation of the turbine thrust-bearing trip cevice prcoaoly due to clogged thrust-bearing oil no::les.
After the four trips, the plant remained shut dcwn by mutual agreement between the licensee and tne NRC until the problems associated with the trips were fully understood, evaluated, and satisfactorily resolved.
The licensee con-ducted an investigation of the causes of the trips and cetermined it was.
likely that that three of tne four events were the result of dirty oil systems, anc the fourth was due to inacequate protection of the valve trip solenoids.
The corrective action plan develeced by the licensee included (1) disassembly, cleaning, and checxout of the main feed pump turbine control oil system, (2) cleaning and inspection of tne main turbine oil system, and (3) construction of protective cages around the main steam isolation valve trip solencies.
All the reactor trips were handled with normal plant and personnel responses and no engineered safeguards feature actuations cccurred.
The plant safety systems associated with the trips functioned as cesigned.
Therefore, it is cons 4e+ red that the events did not satisfy any of tne criteria for abncrmal occurrence reporting.
t
[3.
Multiole Reactor Trios at Brewns Ferry l
Three unexplained reactor trips cccurred in February 1950 at Browns Ferry Unit 2.
NRC review of recordings, cceputer printouts and discussions with piant personnel including operators and instrument engineers did not provide evicence on what caused the trips, except that the reactor protective system (RPS) was actuated anc safety systems performec as intended.
The initial causative facter activating the RPS was not positively identified.
TVA investigations of the events corrected several minor potential prooiems and were able to duplicate spurious inputs to tne reactor protective system, but were not conclusively able to pinpoint the cause.
Specifically, a hycotnesis was developec by TVA cf a person bumping or striking certain instru-ment canels in the reactor building that could have caused the initiation of trip signals being transmitted to the RPS.
By use of a fast speed recorder this hypotheried mechanism was determined to ce valid.
F.xisting instrumentation, in the normal configuration, would not record tne sensor action due to the short duration of the initisting shock pulse.
Celiberate tripping of the reactor through this hypothesis was investigated but not proven.
Tne plant is not designed to prevent a person frcm celiberately ripoing the reactor if ne or she is bent on cerforming such an act.
As incicated previously, in eacn of these events all safety systems remained fully coeracle and performec as recuired.
TVA recuested assistance from the FBI and suscended eight empicyees during the investigations.
Seven of :nese emoioyees were reinstated after one day.
The eigntn, a laborer, was already slated to be laid of f in a recuction-in-force
3-3 and was not returned to work.
It is cur understanding from liaison with the Justice Department (FBI) that their investigation did not provide evidence sufficient to proceed further with indictments or charges. Que:tions of the FBI investigative authority in this area also arose.
This matter has been the subject of a recommendatien to provide Federal sanctions against individuals convicted of damaging facilities licensed by the NRC.
The plant is not designed to prevent an insider from deliberately tripping equipment such as instrumentation or electrical cabinets [Not all safety-related the reactor if he or she is bent on doing such an act.
have individual exterior protective features []ith a challenge to safety systers, the reactor is so Although scrams are not desirable since a transient is introduced along w designed'for shutdowns and places the plant in a safe configuration.
As indicated previously, in each of the scrams in questien, all safety systems remained fully operable and performed as required.
These events were not considered to be a breakdcwn of the physical security system in accordance with present regulatory requirements, and thus are not considered as potential abnormal occurrences at the present time {}
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FEB 21 ISEO gV p MEMCRANDUM FOR:
Norman Haller, Directer Office of Management and Program Analysis FRCM:
Carl Michelsen, Director Office for Analysis and Evaluatien of Operatienal Data
SUBJECT:
PROPOSED AENCRMAL OCCURRENCE - OVEREXPOSURE OF SIX WCRKERS AT TM1-2 We have reviewed the draf t Ccmissien pacer and prcocsed FRN attached to ycur February 13 memorancum on the subject occurrence.
On balance, we believe that the event shculd be included in the fourth cuarter 1979 Abnormal Occurrence (AO) Reper; to Congress as update material en the basic TMI-2 event, instead of as a secarate abnormal occurrence.
Although the calculated skin exposures to two i'idivicuals slightly exceeded the 150 rem technical thres-hold for immeciate reporting, the incident has limitec significance to public health and safety, and thus it cces no: seem to meet tne intent cr spirit of A0 reports tc Congress whien are to acdress highly significant events. Mcwever, the event reflects a breakocwn in training, procadures, and management conteci similar to other TMI-2 related events, such as the waste shi: ment difficulties, radiation menitoring program problems, accicental cnsite spills cf radicactive water, and inadvertent gasecus releases frca containment. We suggest that items of this nature be incluced as upcate material rather than separate reporting as AO's which dces not accear c be warranted.
In such reporting, we believe that the account for thi: even: shculd incluce tne significan: lessens to be learned, such as considering all scurces of raciatien prior to cetermining recuirec radiaticn cratecticn measures for scecial work conditiens and ensuring that the racia:icn meni cring ecui: ment is ca::able of measurinc actual radiaticn 2nvironments.
i'-- a tr ef statemaat shculd e N.:ldedlefplaining :ne limitad si;a.ificance :f :ne 5.e :
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[sa e:9 please inf:rt, us of ycur plans regarding tnis rec:=endaticn.
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