ML19340E499

Report to Congress on Abnormal OCCURENCES.April-June 1980

ML19340E499
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Issue date:	11/30/1980
From:	NRC OFFICE OF MANAGEMENT AND PROGRAM ANALYSIS (MPA)
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References
NUREG-0090, NUREG-0090-V03-N02, NUREG-90, NUREG-90-V3-N2, NUDOCS 8101140736
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Q NUREG-0090 Vol. 3, No. 2 Report to Congress on Abnormal Occurrences April - June 1980 Date Published: November 1980 Office of Management and Program Analysis U.S. Nuclear Regulatory Commission Washington, D.C. 20555

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iii ABSTRACT Section 208 of the Energy Reorganization Act of 1974 identifies an abnormal occurrence as an unscheduled incident or event which the Nuclear Regulatory Commission determines to be significant from the standpoint of public health or safety and requires a quarterly report of such events to be made to Congress.

This report, the twenty-first in the series, covers the period from April 1 to June 30, 1980.

The following incidents or events, including any submitted by the Agreement States, were determined by the Commission to be significant and reportable:

1.

There were two abnormal occurrences at the nuclear power plants licensed to operate.

One involved the loss of decay heat removal capability.

The other involved the failure of control rods to insert fully during a scram.

2.

There were no abnormal occurrences at the fuel cycle facilities (other than nuclear power plants).

3.

There were no abnormal occurrences at other licensee facilities.

4.

There were no abnormal occurrences reported by the Agreement States.

This report also contains information updating some previously reported abnormal occurrences.

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TABLE OF CONTENTS PAGE ABSTRACT...........................................................

iii PREFACE............................................................

v INTRODUCTION..................................................

v THE REGULATORY SY5 TEM.........................................

vi REPORTABLE OCCURRENCES........................................

viii AGREEMENT STATES..............................................

ix REPORT TO CONGRESS ON ABNORMAL OCCURRENCES, April-June 1980..................................................

1 NUCLEAR POWER PLANTS..........................................

1 80-5 Loss of Decay Heat Removal Capability...............,

1 80-6 Failure of Control Rods to Insert Fully During a Scram.............................................

6 FUEL CYCLE FACILITIES (Other than Nuclear Power Plants).......

10 OTHER NRC LICENSEES (Industrial Radiographers, Medical Institutions, Industrial Users, Etc.)...............

10 AGREEMENT STATE LICENSEES.....................................

10 APPENDIX A - ABNORMAL OCCURRENCE CRITERIA..........................

11 APPENDIX B - UPDATE OF PREVIOUSLY REPORTED ABNORMAL OCCURRENCES....

14 NUCLEAR POWER PLANTS..........................................

14 APPENDIX C - OTHER EVENTS OF INTEREST..............................

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PREFACE INTRODUCTION The Nuclear Regulatory Commission reports to the Congress each quarter under provisions of Section 208 of the Energy Reorganization Act of 1974 on any abnormal occurrences involving facilities and activities regulated by the NRC.

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 promul-gated in an NRC policy statement which was published in the Federal Register (42 FR 10950) on February 24, 1977.

In order to provide wide dissemination of information to the public, a Federal Register notice is issued on each abnormal occurrence with copies distributed to the NRC Public Document Room and all local public document rooms.

At a minimum, each such notice contains the date and place of the occurrence and describes its nature and probable consequences.

The NRC has reviewed Licensee Event Reports, licensing and enforcement actions (e.g., violations, infractions, deficiencies, civil penalties, license modifi-cations, etc.), generic issues, significant inventory differences involving special nuclear material, and other categories of information available to the NRC.

The NRC has determined that only those events, including those submitted by the Agreement States, described in this report meet the criteria for abnormal occurrence reporting.

This report, the twenty-first in the series, covers the period between April 1 - June 30,1980.

Information reported on each eveni, it.cludes:

date and place; nature and probable consequences; cause or causes; and actions taken to prevent recurrence.

vi 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 standards and rules; issuing licenses and permits; inspecting for compliance; enforcing license requirements; and carrying on continuing evaluations, studies and research projects to improve both the regulatory process and the protection of the public health and safety.

Public participation is an element of the regulatory process.

In the licensing and regulation of nuclear power plants, the NRC follows the philosophy that the health and safety of the public are De n assured through.

the establishment of multiple levels of protection.

These mhitiple levels can be achieved and maintained through regulations which specif* requirements which will assure the safe use of nuclear materials.

The regulations include design and quality assurance criteria appropriate for the various activities licensed by NRC.

An inspection and enforcement program helps assure compli-ance with the regulations.

Requirements for reporting incidents or events exist which help identify deficiencies early and aid in assuring that corrective action is taken tc, prevent their recurrence.

After the accident at Three Mile Island in March 1979, the NRC and other groups (a Presidential Commission, Congressional and NRC special inquiries, industry, special interests, etc.) expended substantial efforts to analyze the accident and its implications for the safety of operating reactors and to identify the changes needed to improve safety.

Some deficiencies in design, operation and regulation were identified that required actions to upgrade the safety of nuclear power plants.

These included modifying plant hardware, l

improving emergency preparedness, and increasing considerably the emphasis on human factors such as expanding the number, training, and qualifications of the reactor operating staff and upgrading plant management and technical support staffs' capabilities.

In addition, each plant has installed dedicated telephone lines to the NRC for rapid communication in the event of any incident.

Dedicated groups have been formed both by the NRC and by the industry for the detailed review of operating experience to help identify safety concerns early, to improve dissemination of such information, and to feed back the experience into the licensing and rc.gulation process.

Most NRC licensee employees who work with or in the vicinity of radioactive I

l materials are required to utilize personnel monitoring devices such as film l

badges or TLD (thermoluminescent dosimeter) badges.

These badges are processed periodically and the exposure results normally serve as the official and legal record of the extent of personnel exposure to radiation during the period the l

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THE REGULATORY SYSTEM (Continued) 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 values are permitted to the extremities or skin of the whole body.

For unrestricted areas, permissible levels of radiation are considerably smaller.

i Permissible doses for restricted areas and unrestricted areas are stated in 10 CFR Part 20.

In any case, the NRC's policy is to maintain radiationtexposures to levels as low as reasonably achievable.

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viii REPORTABLE OCCURRENCES Since the NRC is responsible for assuring that regulated nuclear activities are conducted safely, the nuclear industry is required to report incidents or events which involve a variance from the regulations, such as personnel over-exposures, radioactive material releases above prescribed limits, and malfunc-tions of safety-related equipment.

Thus, a reportable occurrence is any incident or event occurring at a licensed facility or related to licensed activities which NRC licensees are required to report to the NRC.

The NRC evaluates each reportable occurrence to determine the safety implications involved.

Because of the broad scope of regulation and the conservative attitude toward safety, there are a large number of events reported to the NRC. The information provided in these reports is used by the NRC and the industry in their continuing evaluation and improvement of nuclear safety.

Some of the reports describe events that have real or potential safety implications; however, most of the reports received from licensed nuclear power facilities describe events that did not directly involve the nuclear reactor itself, but involved equipment and components which are peripheral aspects of the nuclear steam supply system, and are minor in nature with respect to impact on public health and safety.

Many are discovered during routine inspection and surveillance testing and are corrected upon discovery.

Typically, they concern single malfunctions of components or parts of systems, with redundant operable components or systems continuing to be available to perform the design function.

Information concerning reportable occurrences at facilities licensed or other-wise regulated by the NRC is routinely disseminated by NRC to the nuclear industry, the public, and other interested groups as these events occur.

Dissemination includes deposit of incident reports in the NRC's public document rooms, special notifications to licensees and other affected or interested groups, and public announcements.

In addition, a biweekly computer printout containing information on reportable events received from NRC licensees is sent to the NRC's more than 120 local public document rooms throughout the United States and to the NRC Public Document Room in Washington, D.C.

The Congress is routinely kept informed of reportable events occurring at licensed facilities.

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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 publicly available at the State level.

Certain information is also provided to the NRC under exchange of information provisions in the agreements.

NRC prepares a semiannual summary of this and other information in a document entitled, " Licensing Statistics and Other Data," which is publicly available.

In early 1977, the Commission determined that abnormal occurrences happening at facilities of Agreement State licensees should be included in the quarterly report to Congress.

The abnormal occurrence criteria included in Appendix A is applied uniformly to events at NRC and Agreement State licensee facilities.

Procedures have been developed and implemented and any abnormal occurrences reported by the Agreement States to the NRC are included in these quarterly reports to Congress.

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REPORT TO CONGRESS ON ABNORMAL OCCURRENCES APRIL - JUNE 1980 NUCLEAR POWER PLANTS The NRC is reviewing events reported at the nuclear power plants licensed to operate during the second quarter of 1980.

As of the date of this report, the NRC had determined that the following events were abnormal occurrences.

80-5 Loss of Decay Heat Removal Capability During the preparation of this report, the following item was determined reportable using the criteria given in Appendix A of this report.

Example 11

("For All Licensees") notes that serious deficiency in management or procedural controls in major areas can be considered an abnormal occurrence.

Federal Register noticing is being made in conjunction with the noticing of issuance of this report.

Date and Place - On April 19, 1980, Toledo Edison Company reported the tempo-rary loss of decay heat removal capability at Davis-Besse Unit I while the plant was in a refueling outage.

The Davis-Besse Unit 1 nuclear plant utilizes i

a pressurized water reactor, designed by Babcock & Wilcox, and is located in Ottawa County, Ohio.

Nature and Probable Consequences - On April 8, 1980, the plant was taken to cold shutdown for refueling, maintenance, and modifications.

On April 19, 1980, the plant experienced a loss of two of four 120 VAC essential instrument buses resulting in a loss of decay heat removal (OHR) capability for about two and one-half hours.

At the time of the event, reactor coolant system (RCS) temperature was 90 F, decay heat was being removed by Decay Heat Loop No. 2, the vessel head was detensioned with bolts in place, the reactor coolant level was slightly below the vessel head flanges, and the manway covers on top of the once-through steam generators (OTSG) were removed.

Since the plant was in a refueling mode, many systems or components were out of service for maintenance or testing purposes.

In addition, other systems and components were deactivated to preclude their inadvertent activation while in a refueling mode.

Systems and components that were not in service or leactivated included:

Containment Spray System; High Pressure Injection System; Source Range Channel 2; Decay Heat Loop No.1; Station Battery IP and 1N; Emergency Diesel-Generator No. 1; 4.16 KV Essential Switchgear Bus C1; and 13.8 KV Switchgear Bus A (this bus was energized but not aligned).

The event was due to the tripping of a non-safeguards feeder breaker in 13.8 KV Switchgear Bus B (probably due to mechanical vibration or bumping of the w

2 breaker by construction workers who were working in the area).

Because of the extensive maintenance and testing activities being conducted at the time, Channels 1 and 3 of the Reactor Protection System (RPS) and Safety Features Actuation System (SFAS) were being energized from only one electrical source, the source emanating from the tripped breaker.

Since the SFAS logic used at Davis-Besse is a two-out-of-four input scheme in which the loss (or actuation) of any two input signals results in the actuation of all four output channels (i.e., Channels 1 and 3, and Channels 2 and 4), the loss of power to Channels 1 and 3 bistables also resulted in actuation of SFAS Channels 2 and 4.

The actuation of SFAS Channels 2 and 4, in turn, affected Decay Heat Loop No. 2, the operating loop.

Since the initiating event was a loss of power, all five levels of SFAS and the SFAS interlock channel on the Decay Heat Isolation Valve DH-12 actuaced on the loss of 120 VAC power to Channels 1 and 3 (i.e., Level 1 - High Radiation; Level 2 - High Pressure Injection; Level 3 - Low Pressure Injection; Level 4 -

Containment Spray; and Level 5 - ECCS Recirculation Mode).

Loss of power in the interlock circuit resulted in loss of decay heat pump suction from RCS hotleg No. 2 when containment isolation valve DH-12 closed.

Actuation of SFAS Level 3 aligned the Decay Heat Pump No. 2 suction to the Borated Water Storage Tank (BWST) in the low pressure injection mode.

Actuation of SFAS Level 5 represents a low level in the BWST; therefore, upon its actuation, ECCS opera-tion was automatically transferred from the Injection Mode to the Recirculation Mode.

Transfer to the Recirculation Mode closed the supply valve from the BWST and opened the valve to the dry Containment Emergency Sump.

During the opening and closing of these two valves (60 to 90 seconds), approximately 3,500 gallons of water from the BWST were injected into the RCS via the Decay Heat Pump No. 2 and approximately 1,500 gallons backflowed into the Emergency Sump by gravity.

The operator manually stopped Decay Heat Pump No. 2 approximately 2 minutes into the event to stop the injection of water into the RCS to pre-vent pump damage due to loss of suction.

The above sequence of events resulted in the loss of decay heat removal capability for approximately two and one-half hours, the time to properly check out and realign the electrical systems and realign and vent air from No. 2 Decay Heat Loop. Decay Heat Loop No. 1 was drained at the time in preparation for maintenance and not available as an alternate decay heat removal system.

During the time of the event, the reactor coolant temperature increased from 90*F to about 170 F (the Technical Specification definition for refueling mode is an average temperature of <140 F); however, the final temperature reached was still considerably below that which could adversely affect the heat transfer characteristics of the fuel such that fuel damage could result.

There were no offsite releases of radioactivity, and there were no overexposures or injuries to personnel associated with the event.

3 The loss of power a'.a caused communication problems in addition to the loss of decay heat remova: apability.

The Gaitronics System (internal communica-tions system) was without power for about 33 minutes.

This complicated com-1 munications between the control room personnel and personnel in other parts of the plant, which may have contributed to the delay in restoring decay heat removal capability.

There have been other incidents involving the decay heat removal systems at Davis-Besse Unit 1 during the refueling outage which began on April 8, 1980.

On April 18, 1980, while the reactor was still in cold shutdown prior to entering the refueling mode - an operator discovered that the water level in the RCS had dropped significantly.

The loss of reactor coolant inventory was from an open valve which had been opened to facilitate draining of the out-of-service Decay Heat Loop 1 and from a partially open valve (due to the valve's remote operator being out of adjustment) in the operable Decay Heat Loop 2.

A few minutes into the event, the running decay heat pump was tripped because of concern for possible loss of suction; the pump was restarted about 29 minutes later.

This was a violation of the Technical Specifications which require that while in Mode 5 (cold shutdown), at least one reactor coolant loop must be in operation with an associated reactor coolant pump or a decay heat pump operating.

On May 28, 1980, Decay Heat Isolation valve DH-11 was inadvertently tripped closed resulting in loss of decay heat removal for about two minutes until the valve was reopened and the pump restarted.

An Instrumentation and Control (I&C) mechanic was preparing to test an NRC required modification using the same pressure instrument used to activate DH-11 interlock circuit.

Due to a test procedural inadequacy, the valve interlock circuit actuated when a test connection caused the pressure input to spike high.

On May 31, 1980, the control room operator stopped the decay heat pump when the flow meter indication for the decay heat loop dropped off scale.

The pump was returned to service when it was discovered that an I&C mechanic had taken the flow meter out of service for surveillance testing without informing the control room.

On June 14, 1980, an inadvertent SFAS Levels 1, 2, and 3 actuation resulted in a loss of decay heat removal flow for about two minutes.

An I&C mechanic was in the process of restoring containment pressure inputs to SFAS following an Integrated Leak Rate Test.

Due to a procedural inadequacy, the SFAS actuated aligning the operating decay heat pump to the BWST and injecting water into the RCS and refuel.cg canal.

The BWST level dropped to the low level limit, actuating SFAS Leeel 5, closing the BWST isolation valves.

Tris caused a loss of suction to the tecay heat pump.

On July 10, 1980, at 1050 hours0.0122 days <br />0.292 hours <br />0.00174 weeks <br />3.99525e-4 months <br />, the decay heat remeval flow was inadvertently reduced to about 2000 gpm for approximately 51 seconds on restoring power to the flow control valve.

Due to a procedural error, power was lost to the flow control valve when SFAS Channel 2 was de-energized for maintenance work on the 120-VAC essential Bus Y-2.

When power was restored to the control valve, the

4 valve controller caused the decay heat flow to momentarily decrease below the minimum required flow of 2800 gpm.

On July 24, 1980, at 0935 hours0.0108 days <br />0.26 hours <br />0.00155 weeks <br />3.557675e-4 months <br />, a blown fuse caused the Decay Heat Isolation valve DH-12 to close resulting in loss of decay heat removal for about 50 minutes until manual bypass valves were open.

The blown fuse was caused by an electrician pulling wire, (in the cabinet containing the isolation valve control wires) for an NRC-required design change.

Also on July 24, 1980, at 2232 hours0.0258 days <br />0.62 hours <br />0.00369 weeks <br />8.49276e-4 months <br />, Decay Heat Isolation valve DH-11 was inadvertently tripped closed resulting in loss of decay heat removal for about two minutes until the valve was reopened and the pump restarted.

Inadequate job planning for restoration of the system from a modification led in I&C mechanic to perform steps out of sequence.

On August 8, 1980, Decay Heat Isolation valve OH-11 was inadvertently tripped closed result %g in loss of decay heat removal for about three minutes until the valve was reopened and the pump restarted.

A bistable in the valve circuit was removed during maintenance causing the valve to trip.

In the planned maintenance, the function of the bistable had been improperly designated.

On August 13, 1980, Decay Heat Isolation valve DH-11 was inadvertently tripped closed resulting in loss of decay heat removal for about five minutes until the valve was reopened and the pump restarted.

An I&C mechanic failed to fully defeat the automatic trip on DH-11 prior to performing modification work on an SFAS channel.

Cause or Causes - Two factors identified as major contributors to the events are (1) extensive maintenance activities which led to a loss of the DHR capability, and (2) inadequate procedures and/or administrative controls which, if corrected, could have precluded the events or at least ameliorated their effects.

For example, for the rather extended loss of DHR capability on April 19, 1980, the high pressure injection and containment spray pumps had been deae.tivated to preclude their inadvertent actuation during refueling.

If SFAS Level 5 had also been bypassed or deactivated, or if the emergency sump isolation valves had been closed and their breakers opened, this event'would have resulted in a minor interruption of decay heat flow.

It is also believed that the event could have been avoided, or ameliorated, if the maintenance activities had been less extensive during the outage, or at least better coordinated.

If i

activities had been restricted so that two SFAS channels would not be lost by l

a single event (e.g., serving Channels 1 and 3 frc:n separate sources), the loss of DHR capability would not have occurred.

In addition, if a backup DHR l

system would have been readily available, consequences of the loss of the operating DHR loop would have been lessened.

Actions'Taken to Prevent Recurrence Licensee - As a result of the April 19, 1980 event, to increase the reliability of decay heat removal during the refueling outage, the licensee (1) closed and electrically disabled the isolation valves to the containment emergency sump, I

5 (2) kept second decay heat loop in standby until the refueling cinal was filled, and (3) reviewed future electrical distribution system maintenance, modification, and testing to provide maximum diversity to the 120-VAC instru-m e t power buses.

Appropriate operating procedures were modified.

Corrective actions to the internal communication system problem will be included in the licensee's response to NRC's Inspection and Enforcement Circular No. 80-09 (Problems with Plant Internal Communications Systems) recommendations issued to all holders of a power reactor operating license or construction permit on April 28, 1980.

To date, the licensee has made several changes including the placement of the Red Phone on uninterruptible power, installing internal antennas in the auxiliary building for better radio communication and other telephone changes.

However, the power source for the Gaitronics System has yet to be modified.

Long-term corrective actions were taken by the licensee in accordance with the NRC's Inspection and Enforcement Bulletin No. 80-12:

1.

Additi,nal revisions to EP 1202.32, Loss of DHR Frequency Procedure, to include alternate methods to those previously listed to supply water to the reactor core and reference to appropriate procedures for monitoring core temperatures using the incore thermocouples.

2.

Additional guidance was provided for venting the DHR if air is drawn into the system.

3.

Five procedures were revised to insure power is removed for Emergency Sump Isolation valves OH-9A and 98 in Modes 5 and 6.

4.

Instrument AC System Procedure SP 1107.09 was revised to allow the 120-VAC instrument power inverter to be supplied from the DC Bus when normal AC feed is rot cvailable.

This will minimize the possible loss of power to two instrument channels at one time.

5.

A special procedure was written to require, whenever possible, that recundant decay heat system not be intentionally removed from service in Modes 4, 5, or 6 unless at least one steam generator is available for decay heat removal, the refueling canal is filled, or the decay heat pump can be restored to service or a gravity flow path to the RCS can be established within four hours.

The special order also covers expediting the restoration of redundant or diverse methods if component failure causes loss of alternate decay heat removal methods.

NRC - After the licensee notified the NRC's Inspection and Enforcement Incident Response Center on April 19, 1980, the Region III (Chicago) Incident Response Center was manned.

The resident inspector reached the site a short time later.

Two regional inspectors were dispatched to the site the next day to assist in the review of the event and corrective actions.

The investigation revealed only one item of noncompliance - failure to establish fire watches within the required time when the fire alarm system became inoperable due to the loss of power.

All other limiting conditions for operation were satisfied.

6 Pressurized water reactors (PWRs) are most susceptible to losing DHR capability when steam generators, or other means of removing decay heat, are not readily available.

Such conditions often occur when the plants are in a refueling or cold shutdown mode, and when concurrent maintenance activities are being per-formed.

The risk and frequency associated with this type' of event dictated investigation of other PWRs.

Actions to be taken by licensees were listed in NRC's Inspection and Enforcement Bulletin No. 80-12, " Decay Heat Removal System Operability," issued to all licensees of PWR facilities on May 9, 1980.

These actions include review of each facility for all DHR degradation events expe-rienced, especially those similar to the Davis-Besse Unit 1 event; review of the facility's hardware capability for prevention of DHR loss events; analysis of procedures for safeguarding against loss of redundancy and diversity of DHR capability; and analysis of procedures for adequacy of responding to DHR loss avents.

This incident is closed for purposes of this report.

80-6 Failure of Control Rods to Insert Fully During a Scram Preliminary information pertaining to this incident was reported in the Federal Register (45 FR 69318).

One of the general criteria listed in Appendix A of this report notes that major degradation of essential safety-related equipment can be considered an abnormal occurrence.

Date and Place - On June 28, 1980, Browns Ferry Unit 3, a boiling water reactor (BWR), located in Limestone County, Alabama, reported that 76 control rods failed to insert fully during a routine shutdown by a manual scram actuation at about 35% power.

Nature and Probable Consequences - Following the manual scram actuation, 76 of 185 control rods failed to insert fully. The partially inserted rods were all (with one exception) on the east side of the core where reactor power level was indicated to be 2% or less.

The west side of the core was subcritical.

A second manual scram was initiated 6 minutes later and all partially inserted rods were observed to drive inward, but 59 remained partially withdrawn.

A third manual scram was initiated 2 minutes later, and 47 rods remained partially withdrawn.

Six minutes later, an automatic scram occurred and all the rods inserted fully when the scram discharge level bypass switch was returned from

" bypass" to " normal" and there was a high water level in the scram discharge instrument volume.

It appears that this was a coincidence in that a manual scram would probably hava produced the same result.

Core coolant flow, tem-perature, and pressure remained normal for the existing plant conditions.

There was no danger to the general public or plant employees as a result of this event, No radioactivity was released to the environment.

There was no indication of fuel damage.

This type of occurrence could result in failure of the control rods to insert fully in part or all of the core on any automatic or manual scram signal.

Such a failure to scram on demand has the potential to cause significant fuel damage.

7 Cause'or Causes - The problem has been determined to bc hydraulic in nature rather than electrical or mechanical.

The control rod drives (CRDs), which insert and withdraw the attached control rods in a General Electric BWR, are essentially water-driven hydraulic pistons.

On a scram, a relatively high water pressure is applied to the bottom side of the piston by opening a scram inlet valve; a scram outlet valve opens to relieve water and pressure above the piston and the rods are rapidly driven up into the reactor core.

Water discharged from the 185 individual CRDs during scram insertion is collected in two separate headers consisting of a series of interconnected 6-inch-diameter pipes (four on each side of the reactor) called the scram discharge volume (SDV).

During normal operation, both SDVs are designed to remain empty by being continuously drained to a separate scram discharge instrument volume (SDIV) tank.

The SDVs are therefore normally ready to receive the scram discharge water when a scram occurs.

This instrumented tank is monitored for water level and initiates an automatic scram on high level, in anticipation of too much water in the SDV preventing a scram.

The control rod drives at Browns Ferry Unit 3 are grouped in such a manner that the east and west sides of the reactor core are connected to separate SDVs.

Later tests, inspections, and analyses resulted in the con-cit,sion that the east SDV was substantially full of water at the time of the event, leaving insufficient room for the discharge water.

Accordingly, upon scram acttation, the CRDs rapidly drove the control rods partially into the core but rod motion prematurely ceased when pressure quickly equalized on each side of the pistons.

Following each scram actuation, the scram signal was reset by the operator, allowing some water to drain from the SDV, permitting the rods to insert further with each scram attempt.

Sufficient water was finally drained from the SDV to allow the rods to insert fully on the fourth scram signal.

It is believed that the east SDV water accumulation problem resulted from improper drainage,into the SDIV from the SDV due to inadequate SDV venting, an obstruction in the line between the SDV and SDIV, or a combination of these problems.

Actions Taken to Prevent Recurrence Licensee - The unit remained shut down while a series of tests was performed in an attempt to determine the cause of the water accumulation in the SDV.

Ultrasonic probes were installed on the SDVs to continuously monitor the water level in the SDVs.

The unit was authorized to restart on July 13, 1980, as discussed below.

NRC - Immediately following the event, Region II dispatched a specialist to the site to assist the two NRC resident inspectors.

Region II also issued a letter confirming the licensee's (Tennessee Valley Authority) commitment to obtain NRC concurrence prior to restart.

An evaluation team consisting of the Region II Director, Region II specialists, and NRC Headquarters personnel was assembled at the site to evaluate the significance of this event.

A Preliminary Notification was issued to inform other NRC offices promptly.

On July 3, 1980, IE Bulletin No. 80-17 was issued to all licensees operating BWRs and required them to conduct prompt and periodic inspections of the SDV; perform two reactor scrams within 20 days while monitoring pertinent parameters to further confirm

8 operability; review emergency procedures to assure pertinent requirements are included; and conduct additnal training to acquaint operating personnel with this type of problem. On J r 18, 1980, Supplement 1 to Bulletin 80-17 was issued to all licensees operating BWRs.

This supplement required an analysis of the "as built" SDV; revised procedures on initiation of the standby liquid control system (SLCS); specifying in operating procedures action to be taken if sater is found in the SDV; daily monitoring of the SDV until a continuous monitor can be installed; and studying of designs to improve the venting of the SDV.

During testing required by IE Bulletin 80-17, the following anomalies were found:

1.

On July 19, following the manual scram at Dresden 3, the SDV was aligned for draining. An ultrasonic test (UT) of the SDV showed the west SDV to

.be 80% filled with water when it was thought to be empty. The SDV con-tains a ball check valve in the vent line that serves as a vacuum breaker.

The ball check provides a vent path directly from the reactor building atmosphere in the event the vent header does not function.

The vent header terminates in the reactor building equipment drain tank (RBEDT).

l This line extended into the tank and below the surface of the water.

Because the SDV was draining slower than the operators had anticipated, l

the ball check valve was unseated by the operators and the SDV drain rate increased.

2.

At Duane Arnold, the SDIV drain valve was found installed so that pressure in the SDIV tended to ur, seat the drain valve disk.

This resulted in leakage out of the SDIV during the scram.

This was corrected by reversing and reinstalling the valve.

The scram tests were performed on July 12 and 13, and the drain valve was corrected before return to power operations on July 17, 1980.

3.

At the Millstone Unit 1, the scram tests were performed successfully on July 11 through 14.

The function of the 10-second delay on scram reset was tested separately from the scram tests.

Review of the separate test results by plant personnel established that the scram reset delay feature was not functioning in the scram circuits due to a wiring error on the circuit boards.

This was corrected.

4.

At Browns Ferry Unit 1, a test scram involving two rods was performed on July 19, 1980.

The test showed normal response of level switches in the SDIV. When proceeding to drain the SDIV, however, the SDV did not empty as required and expected.

A vacuum in the SDV apparently existed which kept the system from draining.

Subsequently, the vacuum was cleared by operator actions and the volume drained properly.

Tests are continuing toward determination of the cause and to measure the vacuum.

5.

At Nine Mile Point Unit No. 1, one rod failed to scram during the manual scram test on July 14, 1980.

This was due to a failure of the scram pilot valve for that rod.

9 6.

On July 21, 1980, the licensee for Peach Bottom 2 and 3 reported that improperly rated solenoids for the backup scram valves had been installed.

The installed solenoids were rated at 250 volts DC but were connected to a 125-volt DC power source.

These valves are not considered essential to safety because the remainder of the scram system satisfies all NRC licensing requirements.

The solenoids were replaced with ones requiring 125 volts DC.

As a result of the above findings, Supplement 2 to IE Bulletin 80-17 was issued on July 22, 1980. This required the BWR licensees to previde a vent path from the SDV directly to the building atmosphere without any intervening component except for the vent valve itself.

These modifications had to be completed within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> for plants operating or prior to startup for plants shut down.

Browns Ferry Unit 3 was authorized to restart on July 13, 1980, following com-pletion of the actions required by IE Bulletin 80-17 and other extensive tests.

A letter was issued on July 14, 1980, confirming the actions taken by the licensee and to confirm that the licensee will perform an expedited review of design changes recommended by the reactor vendor (General Electric Company).

Continuing staff review of this event identified a potential for unacceptable interaction between the control rod drive system and the nonessential control air system; therefore, IE Bulletin 80-17 Supplement 3 was issued on August 22, 1980.

This Supplement required affected BWR licensees to implement operating procedures within five days which required an immediate manual scram on low control air pressure, or in the event of multiple rod drift-in alarms, or in the event of a marked change in the number of control rods with high tempera-ture alarms.

In addition, the licensees were requested to implement procedures which require a functional test using water for the instrument volume level alarm, rod block, and scram switches after each scram event.

On October 2,1980, the NRC issued Confirmatory Orders to the licensees of 16 BWR plants requiring the installation of equipment to continuously monitor water levels in all SDVs and provisions for water level indication and alarm for each SDV in the control room.

Until the system is installed and operating satisfactorily, the licensees shall increase their surveillance of the SDV water level.

The equipment provides information to the reactor operator such that if water accumulates in the SDV, reactor operators may take timely action.

This equipment is to be operable by December 1,1980 or prior to restart for those reactors in refueling, except installation for Browns Ferry Units 1 and 2 is required by December 22, 1980.

Browns Ferry already has continuous monitors located outside the control room.

The various aspects of the problem have been and continue to be actively studied by the NRC, the BWR licensees, and the reactor vendor.

The NRC Office for Analysis and Evaluation of Operational Data has prepared two detailed reports

(" Report on the Browns Ferry 3 Partial Failure to Scram Event on June 28, 1980,"

dated July 30, 1980, and " Report on the Interim Equipment and Procedures at Browns Ferry to Detect Water in the Scram Discharge Volume," dated September 1980 - both reports have been placed in the NRC Public Document Room).

The

10 recommendations of these reports are being actively pursued by the NRC staff.

A BWR Owners Sub-Group, which was organized for this issue, proposed criteria which must be satisfied for any further changes (i.e., design, procedures) which may be necessary in the future; these criteria were submitted to the NRC and were approved subject to comments.

The NRC is also preparing generic and plant-specific safety evaluation reports of the problem including short-and long-term corrective actions taken and required, respectively.

It is planned to issue a report by December 1, 1980 which will describe the acceptable criteria for any long-term procedure and system changes which may be necessary.

The BWR owners are to prov'de the NRC by December 15, 1980 with the specific proposed changes, and a schedule for their implementation.

The modifications proposed will likely be made within two years for the various affected plants.

Further reports will be made as appropriate.

FUEL CYCLE FACILITIES (Other than Nuclear Power Plants)

The NRC is reviewing events reported by these licensees during the second quarter of 1980.

As of the date of this report, the NRC had not determined that any events were abnormal occurrences.

OTHER NRC LICENSEES (Industrial Radiographers, Medical Institutions, Industrial Users, etc.)

There are currently more than 8,000 NRC nuclear material licenses in effect in the United States, principally for use of radioisotopes in the medical, indus-trial, and academic fields.

Incidents were reported in this category from licensees such as radiographers, medical institutions, and byproduct material users.

The NRC is reviewing events reported by these licensees during the second quarter of 1980.

As of the date of this report, the NRC had not determined that any events were abnormal occurrences.

AGREEMENT STATE LICENSEES Procedures have been developed for the Agreement States to screen unscheduled incidents or events using the same criteria as the NRC (see Appendix A) and report the events to the NRC for inclusion in this report.

During the second quarter of 1980, the Agreement States reported no abnormal occurrences to the NRC.

11 APPENDIX A ABNORMAL OCCURRENCE CRITERIA The following criteria for this report's abnormal occurrence determinations were set forth in an NRC policy statement published in the Federal Register (42 FR 10950) on February 24, 1977.

Events involving a major reduction in the degree of protection of the public health or safety.

Such an event would involve a moderate or more severe impact on the public health or safety and could include but need not be limited to:

1.

Moderate exposure to, or release of, radioactive material licensed by or otherwise regulated by the Commission; 2.

Major degradation of essential safety-related equipment; or 3.

Major deficiencies in design, construction, use of, or manage-ment controls fc-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 whale body of any individual to 25 rems or more of radiation; exposure of the skin of the whole body of any individual to 150 rems or more of radiation; or exposure of the feet, ankles, hands or forecrms 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 averaged over a period of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, exceed 500 times the regulatory limit of Appendix B, Table II, 10 CFR Part 20 (10 CFR Part 20.403(b)).

4.

Radiation or contamination levels in excess of design values on packages, or loss of confinement of radioactive material such as (a) a radiation dose rate of 1,000 mrem per hour three feet from the surface of a package containing the radioactive material, or (b) release of radioactive material from a package in amounts greater than the regulatory limit (10 CFR Part 71.36(a)).

l l

12 5.

Any loss of licensed material in such quantities and under such circumstances that substantial hazard may result to persons in unrestricted areas.

6.

A substantiated case of actual or attempted theft or diversion of licensed material or sabotage of a facility.

7.

Any substantiated loss of special nuclear material or any substantiated inventory discrepancy which is judged to be significant relative to normally expected performance and which is judged to be caused by theft or diversion or by substantial breakdown of the accountability system.

8.

Any substantial breakdo i; of physical security or material control (i.e., access control, containment, or accountability systems) that significantly weakened the protection against theft, diversion or sabotage.

9.

An accidental criticality (10 CFR Part 70.52(a)).

10.

A major deficiency in design, construction or operation having safety implications requiring immediate remedial action.

11.

Serious deficiency in management or procedural controls in major areas.

12.

Series of events (where individual events are not of major importance),

recurring incidents, and incidents with implications for similar facilities (gencric incidents), which create major safety concern.

For Commercial Nuclear Power Plants 1.

Exceeding a safety limit of license Technical Specifications (10 CFR Part 50.36(c)).

2.

Major degradation of fuel integrity, primary coolant pressure boundary, or primary containment boundary.

3.

Loss of plant capability to perform essential safety functions such that a potential release of radioactivity in excess of 10 CFR Part 100 guidelines could result from a postulated transient or accident (e.g., loss of emergency core cooling system, loss of control rod system).

4.

Discovery of a majcr condition not specifically considered in the Safety Analysis Report (SAR) or Technical Specifications that requires immediate remedial action.

~

13 5.

Personnel error or procedural deficiencies which result in loss of plant capability to perform essential safety functions such that a potential release of radioactivity in excess of 10 CFR Part 100 guidelines could result from a postulated transient or accident (e.g., loss of emergency core cooling system, loss of control rod systems).

For Fuel Cycle Licensees 1.

A safety limit of license Technical Specifications is exceeded and a plant shutdown is required (10 CFR Part 50.36(c)).

2.

A major condition not specifically considered in the Safety Analysis Report or Technical Specifications that requires immediate remedial action.

3.

An event which seriously compromised the ability of a confinement system to perform its designated function.

1

L->-

J

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L 14 APPENDIX B UPDATE OF PREVIOUSLY REPORTED ABNORMAL OCCURRENCES During the April through June 1980 period, the NRC, NRC licensees, Agreement States, Agreement State licensees, and other involved parties, such as reactor vandors and architects and engineers, continued with the implementation of actions necessary to prevent recurrence of previously reported abnormal occur-rences.

The referenced Congressional abnormal occurrence reports below provide the initial and any updating information on the abnormal occurrences discussed.

Those occurrences not now considered closed will be discussed in subsequent reports in the series.

NUCLEAR POWER PLANTS The following abnormal occurrence was originally reported in NUREG-75/090,

" Report to Congress on Abnormal Occurrences:

January - June 1975," and updated in subsequent reports in this series, i.e., NUREG-0090-1, 2, 3, 9, Vol. 1, No. 3, Vol. 2, No. 2, and Vol. 2, No. 4.

It is further updated as follows:

75-5 Cracks in Pipes at Boiling Water Reactors (BWRs)

The NRC staff updated the implementation document NUREG-0313 as a subtask under Generic Task A-42, " Pipe Cracks in Boiling Water Reactors." The objective of other subtasks is to identify and recommend additional measures to reduce the susceptiblity of stainless steel piping to stress corrosion i

cracking.

A report (NUREG-0313, Rev. 1) on the results of this task was published in October 1979.

Following the issuance of.this report, the NRC noticed the availability of this report in the Federal Register and requested interested parties to provide any comments to the NRC by January 16, 1980.

To date, comments from 11 organizations and individuals have been received.

The organizations and individuals that submitted comments are General Electric l

Company, seven utilities, BWR Owners Group, and two individuals--one from NBS and th,e other from the University of Tokyo.

The largest group of comments were those requesting approval of several recently developed fabrication and evaluation techniques such as Heat Sink Welding (HSW), Induction Heating Stress Improvement (IHSI), Electrochemical Potentiokinetic Reactivation (EPR), and Stress Rule Index that had not yet been accepted by NRC.

There were a number of comments objecting to certain requirements with accompanying technical justification and suggesting clari-fication of certain topics by changes in language or format.

All the comments were evaluated to determine their significance, and several modifications to the report were made to accommodate those significant comments.

The final NUREG-0313, Rev. 1 was published in July 1980.

Further reports will be made as appropriate.

l l

15 The following abnormal occurrence was originally reported in NUREG-0090-5,

" Report to Congress on Abnormal Occurrences:

July-September 1976," and updated in subsequent reports in the series, i.e., NUREG-0090-8, Vol. 1, No. 4, Vol. 2, No. 3, Vol. 2, No. 4, and Vol. 3, No. 1.

It is further epdated as follows:

76-11 Steam Generator Tube Integrity Westinghouse Designed Units San Onofre Unit 1 shut down on April 8,1980, because of an increasing primary to secondary coolant leak, just two days before a scheduled maintenance and refueling outage.

Subsequent hydrostatic testing of the steam generators revealed confirmed leakers in one steam generator and probable leakers in one other of the three steam generators.

Multi-frequency eddy current inspection was performed on 100% of the tubes and revealed large numbers of tubes with defect indications located at the top of the tubesheet.

One defect was also identified two inches below the top of the tubesheet.

Tube specimens were removed from the steam generators and subjected to laboratory examination and a new eddy current probe, sensitive to circumferential defects, was used for additional inspection.

As a result of the exhaustive inspection and examination, the licensee, Southern California Edison Company (SCE), deter-mined that caustic intergranular attack was occurring within a -inch band at the top of the tubesheet for the majority of tubes within the sludge piles. A hardened sludge pile with a maximum height of approx Nately 18 inches exists over approximately two thirds of the tubesheet on the hot leg side of each steam generator.

After defining regions of active integranular attacks, SCE decided to repair i

the steam generator tubes by installing leak tight sleeves inside approximately l

two thirds of the tubes; i.e., approximately 2500 tubes out of 3800 per steam i

generator.

The tubes that will not be sleeved are basically the tubes near l

the periphery of the steam generator.

Along the periphery, the sludge pile is very shallow or non-existent and SCE contends that their examinations show very l

low levels of the intergranular attack and therefore the peripheral tubes are well within acceptable operating limits.

The sleeves are rolled and brazed into place after honing the tube inside surface.

Each sleeve is 36, 30, or 27 inches in length, depending on installation restrictions imposed by the i

i geometry of the steam generator channel head bowl.

Tubes that have been identified as having a high level of intergranular attack, but cannot accept a l

sleeve because of their geometric location, will be mechanically (removable) plugged until another method of repair is devised.

l SCE and their vendor, Westinghouse, held several meetings with the NRC staff commencing in June 1980 and extending into September 1980 to describe the repair and plugging of the steam generator tubes as well as ALARA activities prior to and during the repair program.

On October 23 and 24, 1980, SCE responded to an NRC recommendation by sponsoring a third party independent review board review

16 of the Westinghouse repair program.

The NRC staff attended the review as inter-ested observers and in general was satisfied with the thoroughness of the board review.

SCE is expected to submit a request for restart of San Onofre Unit 1, along with a transcript of the review board meeting, during November 1980.

Prior to the SCE announced restart date of December 15, 1980, resubmittal of the ECCS evaluation will be required for NRC staff review and approval for Cycle 8 operation.

The NRC staff plans to issue their evaluation of the repair program, the ECCS reevaluation, and an environmental impact document within a few weeks following receipt of the needed SCE documentation.

Further reports will be made as appropriate.

The following abnormal occurrence was originally reported in NUREG-0090-10,

" Report to Congress on Abnormal Occurrences:

October - December 1977," and updated in subsequent reports in this series, i.e., NUREG-0090, Vol.1, No.1, Vol. 1, No. 2, and Vol. 2, No. 2.

It is further updated as follows:

77-9 Environmental Qualification of Safety-Related Electrical Equipment Inside Containment On May 23, 1980, the Nuclear Regulatory Commission issued a memorandum and order that addresses.this subject and that directed an accelerated environmental quali-fication review of safety-related electrical equipment that could be exposed to a harsh environment in the event of a design basis accident at a nuclear facility.

The NRC has requested pertinent information for all facilities and has initiated a review of these submittals.

The review is scheduled to be completed by February 1,1981.

The order requires that, by no later than June 31, 1982, all safety-related electrical equipment in all operating plants be qualified.

Further reports will be made as appropriate.

The following abnormal occurrence was originally reported in NUREG-0090, Vol. 2, No. 1, " Report to Congress on Abnormal Occurrences:

January-March 1979," and updated in subsequent reports in this series, i.e., NUREG-0090, Vol. 2, No. 2, Vol. 2, No. 3, Vol. 2, No. 4, and Vol. 3, No. 1.

It is further updated as follows:

79-3 Nuclear Accident at Three Mile Island (1)

Decontamination of the TMI-2 Reactor Building Atmosphere After review of the comments received on the draft environmental assessment and further staff analysis of alternatives, the NRC/TMI Program Office staff

17 prepared and presented to the Commission on May 30, 1980, the " Final Environ-mental Assessment for Decontamination of the Three Mile Island Unit 2 Reactor Building Atmosphere, NUREG-0662, Volume 1, May 1980."

In this final environ-mental assessment, the staff continued to recommend that controlled purging of the TMI-2 reactor building atmosphere to the environment be authorized.

The Commission authorized this action in its Memorandum and Order of June 12, 1980.

l Controlled purging of the TMI-2 reactor building atmosphere was initiated at approximately 8:00 a.m. on June 28, 1980.

Approximately 4 minutes into the purging, the system was shut down due to high radiation alarms on the particu-late detectors.

Subsequent licensee analysis of the monitor sample system filters revealed no particulate activity.

It was then concluded that the particulatc detectors were responding to the noble gas (Kr-85) concentration in the system.

EPA and NRC independent analysis reaffirmed this conclusion.

The radiation monitor was modified to eliminate the interference due to noble gas, and test purging was resumed between 5:00 p.m. and 10:00 p.m. on June 28, 1980.

During this test purging, additional filter samples were taken and ana-lyzed to reaffirm that no particulate activity concentrations were present.

All analyses by the NRC were in agreement with the licensee's results.

Purging resumed at 2:00 p.m. on June 29, 1980, and by 12:00 midnight on Jure 30, 1980, 4295 Curies of Kr-85 had been discharged.

All releases were made in accordance with the Commission's Order, the Technical Specifications, and the licensee's procedures.

The NRC Region I mobile laboratory was onsite and was used to verify the licensee's analytical results.

The NRC/TMI Program Office staff provided 24-hour onsite coverage of the purging operation.

Purging continued until July 11, 1980.

The purge systems were shut down based on the licensee's declaration that the purge was essentially complete.

Calcu-lations indicated a total release of about 43,800 curies.

In the days following the purging, periodic sampling of the reactor building atmosphere showed a gradual increase in Kr-85 concentration.

The water in the reactor building sump is suspected to be off gassing Kr-85.

The licensee made plans to purge the reactor building using a modified purge system in accordance with their Environmental Technical Specification release rates.

Subsequent periodic purges of the reactor building are also being planned.

(2) Reactor Building Personnel Airlock On May 20, 1980, the first post-accident manned entry into the TMI-2 reactor building was attempted.

However, the inner door of the airlock could not be opened; therefore, the entry attempt was aborted.

Subsequent evaluations dis-closed that opening of this door was being prevented by a differential pressure interlock pin which had not retracted.

Attempts to retract this pin were unsuccessful until a 1-inch-diameter hole was drilled in the door on June 30, 1980.

The pin was then manually actuated to its retracted position.

A func-i tional test of the door's hand-wheel was conducted on July 3, 1980. A hhough no attempt was made to actually open the door at that time, this test indicated l

l

18 that the door should now be capable of being opened.

On July 16, 1980, the door was successfully opened.

However, upon closure of the door, the seal rings for the door failed a leakage test which was conducted following the test of the door operability.

Temporary measures (such as mating surface cleaning, taping) were used to attempt to correct the excessive seal leakage.

However, a subsequent seal leakage test conducted on July 16, 1980, on the inner door also failed to meet the test acceptance criteria.

A seal leakage test was successfully conducted on the outer door.

It was decided to wait until the first actual containment entry to perform a more thorough cleaning of the door seals / mating surfaces to correct the leakage problem.

While the airlock doors were open, radiation surveys were taken inside the reactor building near the inner door.

General area radiation levels ranged from about 20 mr/hr at the door to approximately 700 mr/hr 12-15 feet from the door.

(3) Reactor Building Entry On July 23, 1980, two licensee engineers made the first post-accident entry into the reactor building.

The entry commenced at 10:06 a.m. and was completed 20 minutes later.

The entry was limited to a sector of the reactor building between the concrete shield ring around the reactor and the reactor building wall.

The entry team remained on the 305-ft elevation (ground level).

Measure-ments indicated that each entry team member received a gamma dose of less than 200 mr.

Gamma radiation readings inside the reactor building ranged from 400 to 700 mr/hr in shielded areas over the concrete floor surfaces.

Readings of 10 r/hr were recorded over the two reactor building stairwells.

The reading in the vicinity of the reactor building air coolers was 7 r/hr.

In most areas, beta readings ranged from 1 rad /hr at waist level to 2 rad /hr on contact with the floor surface.

Photographs and swipe surveys were also taken.

The photographs indicated that the equipment inside the building was intact.

During the entry, a health physics technician, who was stationed inside the personnel airlock, cleaned the inner door seal surface, and the door was suc-cessfully leak tested following the entry.

Further reports will be made as appropriate.

The following abnormal occurrence was originally reported in NUREG-0090, Vol. 3, No. 1, " Report to Congress on Abnormal Occurrences:

January-March 1980," and is further updated as follows:

80-1 Occupational Overexposure to Skin and Extremities The independent dose assessment of the six individuals has been performed by a consultant.

It is the conclusion of the consultant that the calculated beta

19 dose assignment is probably conservative in the limiting organ aid sufficient documentation exists to support the reported doses.

Enforcement action con-cerning these overexposures is pending and the onsite NRC inspectors are reviewing licensee progress in upgrading of the Met-Ed Health Physics program to prevent recurrence.

Generic aspects of the event are under review to assess possible inadequacies of present practices and regulatory requirements for occupational radiation monitoring in post-accident plant environments.

During routine and special health physics appraisal inspections at all operating reactor licensees, NRC inspectors are reviewing survey equipment inventories to make certain that proper types of equipment are available and that instruc-tions and calibration curves for this equipment are available and are being used.

Further reports will be made as appropriate.

i i

i 1

{

20 APPENDIX C OTHER EVENTS OF INTEREST The following events are described below because they may possibly be perceived by the public to be of public health significance.

None of the events involved a major reduction in the level o' protection provided for public health or safety; therefore, they are not reportable as abnormal occurrences.

1.

Construction Deficiencies During NRC investigations conducted between November 1979 and February 1980 of construction activities at the Washington Public Power Supply System Washington Nuclear Project 2 facility and their contractor's shops in Seattle, Washington, several problems were identified that demontrated inadequacies in the licen-see's quality assurance program.

Significant deficiencies were found in the construction of the sacrificial shield wall, a steel and concrete cylindrical structure surrounding the reactor vessel which is designed to support piping systems within the primary containment vessel and also protect plant personnel and equipment from harmful levels of radiation.

Several welds in the shield wall were improperly made and the wall as constructed may have been incapable of withstanding shear forces resu-lting from postulated accident conditions.

Also, voids existed in the wall such that adequate shieldins from radiation within the primary containment vessel would not have been provided.

These findings, together with other identified quality assurance problems associated with the work of other contractors, led to the cessation of further work on the sacrificial shield wall on November 21, 1979, and the NRC issued a notice of proposed imposition of civil penalties on June 17, 1980, in the cumulative amount of $61,000.

The licensee has submitted plans for repair and modification of the sacri-ficial shield wall.

These plans are being reviewed by the NRC.

The licensee is also required to submit a plan and schedule for reviewing completed safety-related work accomplished by certain contractors to determine whether such work was properly performed.

The licensee's response to.the proposed civil penalty and a description of the corrective actions to be taken to resolve the quality assurance inadequacies were received by the NRC on July 17, 1980, and are presently under review by the NRC staff.

The deficiencies were found during the normal inspections performed for a plant under construction.

There is no reason to believe that the deficiencies would have remained undetected until after the plant was fueled and operating.

Therefore, there was no major reduction in the degree of protection of the public health or safety, and the event is not considered reportable as an abnormal occurrence.

21 2.

BWR Jet Pump Assembly Failure On February 2, 1980, Commonwealth Edison Company (CECO) reported that a jet pump failed in Dresden Unit 3 while operating at about 67% of full power and in a coastdown mode to a refueling shutdown.

The Dresden Unit 3 nuclear plant utilizes a boiling water reactor, designed by General Electric Company (GE),

and is located in Grundy County, Illinois.

The jet pumps, located peripherally outside the core shroud and inside the reactor vessel, are part of the recircu-lation system which provides forced convection cooling of the reactor core.

Dresden Unit 3 contains 20 jet pumps, 10 in each of the 2 recirculation loops.

Observed changes in plant parameters during the event indicated a jet pump failure had occurred.

In accordance with Technical Specifications, an orderly plant shutdown was begun to bring the unit to cold shutdown within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Upon further investigation, pump No. 13 was determined to have failed.

Following vessel head removal and defueling, TV camera and visual inspections of the jet pumps and vessel annulus revealed that the holddown beam assembly had broken which allowed the jet pump components to disassemble.

Subsequent nondestructiva examinations of the remaining 19 holddown beams identified crack indications in 6 additional beams.

All defective beams were subsequently removed and replaced with new beams.

Investigations by GE showed that intergranular stress corrosion cracking under sustained loading was the cause of the beam failures.

GE notified utilities having operating BWR units with jet pumps of the problem on March 19, 1980.

Also during March, while Quad Cities Unit 2 and Pilgrim Unit 1 were shut down for refueling, examinations revealed crack indications in one beam and three beams, respectively.

These beams were replaced during the refueling outages.

On April 4, 1980, the NRC issued IE Bulletin No. 80-7 to all GE facilities utilizing jet pumps. The bulletin described inspections required to assure jet pump integrity, and described a jet pump surveillance program which would allow the detection of an impending holddown beam failure well in advance of such a failure occurring.

The jet pump failure at Dresden Unit 3, while significant, was not considered to be an abnormal occurrence.

To result in a significant reduction in the degree of protection of the public health or safety, the failure would have to be superimposed on two additional failures, i.e., a design basis recirculation pipe break - loss-of-coolant accident coupled with the worst single active failure of a component (in this case, failure of the low pressure coolant injection valve); in this case, short-term and long-term core reflood capa-bility would be significantly degraded.

The Dresden Unit 3 plant restarted on May 3, 1980, following replacement of the defective beams, implementation of a jet pump surveillance program, and refueling / maintenance activities.

1 22 l

3.

Show Cause Order - Irigaray Solution Mining Project J

On April 21, 1980, the NRC's Director, Office of Nuclear Material Safety and Safeguards, issued a Show Cause Order to Wyoming Mineral Corporation, owner and operatcr of the Irigaray Uranium Solution Mining Project, to immediately suspend normal production operations at the Irigaray site due to evidence of uncontrolled vertical excursions of leaching solutions.

Any solution excur-sions in the vertical or horizontal directions fa* nne*atises such as these are of concern to NRC, for such excursions have tm pnentid to contaminate 1

water supplies.

Since this is an h situ uraniJG MQvery PCeration, no individual was exposed as a result of the excuri<o!s.

E:Wed on the results of shallow zone well drilling and the sampling pro g a.a conducted by the licensee, the data show that the lixiviant and pollutants are, in the short term, being confined laterally to the well field vicinity.

Therefore, in the short term, there has been no major degradation of the environment.

Based upon this information, the event is not consir'ered an abnormal occurrence, for the event did not involve a major reduction 0 the degree of protection to the public health, safety, or environment.

On May 23, 1980, an immediately effective Order modifying the license and terminating the Show Cause Order of April 21, 1980, was issued to the licensee.

Licensee modifications directed by this Order included:

(1) requiring the licensee, within 90 days, to demonstrate that continued operation of this project should be authorized by providing a detailed geologic and hydrologic analysis that shows that control of the M situ leaching process and restoration of the groundwater are achievable, (2) changing lixiviant to sodium carbonate /

sodium bicarbonate solution from previously used ammonia-based solutions, (3) continuing monitoring of diagnostic wells, (4) conducting additional aquifer pumping tests, (5) limiting production to previously mined well fields, (6) continuing efforts to remove existing contamination.

The licensee submitted the required information within the 90-day time frame.

The NRC will make a determination in the near future on the advisability of continued operation and what additional license conditions will be warranted.

4.

Reactor Coolant Pump Seal Failure On May 10, 1980, Arkansas Power and Light Company reported that a reactor coolant pump seal had failed at Arkansas Nuclear One - Unit 1 (ANO Unit 1),

a pressurized water reactor (PWR), located in Pope County, Arkansas.

The nature of the failure of a reactor coolant pump seal is a degradation of the primary coolant pressure boundary.

This boundary is one of three barriers designed to contain radioactive materials generated by the nuclear reactor - the other two are the fuel cladding and the containment boundary.

Each reactor coolant pump has a shaft seal system which is to maintain essentially zero reactor coolant leakage; should seal leakage occur, it requires a reactor shutdown when the leak rate exceeds 10 gallons per minute.

23 At 0145 on May 10, 1980, while ANO Unit 1 was at approximately 86 percent full power, operations personnel were taking reactor coolant system (RCS) leak rate data when a step decrease in makeup tank level occurred, indicating an unusual RCS inventory loss.

The "C" reactor coolant pump (RCP) seal instrumentation indicated either a seal failure or piping break.

Based on this information, power was reduced in preparation for shutdown as regi: ired by the license.

Due to the observed indications, operations personnel actions were governed by the small-break loss-of-coolant response procedure.

The NRC Resident Inspector l

and NRC Headquarters Maryland were notified of the controlled shutdown.

After 62 minutes, the main turbine generator was tripped by the operators.

The power reduction initially started at a rate of approximately 5 percent per i

minute decrease when the estimated RCS leak rate was 10 to 20 gpm; when the RCS leak rate increased, the load reduction rate was increased to approxi-mately 20 to 30 percent per minute.

The "C" RCP was stopped 1 minute after the turbine was taken offline with the reactor still critical.

The RCS leak rate then took a step increase to an estimated maximum leak rate of 350 gpm.

The "C" RCP lift pumps were started and stopped four times in succession and, after the fourth try, caused a decrease in RCS leak rate.

The reactor was manually scrammed by the operators from approximately 10 percent full power 3 minutes after the main generator shutdown.

In order to maintain pressurizer level and RCS pressere following the reactor trip, the two additional high pressure injection pumps were manually started, and all four high pressure injection valves were manually opened.

The "C" RCP seal return was then isolated to prevent RCS inventory loss through the seal return line, and seal flow was increased to quench the steam / water that was leaking by the failed seal.

At this time, operators noted that the reactor building pressure had increased from 14.7 psia to 15.2 psia and that radiation levels had increased, confirming that RCS leakage was inside the containment.

The operators then put the reactor building emergency coolers in service to reduce the containment building pressure increase.

Shortly thereafter, the operators secured the "A" Reactor Coolant Pump.

Fifteen minutes after starting the two additional High Pressure Injection (HPI) pumps, the "C" HPI pump was secured and normal RCS makeup established with "A" and "B" makeup pumps taking suction from the Borated Water Storage Tank (BWST).

The RCS was then cooled down with a relatively high reactor coolant system cooldown rate, and depres-surized to minimize pressure on the RCP seals and thus leakage through the seals.

l To prevent discharging the Core Flood Tanks into the Reactor Coolant System during depressurization, one person from the operations staff and one health i

physics technician entered the containment building to energize the core flood tank discharge valves controls which are required to be locked out in certain operational modes.

The two persons were in the reactor building for about five minutes receiving 53 mrem and 44 mrem exposures, respectively.

The State Health Department and the Office of Emergency Services were notified i

at 0850.

The RCS cooldown was essentially complete at 0900 with the decay heat removal systems in service and all four reactor coolant pumps off.

The margin to saturation in the RCS hot leg was always maintained greater than l

24 100 F.

The total amount of water transferred to the RCS from the BWST was estimated to be 64,000 gallons as of 0900:

approximately 25,000 gallons were required for makeup as a result of RCS shrinkage and about 39,000 gallons collected in the reactor building basement during cooldown.

Subsequently, 11,000 more gallons were drained from the hot legs to the containment building.

Radioactivity levels at the stack and reactor auxiliary building areas were at background levels.

There were no personnel injuries or high radiation exposures.

Following NRC approval, the reactor containment building was vented on May 13, 1980.

The release was monitored by the EPA, the State of Arkansas, and the licensee.

Venting was completed on May 15 and monitoring results detected traces of Xenon-133 that were a small fraction of the regulatory limits.

The water that collected in the basement of the Reactor Building was reprocessed for later use in the RCS.

A failure investigation was initiated, which included:

(1) Examination of the failed seal, the three remaining seals, and one spare seal.

(2) Review of the seal failure history at the facility and other power reactors with pumps of similar design.

(3) Review of the possible failure mechanisms.

(4) Recommendations to prevent recurrence.

Damage to the RCP "C" seal was severe.

The licensee replaced the seals on all four reactor coolant pumps at ANO Unit 1.

The licensee is working closely with Byron Jackson, the pump manufacturer, and Babcock and Wilcox, the Nuclear I

Steam System Supplier, on the failure analysis investigations.

The problem was reviewed with the plant operators and maintenance personnel.

The licensee also relocated the electrical breakers for the core flood tank discharge valves outside of the reactor containment building for ease of access.

The NRC investigated the various aspects of the incioant.

The NRC met with the licensee and reviewed the planned corrective actions and the results of the licensee's preliminary analyses.

The NRC is continuing with these review efforts.

In conjunction, the NRC has been following the various types of seal failures that have occurred on reactor coolant pumpr, and the industry efforts related to pump seal development to improve reliability and availability, including efforts to reduce failure modes and frequency.

In addition, the NRC licensing staff is reviewing concerns on large amounts of water collecting in the containment during operational transients and accidents, since useful equipment may become submerged and fail to operate.

Several other reactor plants have experienced excessive RCP seal leakages.

One, which occurred in May 1975 at H. B. Robinson Unit 2, was reported as an l

25 t

l abnormal occurrence in NUREG-75/090 (" Report to Congress on Abnormal Occurrences:

January-June 1975").

The latter event was more serious since abnormal procedures were implemented due to lack of operating limits and instructions for operating under these conditions; this and other problems compounded the damage and delayed shutdown operations.

Although the pump seal failure at ANO Unit I was unusual and a degradation of the primary coolant pressure boundary, the event did not involve a major reduc-tion in the degree of protection of the public health or safety and consequently is not considered to represent an abnormal occurrence.

The plant responded as designed and expected and plaat personnel reacted in an established and orderly manner to safely shut down the plant.

Neither fuel integrity nor the containment boundary was degraded.

5.

Development of Steam Void Under Vessel Head During Reactor Cooldown On June 11, 1980, Florida Power & Light Company reported an event at St. Lucie Unit 1.

This nuclear plant utilizes a pressurized water reactor designed by Combustion Engineering, and is located in St. Lucie County, Florida.

At 2:26 a.m., with St. Lucie Unit 1 at full power, an electrical short across a solenoid valve terminal board caused one of two series containment isolation valves in the component cooling water (CCW) return from all reactor coolant pumps (RCP) to fail closed.

The electrical short resulted from environmental effects of a minor steam leak in the immediate vicinity of the solenoid valve.

After unsuccessful attempts to restore CCW flow, the reactor was tripped manually at 2:33 a.m.

Within 2 minutes, all four RCPs were also manually tripped.

Natural circulation cooldown was initiated at approximately 3:00 a.m. to prevent damage to the RCP seals.

Component cooling water flow to the RCPs was restored at 3:50 a.m.

Although variations in seal leakoff flow rates were observed, the seals on the four idle RCPs did not fail.

St. Lucie has Byron Jackson reactor coolant pumps with mechanical seals.

Controlled reactor coulant bleedof f flow is used for seal cooling and lubrication.

The pumps do not have a supplementary seal water injection system.

l The natural circulation cooldown continued uneventfully until after 6:00 a.m.

l The highest cooldown rate achieved was approximately 65 to 70 F per hour, which is withir operational limits.

Between 6:00 a.m. and 6:30 a.m., RCS pressure was reduced from 1140 to 690 psi by charging water through the pressurizer j

auxiliary spray line.

Around 7:00 a.m. while still charging via the auxiliary spray line, pressurizer level increased at rates faster than the rate at which water was being added.

Pressurizer level then experienced wide variations which continued for approximately five hours while the cooldown and depressurization continued.

The pressurizer level variations have been shown to be due to the formation of a relatively large steam void in the reactor head area that persisted for a number of hou"s.

The void was due to a temperature lag between the bulk coolant and the vessel head area because of lower cooling flow in the head area during

26 natural circulation cooldown.

Gas concentration in the RCS was not high enough to cause a significant volume of gas to come out of solution.

The indicated subcooling margin of the bulk coolant ranged between 220 and 150 F when the reactor head steam void developed. The indication of minimum required subcool-ing of 50 F was not approached during the cocidown until around 12:19 a.m. at a pressurizer pressure of approximately 110 psig.

The reactor designer has calculated that the steam void could have been as large as 700 cubic feet.

A volume of more than 1000 cubic feet would be required before any steam would have been swept into the RCS piping.

The event is significant because it is an example of a natural circulation coo.;-

down during which a steam void large enough to cause large, rapid variations in pressurizer level formed under seemingly normal conditions.

In addition, the possibility of a total loss of component cooling water to reactor coolant pumps due to the single failure of any one of four CCW containment isolation valves was highlighted.

In a pressurized water reactor, steam voids in the reactor coolant system (in locations other than the pressurizer) are generally undesirable and an opera-tional concern.

Steam voids form in the reactor coolant system when the system pressure is reduced below the saturation pressure corresponding to the highest temperature of the reactor coolant.

A safety concern develops when the presence of steam voids is not recognized or proper corrective actions are not taken.

Aside from the condition that existed during the natural circulation cooldown at St. Lucie, steam void formation has previously been noted in transient events at pressurized water reactors when the reactor coolant system has been rapidly cooled down causing a severe shrinkage of the reactor coolant and a decrease in reactor coolant system pressure. With the exception of the Three Mile Island accident, the high pressure safety injection (HPSI) system has been successfully used to refill and repressurize the RCS system and collapse steam voids formed during a transient.

St. Lucie did not use HPSI as they did not want to over-pressurize the seals.

Based on the studies performed to date, no significant safety problem has been identified; therefore, the event is not considered an abnormal occurrence.

However, the event remains under study by the NRC's Office of Analysis and Evaluation of Operational Data.

6.

Public Concern Over Groundwater Contamination An article in a local newspaper in the Wood River Junction, Rhode Island area, caused considerable amount of public concern by quoting a U.S. Geological Survey report on groundwater contamination in monitoring wells at Wood River Junction.

The data reported in the article was 1977 data that the NRC staff had reviewed with the licensee and appropriate Federal and State authorities. The contami-nation in the aquifer apparently came from a liquid effluent storage pond located on United Nuclear Corporation (UNC) property. When this was discovered, UNC ceased using this pond and fabricated new plastic-lined ponds to prevent any further seepage into the groundwater.

l

27 On June 23, 1980, the Governor of Rhode Island wrote NRC Chairman J. F. Ahearne expressing concern about the underground water contamination and the ongoing decommissioning of the UNC facility.

The Governor also requested that the NRC staff discuss this matter with him and State officials.

Mr. W. T. Crow of NRC's Office of Nuclear Material Safety and Safeguards met with the Governor's staff on July 8, 1980, and attended a public meeting held in the evenir.g of the same day in the Town of Charlestown, Rhode Island, to discuss the UNC situation.

The responses to the Governor's concerns, expressed in his letter to the Chairman, appeared to satisfy State and local officials as well as the public.

Media interest was high; reporting was factual and stressed the positive aspects of the situation.

The releases involved were below the concentrations specified in 10 CFR 20, Appendix B, Table II, which specifies the concentrations of radionuclides in water that may be released to an unrestricted area.

Therefore, since there was no major reduction in the degree of protection of the public health or safety, this event is not considered reportable as an abnormal occurrence.

I

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RYh 00k U.S. NUCLE A3 REGULATORY COMMIS$40N BIOLIOGRAPHIC DATA SHEET Vol. 3 No. 2 4 TtiLE AND SU8 TITLE (Add volume No., of worconate)

2. (Leave b:enk)

Report to Congress on Abnormal Occurrences April-June 1980 3 RECIPIENT 3 ACCESSION NO.

7. AUTHOR (S)

5. DATE REPORT COMPLE TED MONTH l YEAR November 1980

9. PE RFORMING ORGANIZATION N AME AND MAILING ADDRESS (inc'uele Inp Codel DATE REPORT ISSUED U. S. Nuclear Regulatory Comission uCNTa lvEaa Office of Management and Program Analysis November 1980 Washington, D. C. 20555 s.(teeve uan*>

8. fleave Nanki 12 SPONSORINC ORGANIZ ATION N AME AND MAILING ADDRESS (include 2,p Cooel

10. PROJECT / TASK / WORK UNIT No.

U. S. Nuclear Regulatory Commission Office of Management and Program Analysis M CONTRACT NO.

Washington, D. C. 20555 13_

• TYPE OF REPORT PE RIOD COVE RE D Iloc/usere dates)

Quarterly April-June 1980

15. SUPPLEMENTARY NOTES

14. ILeave o'"'* /

16. ABSTR ACT 000 words or less)

Section 208 of the Energy Reorganization Act of 1974 identifies an abnormal occurrence as an unscheduled incident or event which the Nuclear Regulatory Comission determines to be sign.'icant from the standpoint of public health or safety and requires a quarterly report of such events to be made to Congress. This report, the twenty-first in the series, covers the period April 1 to June 30, 1980.

There were two abnormal occurrences at the nuclear power plants licensed to operate.

One involved the loss of decay heat removal capability. The other involved the failure of control rods to insert fully during a scram. There were no abnomal occurrences at the fuel cycle facilities (other than nuclear power plants). There were no abnormal occurrences reported by the Agreement States.

This report also contains information updating previously reported abnormal occurrences, i

l

17. KE Y WOR DS AND DOCUME NT AN ALYSIS 17a DESCRIPTORS i

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17o IDENTIFIE RS OPEN ENDED TERMS l

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