ML20128P828
| ML20128P828 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry, Diablo Canyon, 05000000 |
| Issue date: | 05/13/1985 |
| From: | Felton J NRC OFFICE OF ADMINISTRATION (ADM) |
| To: | Freire A AFFILIATION NOT ASSIGNED |
| References | |
| FOIA-85-240 NUDOCS 8507130314 | |
| Download: ML20128P828 (2) | |
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'o, UNITED STATES N
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NUCLEAR REGULATORY COMMISSION g
j WASHINGTON, D C. 20555
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MAY i s 3 5 Mr. Antonio Freire MM3 A-DIV USS Niagara Falls, AFS-3 FP0 San Francisco IN RESPONSE REFER "A" Gang 96673-3032 T0'F01A-85-240
Dear Mr. Freire:
This is in partial response to your undated letter in which you requested, pursuant to the Freedom of Information Act (FOIA), records related to fuel loading and/or low power testing of the Diablo Canyon nuclear power plant; a copy.of the damage claim against NRC filed by GPU on December 8, 1980; and a copy of the September 1980 report prepared by the NRC Office for the Analysis and Evaluation of Operational Data (AE0D) regarding interim equipment and procedures at the Browns Ferry nuclear power plant to detect water in discharge.
We are enclosing copies of the two documents listed on the enclosed Appendix A which respond to your request.
The NRC has not completed its search for and review of any additional documents which may be subject to your request. We will communicate with you again when search and review are completed.
Si erel
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.' M. Felton, Director Division of Rules and Records Office of Administration
Enclosures:
As stated 8507130314 850513 PDR FOIA FREIRE85-240 PDR
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Re: FOIA-85-240 APPENDIX A 1.
09/80 AEOD Report: The Interim Equipment and Procedures At Browns Ferry to Detect Water In The Scram Discharge Volume (34 pages) 2.
12/08/80 GPU Danaga Claim Against NPC (34 pages)
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..,1 AE00/C002 REPORT ON THE INTERIM EQUIPMENT AND PROCEDURES AT BROWNS FERRY TO DETECT WATER IN THE SCRAM DISCHARGE VOLUME by the OFFICE FOR ANALYSIS AND EVALUATON OF OPERATIONAL DATA September 1980 i
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Prepared by:
George Lanik l
l NOTE: This report documents results of studies completed to date by the Office for Analysis and Evaluation of Opera-f(Jntional Data with regard to a particular operating event.
M VThe findings and recommendations contained in this report f
are provided in support of other ongoing NRC activities
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concerning this event.
Since the studies are ongoing, F1 the report is not necessarily final, and the findings and recomendations do not represent the position or efybg
- ' u " M / requirements of the responsible program office of the l
0 Nuclear Regulatory Comission.
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.. ' s EXECUTIVE
SUMMARY
On June 28, 1980, the Browns Ferry Unit 3 reactor experienced a partial failure to scram while shutting down for a scheduled outage.
As reported by the Office for Analysis and Evaluation of Operational Data (AE0D) on July 30, 1980, the apparent cause of this event was found to be water accumulation in the Scram Discharge Volume (SDV) prior to the attempted scram.
The AE00 study identified possible fundamental deficiencies in the SDV which cast doubt on the ability of the Scram Discharge Volume / Scram Instrument Volume (SIV) to adequately perform their intended functions.
In view of these deficiencies, AEOD recommended design changes to improve the performance of the scram system
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for the long term.
Following the event, the Office of Inspection and Enforcement (IE) issued Bulletin 80-17 and Supplement Nos. 1, 2, and 3.
Supplement 3 was issued in response to the concerns raised by the AEOD memorandum of August 18, 1980 which identified degraded air pressure in the control air system as a mechanism which could rapidly fill the SDV.
The equipment and procedural changes required I
by Bulletin 80-17 and its Supplements are intended to provide the basis for continued operation of BWR's during the period prior to completion of design changes to the scram system.
AE00 has evaluated the procedures and equipment at the Browns Ferry Units 1, 2 and 3 to determine their adequacy with respect to providing assurance that the SDV will not fill with water and interfere with a successful scram.
This
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evaluation applies specifically to the Browns Ferry units.
However, the l
findings and recommendations should be considered in the review process for all applicable BWR's.
The principal findings of the study are summarized below:
e The present system, which uses recently installed ultrasonic water detec-tion equipment and special procedures, in conjunction with previously installed instrumentation and procedures, does not restore the level of i
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scram protection capability thought to be assured in the original design.
However, except for degraded control air pressure events, it does provide adequate assurance for the interim that, accumulation of water in the Scram Discharge Volume (from currently identified sources),which could result in a loss of scram capability,will be reliably detected and adequately responded to by the operator.
e Degraded HCU control air pressure could result in scram outlet valve leakage to the SDV which would require operator action to manually scram the reactor within a few minutes before scram capability would be com-pletely lost.
Control air related disruptions in the plant would likely also initiate a plant disturbance which would require a scram.
Such an event would be accompanied by numerous control room alarms and indications which could distract the operator from a prompt manual scram actuation.
The current system does not adequately assure sufficient time for operator diagnosis and actions for this event.
e Operating experience indicates that a significant number of reactor scrams attributed to loss of HCU control air pressure have occurred.
These provide evidence that rapid filling of the SDV is a credible event.
The principal-recommendations of the study are as follows:
An immediate manual scram should be required based on control room indi-e cation of degraded HCU control air pressure.
Review of licensee proposals l
should include consideration of the available pressure indications and procedures to assure that other alarms and indications do not divert operator attention from this priority action, Redundant HCU air header pressure instrumentation should be provided in e
the control room.
To aid the operator in quickly focusing his attention on the need for protective action, a distinctive alarm for degraded air pressure should be provided.
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i e Because of the possibility that a currently unidentified water source could result in water. accumulation in the SDV, it would be prudent to monitor the ultrasonic system alarm output in the control room and require an immediate verification of a sustained alarm by operator dispatch to the equipment.
Operability and calibration checks of the system should be continued on a schedule of once per shift.
The conclusions of the study are summarized below:
AE0D has reviewed the interim surveillance system at Browns Ferry used to detect the presence of water in the SDV.
The AE00 assessment considers the procedures and equipment changes initiated in response to IE Bulletin 80-17 with Supplements, 1, 2, and 3 to be adequate for continued interim operation of the Browns Ferry Nuclear Plant, if the recommmendations of this report relating to degraded control air pressure are implemented.
As of the date of this report, the instrumentation and procedures in place to respond to the loss of control air scenario at Browns Ferry are judged to be inadequate.
For this event the operator must respond promptly to a single indistinctive alarm for loss of control air pressure during a period when numerous alarms may be occurring.
Additionally, the operator must take actions l
outside the control room in a very limited time frame because of the absence of a pressure readout in the control room.
IE is currently taking steps to upgrade the procedure for response to the degraded control air pressure event.
l In the past, operator action to perform a vital safety function within less than 10 minutes has not been considered acceptable by the NRC.
- However, l
providing the operator with both a distinctive low pressure alarm and reliable l-air pressure instrumentation in the control room would help assure adequate operator response within the required time period.
Such an arrangement should be acceptable for the interim.
A dedicated operator with adequate alarms and i
instrumentation in the control room could provide even greater assurance of a timely manual scram.
If the provisions made to accomplish a manual scram are iii I
found to be untimely or inadequate, provisions should be made for an automatic scram on low HCU control air pressure.
For the long-term, the scram system should be upgraded according to the recom-mendations of the AE0D report of July 30, 1980.
However, the consequences of degraded air pressure in the HCU control air headers were not fully recognized at the time of that report and were not directly addressed..Although the recommended scram system modifications may be sufficient to' enable the scram system to respond to rapid inflows of water from the scram outlet valves due to degraded HCU control air pressure, design review of the long-term modifications should include specific consideration of the effects of degraded air pressure.
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TABLE OF CONTENTS PAGE EXECUTIVE
SUMMARY
i 1.
INTRODUCTION........................................................
1 2.
SYSTEM DESCRIPTION..................................................
3 2.1 Calibration and Operation......................................
6 2.2' Operating Procedures...........................................
8 2.3 Other Leakage Detection Capabilities...........................
8 2.4 Procedures for Loss of Control Air.............................
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- 2. 5 Procedures for Standby Liquid Control Initiation...............
10 3.
ANA LYS IS AND EVA LUATION.............................................
11 3.1 Water From the Previous Scram..................................
12
- 3. 2 Purge Line Inflow..............................................
13 3.3 Water or Steam from the Drain System...........................
13 3.4 Single Scram Outlet Valve Leakage..............................
14 3.5 Multiple Scram Outlet Valve Leakage from a Common Cause........
16 3.6 Degraded Control Air...........................................
17 3.7 Operating Experience with Degraded Control Air Supply..........
21 4.
FINDINGS............................................................
24 5.
RECOMMENDATIONS.....................................................
25 6.
CONCLUSIONS.........................................................
26 REFERENCES...............................................................
28 TABLE....................................................................
29 FIGURES..................................................................
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1.
INTROCUCTION On June 28, 1980, the Browns Ferry 3 reactor experienced a partial failure of the scram system while shutting down for a scheduled outage.
The operators were able to completely insert all control rods within 14 minutes of the initial scram attempt.
Because of the initial partial success of inserting rods on the first scram and because no unplanned transient requiring a scram was in progress, no immediate challenge to reactor safety and integrity developed.
As documented in the AE00 report dated July 30, 1980, (I) the cause of this event was found to be water accumulation in the East Bank Scram Discharge Volume (SOV) prior to the first attempted scram.
Following the event, IE Bulletin 80-17 and Supplement Nos. 1, 2, and 3 were issued.
These directed BWR licensees to begin surveillance of the SDV to detect the presence of water. A requirement for continuous monitoring of the SDV water level in the control room was stated in Supplement No. 1.
Scram system problems revealed by testing subsequent to the Browns Ferry event were reported in Supplement No. 2.
Supplement No. 3 was issued in response to the concerns raised by the AE00 memorandum of August 18,1980(2)
This supplement required operator actions for a loss of control air to the Hydraulic Control Units (HCUs).
The following report is an evaluation of the current measures being taken at Browns Ferry in response to the IE bulletin and supplements to prevent events of the type that occurred en June 28, 1980. This assessment was undertaken by AE0D because of its concern about the adequacy of the interim system which will be used during the period preceding the implementation of long-term corrective measures. The scope of this report is purposely limited to:
a) Browns Ferry Units 1, 2, and 3: b) the interim measures; c) selected bulletin requirements; and d) procedures and equipment in place on the date of this report.
a The findings, recommendations, and conclusions are based on information gathered through informal channels between AE00 and the Tennessee Valley Authority, the General Electric Company, and the U.S. NRC headquarters and regional offices.
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Section 2 of this report contains a description of the present equipment and procedures at Browns Ferry used to prevent a recurrence of the failure to scram event.
Section 3 provides an AEOD evaluation of the effectiveness of the present. system (equipment plus procedures) for providing a timely response to a 4ange of postulated scenarios.
Sections 4 and 5 present, respectively, the findings and recomendations. The conclusions are given in Section 6.
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2.
SYSTEM DESCRIPTION To compensate for the identified deficiencies (1) associated with the protection system instrumentation installed at Browns Ferry prior to the June 28, 1980 event, additonal hardware and operating procedures have been put in place.
The additional equipment installed at Browns Ferry for monitoring the SDV for the presence of water is an ultrasonic (UT) system.
Ultrasonic transducers are mounted on the East and West SDV header low points.
The transducer is driven by a signal generating and processing device which incorporates a cathode ray tube (CRT) display and provides an output to a strip ~ chart recorder.
Unit 3 has eleven transducers located as shown in Figure 1.
Units 1 and 2 each have four transducers located as shown in Figures 2 and 3. Unit 3 was instrumented to a greater extent to attempt to find the cause of the June 28, 1980, partial scram event.
Since completing the testing of SDV drainage, long-term monitoring has been limited to use of transducers #2 and'#7 on Unit 3; transducers #12 and #13 on Unit 2; and transducers #14 and #15 on Unit 1.
In the case of failure of these transducers, a backup transducer is available on each header.
The transducers are bonded to the headers with a high tempera-ture adhesive.
The pulse echo technique of depth measurement is used and is illustrated in Figure 4.
The top illustration shows a cross section of the SDV pipe on which the transducers are mounted.
The bottom illustration shows the CRT display arising from this situation.
Since sound travels one-fourth the speed in water as in steel, the reflection from the inner tube wall is received very quickly following the initial pulse.
This is shown on the left hand side of the CRT display in Figure 4.
Multiple reflections are seen on the CRT because of sound reflections between the inner and outer diameter of the pipe.
These show a decreasing amplitude and die out rapidly.
l The sample illustration is shown containing 5.2 inches of water.
A second series of echoes is received at a later time on the CRT indicating the i
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reflection from the water-air steface at a distance of 5.2 inches.
The instru-ment has been previously calibrated on a pipe with a known water level.
The 4
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numbers shown on the horizontal axis of the CRT display correspond to the depth in inches of water in the.SDV above the transducer location.
A continuousi recorder is provided.
By use of gating devices, it is possible to pick the signal of interest to look at which is the water-air interface and not the. pipe inside diameter (i.d.).
The gating device is set to gate signals which come in at a time later than those corresponding to one inch of water.
This eliminates the reflection frem,the i.d. of the SDV pipe and the associated multiples.. Thegatingdeviceisaisosettogateonlythosesignalswithan amplitude greater than approximately 20% of full-scale amplitude.
The first signal associated with a given guise to pass the gate is transmitted to the i
recorder.
The recorder is a twc: channel recorder;. one channel records the amplitudeoftnegatebskghal,andtheotherrecordsthecalibrateddepthof I
water associated with tha gated signal.
A local alarm is provided.
Any echo signal which passes the gate will generate an audible and visual alarm.
The alarm is generated when the water level is greater than one inch and self-clears when the level is less than one inch.
Twocharacteristicsofphegating;methodusedareofparticularinterestwith
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respect to the recorder output:
(1) only water depths greater than one inch are recorded; and (2) only the first echo received at a depth greater than one inch is recorded.'
When no water is present in he header', the echo from the pipe i.d. is the only return pulse.j Since tnis initial pdise and its multiples indicate less than one inch, nothing is gated to the recorder.
The second pulse indicating water leve'l never,comes. ;The recorder sees this as a long delayed second pulse.
Thus, the normal empty header condition reads full scale on the recorder.
l The recorder full scale reads ten inches.
Since the full pipe condition would
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read only six inches, there is no confusion in the reading.
When a pulse is
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gated, indicating the presence of water in the header, the recorder pen is driven down toward the lower part of the scale.
(See lower portion.of Figure 5).
At Browns Ferry Unit 3, two separate UT devices and recorders are provided to monitor both the East and West SDV headers independently.
On Units 1 and 2, however, a single UT device is used to drive and monitor two transducers at the same time, one on the East SDV header and one on the West SDV header.
This provides another characteristic of the system that must be recognized by those operating the system.
The gating device passes the first pulse which returns corresponding to a depth greater than one inch, and the recorder responds to this pulse.
With water in both headers, the indication seen on the recorder corresponds to the first returning echo greater than one inch.
Thus, if both East and West headers have a water depth above one inch, the smaller depth indication is recorded because it is the first pulse to return.
Individual measurements for either side can be made by disconnecting the cable from the transducer on the header opposite the side of interest.
The following is a brief discussion of the recorder output from a scram test at Browns Ferry 2 (Refer to Figure 5).
Figure 5 shows only the calibrated water depth. trace.
The amplitude trace has been omitted for simplicity.
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Increasing time is from the bottom upward with one division equal to approxi-mately 5 minutes.
Water depth is measured in inches starting from zero on the l
right hand side of the trace to 10 inches on the left.
Note that the pen location prior to the scram at 0202 hours0.00234 days <br />0.0561 hours <br />3.339947e-4 weeks <br />7.6861e-5 months <br /> is full scale left (10 inches).
This is because'no water is present and the second echo never returns, which the instrument interprets as maximum distance from the bottom mounted'trans-ducer.
The momentary readings where the pen is driven downward (to the right) prior to the scram are due to the instrument reacting to the "walkie-talkies" used for communication.
These momentary readings also activate the visual and audio alarms which clear each time the walkie'-talkie transmission stops.
Since Unit 2 was scrammed at 0202 hours0.00234 days <br />0.0561 hours <br />3.339947e-4 weeks <br />7.6861e-5 months <br /> the water level indicates 6 inches.
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.The trace has been blacked in below the 6 inch level for emphasis.
At about 5
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0230 hours0.00266 days <br />0.0639 hours <br />3.80291e-4 weeks <br />8.7515e-5 months <br /> the indicated water level falls from 6 inches to 1 inch.
This is due to the West header going empty. At the time when the level indicator reaches 1 inch, the pen is driven back up to the 6 inch level.
This is because on Unit 2, a single UT instrument is used to monitor both East and West headers.
Since the gating device passes the first returning signal above one inch, the recorder tracks the header which empties first (West side) and when the echo from the West side indicates less than one inch, the gating device begins to pass the echo from the East side header.
Since the East side header has not yet' emptied, the pen is driven back up to about 6 inches.
Between 0234 and 0256 hours0.00296 days <br />0.0711 hours <br />4.232804e-4 weeks <br />9.7408e-5 months <br />, the East side header continues to drain.
When the level reaches one inch on the East side, no returning echo is gated.and the pen returns to the 10 inch position.
As indicated on the trace, a series of momentary indica-tions of water are present at 0440 hours0.00509 days <br />0.122 hours <br />7.275132e-4 weeks <br />1.6742e-4 months <br />.
These are due to some CRD survillance tests which were run at that time.
2.1 Calibration and Operation J
4 All tranducers used in the system were tested prior to use to assure adequate performance.
The gain of the signal generating and receiving equipment is adjusted to provide an adequate signal output from the least responsive trans-ducer.
The minimum acceptable ' signal for reflection from the water interface is adjusted for 80% full scale output on the CRT display.
The gating device is set to pass any signal with an amplitude more than approximately 20% of full scale so as to provide an adequate margin of sensitivity.
The time scale on the CRT is adjusted to read the depth of the water in the header in terms of horizontal divisions on the CRT screen.
As shown in Figure 4, a total of ten horizontal divisions are used on the CRT display.
The sweep time and the horizontal centering of the CRT are adjusted so that the sixth division on the screen corresponds to a water depth of 6 inches while the first division on the screen corresponds to a water depth of one inch.
- Thus, the CRT displays water depth directly.
If no water is present, only the echo from the i.d. of the pipe is displayed on the CRT screen at a position below the one inch mark.
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Initial system calibration and later checking of the calibration is done by use of standard pipes filled with known amounts of water. Once per shift, a level two QC inspector takes a reading from a standard containing 2 inches of water and from a standard containing 6 inches of water.
This is done by disconnecting the cable from the transducers on the headers and connecting it to hand-held transducer which is held against the bottom of the standard sample pipes.
If the reading on the CRT and the recorder does not agree with the known depth of water in the standard pipes, the gain and amplitude of the UT instrument are adjusted to recalibrate the system.
At this time, all transducers are functinally checked by examining the CRT display for indica-tions of transducer deterioration.
The two UT instruments on Unit 3 are physically located at the ends of the rows of HCUs, one on the East side and one on the West side.
On Units 1 and 2, the UT instruments are located on a mezzanine level above the HCU level approximately midway between the two sides.
Browns Ferry has an auxiliary operator on each shift who observes the UT system recorder strip chart for each unit every 30 minutes.
The operator is not qualified or required to monitor the CRT output.
His sole responsibility is to monitor the strip chart recorder and the alarm.
The calibration and operability of each UT device and each transducer is checked once per shift by a level two QC inspector trained in the use of UT equipment. Communication with the control room concerning SDV water accumulation is by a hand held wal kie-tal kie.
As mentioned earlier, a separate UT transducer, CRT and recorder is provided for each side on Unit 3, while Units 1 and 2 each have a single UT, CRT and recorder to monitor both the East and West SDV header transducers.
Thus for Units 1 and 2, since the recorder tracks the first returning pulse for a water level greater than one inch, it is necessary to disconnect one lead at a time to determine separate SDV water levels.
If water is detected, a level two QC inspector must be called to verify the readings.
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2.2 Operating Procedures Procedures have been written for the control room operator to respond to the presence of water in the SDV as detected by the UT system.
If the water level reading in the SDV is less than 1-1/2 inches, procedures call for an operator to:
(1) visually verify that the SDV vent and drain valves are open; (2) check for leaks in the scram discharge valves by observing CRD temperature probe outputs and by touching the HCU discharge risers; and (3) request QA verification of the UT reading.
If the water level reading is between 1-1/2 and 2 inches, procedures call for the control room operator to:
(1) immediately request QA to dispatch a level two QC inspector to verify the reading; and (2) unless the QC inspector deter-mines that the water level is less than 1-1/2 inches, begin an orderly shutdown within one hour.
If water level exceeds 2 inches, procedures call for the control room operator to immediately begin an ord :/ shutdown without verifying the UT reading.
Plant personnel estimate that a level two QC inspector can reach the area of the UT device within approximately three minutes of being notified. At least one level two QC inspector is available for this duty'at Browns Ferry during each shift.
2.3 Other Leakage Detection Capabilities l
Leakage of water through a scram outlet valve to the SDV is recognized as one i
of the. ways for water to reach the SDV.
This leakage may be detectable by means other than the UT system or SIV instrumentation.
l Flow out of a scram outlet valve would change the flow of the CRD cooling water through the CRD seals so as to increase the water temperature at the location of the CR0 temperature probe.
At this time, no good data is available to correlate CRD temperature with scram outlet valve leakage.
In particular, 8
the rate of temperature change for a given leakage is unknown.
However, for
~the leakage rates postulated in this section, it is reasonable to assume that the temperature probe alarm set point would be reached within a relatively short time; on the order of a few minutes.
Although the CRD temperature probe alarms in the control' room, not all 185 CRD temperature probes are read simultaneously.
A sequential scan is used and it is estimated by GE that the cycle time to read all temperature probes is approximately six minutes.
Another means of inferring scram outlet valve leakage is the observation of control rod drift.
Leakage past the scram valve in excess of the CRD seal leakage would cause the associated control rod to begin to drift into the core.
A rod drift alarm is available in the control room.
Another indication of scram outlet valve leakage is movement of the scram outlet valve stem sufficient to actuate the scram valve position indication switches.
This requires a stem movement of approximately 1/32", out of a total valve stroke of approximately 3/16".
Actuation of the stem mounted switch will light the associated scram outlet valve position indication light in the control room.
One means by which the scram outlet valves can open sufficiently to leak is degraded air pressure in the HCU air header.
A low pressure alarm is provided to alert the operator at approximately 70 psia.
The actual pressure reading on the HCU header is available locally in the area of the HCUs.
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2.4 Procedures for. Loss of Control Air l
.The procedures at Browns Ferry for loss of control air were modified in response to IE Bulletin 80-17, Supplement 3, to protect against scram valve leakage on gradual loss of control air.
The details of the concern for loss of control air are discussed in the AEOD memorandum of August 18,1980.(2)
~At Browns Ferry, control room indication of the air pressure in the HCU air header is limited to a single alarm with a setpoint at 70 psia.
A local air l
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pressure guage is available at the HCUs.
Normally air pressure is maintained between 70 and 75 psia.
Since only a slight degradation of air pressure initiates the alarm, the licensee considers it undesirable to initiate a scram based on this alarm alone.
Upon receipt of the 70 psia alarm, procedures call for the control room operator to dispatch an auxiliary unit operator to read the local air pressure gauge.
Plant operati'ns personnel at Browns Ferry have o
told the NRC resident inspector that an operator will be dispatched to read the local pressure guage no later than 2 minutes after receipt of the 70 psia alarm.
If air pressure in the HCU air header is found to be 60 psia or less, the auxiliary operator informs the control room operator. The control room operator then initiates a manual scram.
Communication between the auxiliary and control room operators is maintained via walkie-talkie.
In addition to the above procedure for gradual loss of air to the HCU air header, other procedures have been implemented in response to IE Bulletin 80-17, Supplement 3.
These call for manual scram initiation in the event of:
(1) multiple rod drift-in alarms; or (2) a marked change in the number of control rods with high temperature probe alarms.
2.5 Procedures for Standby Liquid Control Initiation Bulletin 80-17, Supplement 1, requested that operating procedures be revised to provide clear guidance to the control room operator regarding initiation of the standby liquid control system (SBLC) following a failure of control rods to fully insert.
At Browns Ferry, mandatory SBLC system actuation is required by operating procedure if either of the following conditions exist:
(1) five or more
-adjacent rods are not inserted below 06 position and either reactor water level cannot be maintained or suppression pool water temperature limit of 110 F is reached; or (2) thirty or more rods are not inserted below 06 position and either reactor water level cannot be maintained or suppression pool water temperature of 110 F is reached.
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a 3.
ANALYSIS AND EVALUATION For purposes.of analysis and evaluation of the Browns Ferry failure to scram event, an effort was made by AEOD to identify water sources that could fill the SDV.
The AE00 report of August 1980, identified the following sources of water:
- 1) water left from a previous scram; 2) purge line inflow; 3) water or steam backing up from the clean radwaste drain system; and 4) inflows through the scram discharge valves.
It is recognized that this list of water sources may not be complete. However, at this time,'neither operating experience nor design review has revealed any other sources.
As discussed in Section 2 of this report, the current capability at Browns Ferry to detect and respond to accumulation of water in the SDV is based on the following elements:
previously installed SIV instrumentation; recently installed UT instrumentation; other previously installed instrumentation such as CRD temperature probes, CRD drift alarms, control air flow pressure alarm, etc; and operating procedures for response to the aforementioned instrumenta-tion.
The response of the interim system at Browns Ferry is dependent on both the instrument capability and the operator response.
Both aspects are addressed in this analysis and evaluation.
The original design, as understood prior to analysis of the Browns Ferry event, was thought to have provided continuous, redundant, safety grade and automatic protection which was functional for all water sources.
It was thought to also fail safe on loss of HCU control air pressure.
However, the analysis of the Browns Ferry event showed that this system did not work for all situations of water accumulation.
Accordingly, the original system was supplemented with a functional UT system.
This interim system is neither continuous, redundant, safety grade, nor automatic for many cases.
Further-more, its capability may be inadequate for a loss of control ~ air pressure.
That is, the interim system does not provide the same level of protection as was perceived of the original design.
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At this point,- a short discussion of the capability and reliability of the UT system will be presented.
As stated in S'ection 2, the UT system includes a CRT display of the return echo.
This is shown in Figure 4.
As stated earlier, j
an echo is received from the i.d. of the pipe and is displayed on the CRT as the left most peak.
The presence of this so-called " reference pulse" is interpreted by the inspector performing the calibration ~as verification of the I
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operability of the transducer which is being. monitored.
The calibration technique also requires that the UT system be connected to a separate trans-ducer to detect a known depth of water in a standard pipe.
These surveillance i
and calibration procedures are performed once each shift by a level two QC inspector who has extensive training in UT techniques.
We believe that because of the presence of the " reference pulse" and the high level of training of the i
level two QC personnel who performed the surveillance and calibration of this instrument,.any degradation of the operation of the UT system due to heat, vibration, radiation or other failure mode would be discovered during the scheduled surveillance.
It is recocjnized that during the period of 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> between surveillance of the UT system, it would be possible for equipment failure to go undetected.
However, because of the unlikelihood of a rapid water inflow with an accompanying need to scram occurring during the same period of the equipment failure (except for the degraded air case), this surveillance interval is judged to be adequate.
The following analysis and evaluation addresses the capability of the interim system to detect and respond to water from various postulated sources.
3.1 Water From the Previous Scram Water left from the previous scram after scram reset will be detected by the ultrasonic system.
Because of the time required between a scram and startup operations, a number of ultrasonic readings and equipment calibrations 4
would normally.take place during the shutdown.
For this situation, rapid I
detection is not required and no immediate action is needed if water is detected.
Startup would simply be delayed until it was assured that the situation was corrected and no water' remained.in the SDV.
I 12 4
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F 3.2 Purge Line Inflow Purge line inflows would be detected in time depending on the rate of inflow and the time of purging.
Purging of the SDV is an operation which is per-
~
formed when the reactor is shutdown in order to reduce accumulations of radio-activity in the SDV and its associated piping. The ultrasonic system would be used to check.for the presence of water prior to startup.
Enough time would be available during the plant shutdown for the operator action required to detect and remove all water from the SDV prior to startup.
An administrative or operator error, which allowed purging during normal operation, could provide a flow rate of water into the SDV which might not be detected soon enough by either the interim system or the original system.
However, the likelihood of a properly informed operator performing this unprecedented action is remote.
3.3 Water or Steam From the Drain System Water backing up from the clean radwaste (CRW) drain system would most probably be detected by the SIV. instrumentation of the original design.
An exception to this would be a SIV drain line blockage which could prevent flow from the CRW drain system into the SIV but would allow flow up through the SDV vent lines to fill the SDV.
However, at Browns Ferry, positive vent paths to atmosphere have been provided on the SDV vents.
Any water backing up in the vent line would be released via this path rather than to fill the SDV unless
' the backflow rate was high due to a large water release to the drain system.
Operating experience at Browns Ferry to date has shown that water backing up from the CRW drain system has actuated level switches in the SIV.
Low pressure steam backing up from the CRW drain system due to flashing hot water in the drain pipe would not be detected by the SIV instrumentation.
If the drain line between the SDV and SIV became plugged by the slow drain of
~
condensed vapor mixed with rust, the-steam backing up through the SDV vent lines would slowly fill the SDV with condensed vapor.
The positive vent paths 13
O to atmosphere would vent a portion of the low pressure steam but would not prevent the SDV from filling with condensate.
Op.erating experience at Browns Ferry Unit 1 has shown that flashing hot water can appear in the drain pipe from valve leakoff connections or other sources.
Therefore, the ultrasonic system must be used to monitor this situation and the surveillance interval must be short enough to assure timely discovery during a large hot water release.
Thus, with respect to the three identified sources of water listed above, we believe that the UT system provides adequate interim assurance that water can be detected and actions taken before the plant reaches a condition where the SDV is filled and a scram is required.
It appears that this would be true with a surveillance interval longer than the 30 minutes currently used at Browns Ferry provided that surveillance was performed prior to any start-up.
However, if protection for currently unidentified water sources and flow rates is to be provided, continuous monitoring of the UT system in the control room would be preferable to the current 30 minute surveillance interval. This would allow for a more rapid control room operator response.
As an intermediate method (between continuous monitoring and lengthy surveillance intervals) for providing response to unidentified water sources and rates, an alarm output of the UT device could be provided in the contro1~ room.
This would allow more timely operator response without the complexity of locating the complete UT system output in the control room.
Upon receipt of the sustained alarm, an auxiliary operator could be dis-patched to the UT readout located by-the HCus. We believe this approach would also provide adequate protection for unidentified water sources and flow rates.
3.4 Single Scram Outlet Valve Leakage To evaluate the adequacy of the interim system for various leak rates from the scram outlet valves, it is necessary to identify the causes of leakage.
First, it should be noted that scram valve leakage during normal operrtion is quite low.
At Browns Ferry Unit 3, tests following the June 28, 1980 event indicated an aggregate leakage of from 0 to 3 gallons per hour.
14
Discussions with GE on the leakage characteristics of the scram outlet valves indicate that any leakage is likely to cause degradation of the valve seats and could lead fairly rapidly to greater leakage. Rapid deterioration of the seating surface of one ~ valve would result in cbvious problems with the associated control rod drive but would not affect others.
One aspect of the scram outlet valve leakage problem that must be addressed is the difference in character between a leak arising from a single valve failure and that which could arise from a common mode failure leakage of many valves.
With respect to a single valve failure, the maximum inflow of water into the SDV is limited by leakage past the CR0 seals.
GE has' estimated that with the CRD seals completely destroyed, a leak rate of 10 to 12 gpm into the SDV is the maximum that could occur.
This is the rate if the flow is completely un-restricted by the scram outlet valve.
If the scram valve is only partially open or leaking, the flow rate would be less.
If the CR0 seals are intact, leakage would be expected to be in the range of 1 to 5 gpm.
Thus, a single failure of a scram valve results in only a limited flow into the SDV which would drain out with no accumulation for the current SDV drainage characteristics.(1)
Indications of scram valve leakage would be available to the operator. The CRD temperature probe alarm would be actuated.
If the scram outlet valve leakage is greater than the corresponding CRD seal leakage for a pressure differential across the piston of approximately 550 psig, the rod would move into the core.
When assessing the probability of an event that could cause problems for the SDV, it must be recognized that the probability of a simultaneous failure of more than one scram outlet valve at a given time is very low.
Multiple valve failures would have to occur simultaneously before the drainage capabilities of the current system would be challanged.
Because of (1) the low probability of this event, (2) the likelihood of early detection by rod drift alarm or CRD temperature probe alarm, and (3) because the event does not cause an accompanying plant disturbance, this postulated event is not considered to be a serious concern for the interim period.
The interim precautions should be adequate to protect against failures of this type.
15
i 3.5 Multiple Scram Outlet Valve Leakage from a Common Cause Multiple scram outlet valve leakage due to a common cause can raise serious concerns about the ability to scram the plant successfully.
To date, the only plausible common cause which leads to substantial leakage of a large number of scram outlet valves is degraded air pressure in the control air header for the HCUs.
Loss of air pressure in the control air header has occurred due to a variety of reasons such as failure of an air compressor, improper valve alignment, clogging of filters and dryers, and severance of an air line.
I As air pressure in the header decays, the s' cram outlet valves, which are held closed by air pressure, begin to open.
Although the exact pressure at which a given valve begins to open depends on manufacturing tolerances, the pressure i
for a group of valves is in the range of about 40 to 45 psia.
Information from GE indicates tLat a leakage flow of from 1 to 5 gpm out of a scram outlet valve for a given drive could occur without producing rod motion.
The actual value for a given dri/e would depend on the condition of the seals in that particular drive.
GE has stated that for a typical reactor, if the scram i
discharge valve flow rate to produce rod motion for each individual CRD was averaged with the scram discharge valve flow rate to produce rod motion of all other CRDs, the average would be in the range of 2 to 3 gpm.
With this information it can be postulated that a degraded control air pressure condition could exist for which leakage from a large number of scram outlet valves could exist without producing a scram, In fact, depending on the
{
number of scram valves which partially open and the leakage rate of these valves, it would be possible to generate a.significant flow of water into the SDV without producing significant rod motion.
It is recognized that the l
possibility of the actual occurrence of high flow rates without rod insertion i
depends ~on three factors:
(1) the control air pressure degradation pattern, (2) the range of air pressure over which the scram outlet valves open, and (3) i the seal leakage rate of the CRD associated with each particular scram outlet valve.
However, with the data given above, a flow rate in the range of 1 to i
O 16 l
4 1
2 gpm per drive without significant rod insertion could be possible for certain degraded air pressure scenerios.
4 i
3.6 Degraded Control Air l
Assuming an average l'eak rate that could be generated without significant rod
~
motion (given a specific degraded air pressure) of 2 gpm per CRD, a total of 2 x 185 or.370 gpm flow into the SDV would occur.
Although this large flow rate appears feasible within the characteristics of the system, lower rates of
. leakage to the SDV could also be generated by the same mechanism, and indeed are more probable.
These are discussed below in a framework of average steady-state flow rates.
It is recognized that an actual air system failure would likely lead to continuously changing leakage rates, but the air pressure 1
i degradation might level off and thereby stabilize the leakage rate at any point.
For purposes of evaluation, inflow rates into the SDV can be separated into those for the East SDV header and those for the West SDV header.
Test data s
l show that for Browns Ferry Unit 3 the average drain rate of the East SDV header is normally about 12 gpm with its vent and drain valves open.
- Thus, any steady state in-leakage of less than 12 gpm would not result in water l
accumulation in the East header unless the East side drain line were blocked.
l Similarly, test data show that the average drain rate of the West SDV header, with its vent and drain valves open is normally about 24 gpm.
Thus, any f
steady-state in-leakage of less than 24 gpm would not result in water accumu-lation in the West header unless the West side drain line were blocked.
Test data also show that the average drain rate of the SIV, with the vent and drain lines and valves functioning normally, is about 35 gpm.
To a first approximation, from the above test data and assumptions, the following general statements can be made:
1 1)
For a steady-state in-leakage below approximately 12 gpm per side, no water accumulation would occur, no water measurement with UT is required, and no operator action is necepsary.
17
2)
For a steady-state in-leakage between approximately 12 and 24 gpm per side, water would accumulate on the East side.
As an example, for a steady state flow rate of 24 gpm into each header, the West side would remain empty and the East side would fill within approximately 25 minutes.
The current 30-minute' surveillance interval at Brownv Ferry using the UT system might not detect this accumulation before filling of'the East side.
Also, because the SIV drain rate is greater thary the inflow rate from both the SDV sides to the SIV, the 50 gallons scram level switch would probably not activate.
However, the 3 gallon and perhaps the 25 gallon level switches might be activated.
This level of inleakage could result in a scenario similar to the Browns Ferry event where the West side rods scrammed successfully but the East side rods did not.
3)
For a steady-state in-leakage above approximately 24 gpm per side, water would accumulate in both the East and West side SDVs. As an example, for a steady state flow rate of 36 gpm into each header, the East header would fill within approximately 12-1/2 minutes and the West header within approximately 25 minutes.
The current 30-minute surveillance interval at Browns Ferry using the UT system might not detect this accumulation before filling both the East and West sides.
For this case, the SIV 50 gallon level switches would probably activate somewhere between 12-1/2 and 25 minutes and initiate an automatic scram.
However, the scram capability would be limited on both the East side and the West side due to the previous water accumulation.
4)
For in-leakage at very high rates (approaching 150 gpm per side) water would accumulate in both the East and West side SDVs.
Each side would fill within 3 minutes and probably before sufficient water could flow to the SIV to activate the automatic scram switches at the 50 gallon level.
The current 30-minute surveillance interval at Browns Ferry using UT would not detect this accumulation before a probable loss of scram capa-bility.
Proper operator action would probably be required within less than 2 minutes following the initiation of this scram valve leakage rate to avoid reaching a point where it would become impossible to scram.
18
In summery, the above analysis adresses conditions of degraded pressure in the HCU control. air header which can lead to aggregate leakage rates to the SDV in the range of 24 to 300 gpm.
Flow rates at the high end range probably produce at least some rod motion and perhaps some rods might fully insert.
However, at this time there is no assurance either.
by analysis or testing that a range of leakage rates does not exist which could fill the SDV quickly with insufficient indication to the operator or time for manual scram before tha ability to scram is lost.
The above discussion of the scram system behavior is for different but constant flow rates.
This would probably not be the case for an actual degraded control air event.
The scram valve flow rate would likely pass through the different regimes as discussed above and the characteristics of a particular flow rate would apply at that time.
However, analysis or test results for a variable flow rate, which show acceptable system behavior, do not exist at this time.
Thus, inadequate basis is available to justify disregarding these concerns.
A degraded air supply can also affect the performance of the SDV vent and the SIV drain valves.
Tests done at Browns Ferry show that the SDV vent and the SIV drain valves begin to close at a control air system pressure of about 17 psig.
Thus, the drain and vent valves will remain open during the type of degraded air condition that might lead to loss of scram capability.
It should be noted that the time available for operator action to respond to a degraded air condition can be separated conceptually into two phases:
(1) time available before air pressure degrades from the normal alarm set-point of 70 psig to the pressure at which scram discharge valves leakage begins (about 45 psig) and (2) time available following the beginning of scram discharge valve leakage to the time where the SDV fills to the point where a scram is no longer possible.
The analysis shows that the time available for operator action following the beginning of scram discharge valve leakage can be as little as 2 minutes.
Because of the 19
s short time available for operator action following initiation of scram discharge valve leakage, operator action should be taken prior ~to reaching a degraded or lost scram capability.
Because HCU control air degradation preceeds opening of the scram discharge valves, added time would be available if operator action were based on air pressure indications.
For a rapid air pressure degradation which stabalized at a point where large scram discharge valve leakage occurred, the benefits of operator action based on air pressure indication would be diminished.
From the standpoint of improved assurance of a successful scram during a degraded HCU control air event, however it is preferrable to scram on the indications of degraded air pressure than on the UT system.
This would be true even if UT readout were continuous in the control room.
The UT system (on indica-tion of water in the scram discharge volume) to initiate a manual scram for the degraded HCU control air event does not provide sufficient assurance that adequate time will be available for the required operator diagnosis and action.
The same can be said for reliance on CRD temperature probes, rod drift alarms, and scram outlet valve indicator lights.
IE Supplement 3 to Bulletin 80-17 requires an immediate manual scram on low HCU control in pressure at a minimum pressure of 10 psi above the opening pressure of the scram outlet valves.
This provides additional time for operator. diagnosis and action prior to possible filling of the SDV following receipt of the low pressure alarm.
However, because of the lack of any control room indication of HCU control air pressure (except the low pressure alarm at 70 psig) current procedures at Browns Ferry require an operator to be sent to the HCus to read a local HCU air pres-sure-gauge.
This local operator then reports back the local reading to the control room by walkie talkie.
Given the rapidity of the water inflow possible with the degraded air pressure condition, we judge this.
arrangement to be inadequate.
We believe that the rapid operator response required by a degraded air system condition ncessitates that adequate alarms and instrumentation be available in the control room.
- However, because the present alarm is not safety grade and is a single channel, we 20
l believe that reliance on the alarm alone to initiate a manual scram is not adequate.
Short of installing safety grade instruinentation for this function, we believe that adequate instrumentation could be provided by redundant pressure indication in the control room along with a distinctive alarm on degraded air pressure.
Furthermore, since the instrumentation is not qualified to function during certain postulated events (e.g.
earthquakes), procedures which require immediate manual scrams for such events should be considered.
It is our judgment that if the upgraded instrumentation and procedural changes discussed above are provided, then the system will be adequate to respond to degraded HCU control air for the interim period.
We believe that this analysis supports the position that a scram on degraded HCU control air is sufficient to respond to the complete range of aggregate ~1eakage rates arising from degraded HCV control air pressure as enumerated earlier in this section.
This judgement is based on (1) the additional time available before any discharge valve leakage begins and (2) the relatively low probability of a rapid air pressure degradation which stabilizes in the range of serious scram outlet valve leakage.
3.7 Operating Experience With Degraded Control Air Supply An effort was made to look at reactor operating experience relative to scrams caused by the consequences of a degraded control air supply. By looking through the Annual Report on Nuclear Power Plant Operating Experience (3-6) for the years 1974-78, a total of 21 events were found for BWRs where the description of the event mentioned a loss of control air as the initiating event leading to the scram.
The dates of these evehts are listed in Table 1.
Because of the brevity of the descriptions and the lack of records available to make a more careful study of each event, it is probable that not all of the 21
1 i
l l
events describe a loss of control air which would or could affect the HCUs.
On the other hand, some of the events seemed to be very close descriptively to l
the type of rod behavior that would be expected given a loss of control air to l
the HCUs.
For example, one event description mentioned massive rod drift.
i Another event generated an automatic scram due to high level in the SIV.
This j
. event occurred at Browns Ferry Unit 1 on November 24, 1976. Because of the known drain characteristics at the Browns Ferry units, it is likely that during this event the SDV was at least partially filled.
Because the SDV is designed to provide approximately 3.3 gallons per drive free volume, and a typical scram requires less than one gallon per drive, enough volume was available for a successful scram.
However, there is no doubt that the volume l
margin was reduced. An air degradation of a slightly different character could have lead to a water filled SDV and' inability to scram.
An effort was made by AEOD to find data that would indicate filling of the SDV during some of these events.
From the event at Browns Ferry 3 on June 28,1980,one bit of evidence that. leads to the conclusion that water was in the SDV prior to the first attempted scram was that the SIV high level scram switches were 3
activated more quickly than expected during a manual scram (18 seconds vs. 45 seconds).
This is because for a full SDV, water entering the SDV during the scram will more quickly pressurize the SDV and force water through the drain line to the SIV than if the SDV were not pressurized.
For events where an eventrecorderoutputwasavailable,nosuchchangewasMoted.
However, most events had no data from an event recorder available and no other way of recalling this data.
l In general, this search for operating experience data was unsuccessful. On the i
other hand, the argument that because a plant has successfully scrammed 21
}
times during degraded control air events does not provide a large statistical basis on which to judge the adequacy of the scram system (machine and man) for j
responding to such events.
From the observation of 20 successful scrams in 20 i
scram attempts, one can conclude that the 95% upper (one sided) confidence i
limit for the probability of a scram failure is approximately 3/20 =.15.
1 1
22 i
Alternatively, the 95% lower (one sided) confidence limit for the probability of successful scram is approximately 1 - 3/20 =.85.
Both confidence limits are computed on the assumption of a common probability of a scram attempt failing and the assumption of statistical independence of scram attempts.
These assumptions have been made for mathematical convenience; they are not necessarily plausible.
In fact, there is no doubt that the list of successful scrams includes some events where the HCU control air header system was not affected.
These would not be included in a list developed through a closer investigation of the event which would disclose that fact.
A lower number of successful scrams, due to legitimate control air degradation, even if all are successes, only detracts from the merits of the argument which claims that since no scram failures have occurred to date, the system is adequate.
t 1
4 23
_ ~ _.
i 4.
FINDINGS T
j Based on the system description and evaluation discussed in. Sections 2. and 3.
of,this report, a number of findings have been determined.
Again, it should be emphasized that these are based on Browns Ferry only.
e The present system (ultrsonic level instrumentation, existing SIV instru-mentation, and special operating procedures, etc.) should be capable of providing adequate protection during the interim against filling of the SDV due to all identified water sources except for those related to scram 4
discharge valve leakage due to degraded HCU control air pressure.
e Degraded HCU control air pressure could result in scram outlet valve leakage to the SDV which would require operator action to manually scram the plant within a few minutes before scram capability would be completely i
lost.
This event would likely be accompanied by a plant disturbance requiring a scram due to other control air related disruptions in the plant.
Such an event would be accompanied by numerous control room alarms and indications which could distract the operator from a prompt manual scram actuation.
e Operating experience indicates that a significant number of reactor I
scrams attributed to loss of HCU control air pressure have occurred.
These provide evidence that rapid filling of the SDV is a credible event.
I i
i I
I
, - - -. ~. - _ - -..,, - -, - -, _, -,,,,.,,, - - - -,,.. - - - - - -. -, -..
,,-n.,,
5.
RECOMMENDATIONS The principal recommendations of the study are as follows:
e An immediate manual scram should be required based on control room indi-cation 'of degraded HCU control air pressure.
Review of licensee proposals should include consideration of the available pressure indications and procedures to assure that other alarms and indications do not divert operator attention from this priority action.
Redundant HCU air header pressure instrumentation should be provided in e
the control room.
A distinctive alarm for degraded air pressure should be provided to aid the operator in quickly focusing his attention on the need for protective action.
e Because of the possibility that a currently unidentified water source could result in water accumulation in the SDV, it would be prudent to monitor the ultrasonic system alarm output in the control room and require an immediate verification of a sustained alarm by operator dispatch to the equipment.
Operability and calibration checks of the system should be continued on a schedule of once per shift.
25
l 1
\\
l 6.
CONCLUSIONS f
AEOD has reviewed the interim surveillance system at Browns Ferry used to detect the presence of water in the SDV.
The AE00 assessment considers the procedures and equipment changes initiated in response to IE Bulletin 80-17 with Supplements,1, 2, and 3 to be adequate for continued interim operation of the Browns Ferry Nuclear Plant, if the recommendations of this report for i
response to degraded control air pressure are implemented.
As of the date of this report, the instrumentation and procedures in place to respond to the loss of control air scenario at Browns Ferry are judged to be inadequate.
For this event the operator must respond promptly to a single in-distinctive alarm for loss of control air pressure during a period when numerous alarms may be occurring.
Additionally, the operator must take actions outside the control room in a very limited time frame because he lacks a pressure readout in the control room.
IE is currently taking steps to upgrade the procedure for response to the degraded control air pressure event.
i In the past, operator action to perform a vital safety function within less than 10 minutes has not been considered acceptable by the NRC.
However, pro-viding the operator with both a distinctive low pressure alarm and reliable air pressure instrumentation in the control room, would help assure adequate operator response within the required time period.
Such an arrangement should be acceptable for the interim.
A dedicated operator with adequate alarms and j
instrumentation in the control room could provide even greater assurance of a l
timely manual scram.
If the provisions made to accomplish a manual scram are found to be untimely or inadeqiate, provisions should be made for an automatic scram on low HCU control air pressure.
For the long-term, the scram system should be upgraded according to the recom-l mandations of the AE00 report of July 30, 1980.
However, the consequences of degraded air pressure in the HCU air headers were not fully recognized at the l
time of that report and were not directly addressed.
Although the recommended l
scram system modifications may be sufficient to enable the scram system to 26 l
L
respond to rapid inflows of water from the scram outlet valves due to degraded
- HCU air header pressure, design review of the long-term modifications should include specific consideration of the effects of degraded air pressure.
1 A
w 27
[.
REFERENCES
' 1.
AEOD MEMO (Michelson) to NRR (Denton) dated August 1, 1980 with enclosures.
2.
AE00 MEMO (Michelson) to NRR (Denton) dated August 18, 1980.
3.
USNRC NUREG-0227 dated April 1977.
4.
USNRC NUREG-0366 dated December 1977.
5.
USNRC NUREG-0483 dated February 1979.
6.
USNRC NUREG-0618 dated December 1979.
i i
I i
28
Table 1 Scrams Attributed to Loss of Air (1974-1978)
Browns Ferry 1:
8/1/74, 10/19/76, 11/24/76, 8/15/78, 8/18/78 Browns Ferry 2:
8/18/78 Brunswick 2:
4/5/77 Dresden 2:
9/7/77,7/28/78 Dresden 3:
8/15/74 Quane Arnold:
1/9/78.
Hatch 1:
3/4/76 Millstone 1:
8/6/77. 5/29/78 Nine Mile Point 1: 12/21/74 Pilgrim:
1/19/76 Quad Cities 1:
1/3/77, 4/30/78 Quad Cities 2:
7/1/74,8/31/74,10/25/77 1
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INJURY, OR DEATH
" ' " " ' " ' ' " " '"'3#~""""'~+
!?*
2 Nat.tt A*.D A:,:1!,1 CF C' Ait. TANT INun:brr. srtre t a ny. 5;.
ourJ Zut' C"Jr
- Genera:. Counsel General Public Utilities Corp.
Ntclear P.eculatory Commission 100 I.iterpace Parkway P
Washi..cten, D.C.
20555 Parsipcanv, NJ 07054 et. al Isee attachment}
f
- 3. TY71 CF EvP.cYu.!NT
- 4. AGE
- 5. A'.A3 TAL
- 6. NAME AND ACC2155 CF SPOU$(. IF ANY (Nusnber. Jfrrrt. (ify. Slute. ur.
I STATUS Zip CodeI i
= u a:Aav
= C:.r.iAu NA NA NA NA 5
- 7. F*. ACE OF ACCT TNT IGis e city ur. Iown und Sonor: if outside city limits. inditoose
- 8. DATE AND ort
- p. Tant enitruye or distuner tu nenrest city or town)
OF ACClotNT LA.M OR P.-
Three Mile Island Unit No. 2 March 28, 1979 Beginni ;
I,ondenderry Township, Pennsylvania Wednesday at 4:00.:
10.
AMOUNT CC CLAIM tin J.Jl urso C. W8CNG7UL CI A1H l o TCTAL A 8 8 38141*
- aM4",1 E. PleSC*dAL INJVtv 54,010,C30,000 NA NA p4,010,000,000 l
1: C11*K.21. N CF ACCT; TNT IStore behne. in Jesnil. all Lnun n facts and cirrumnum es unending the J.unnyr. injury. or Jrush. identifs.-
- rren
- s s.*J;'re;r ty h:entred and the runse therrufs e
fgh.,
Q See Attachment 1 *.
FECP!ETY CAMAGE
- .Let 4%3 1::stts C0 ownt E. IF C1>t a 1~ AN CLAtwAN1 INamhrt. astere. ray. $onor. und Zit' C" Jet sa::e s4:18.' :ts:s.t! c v: Aho LecAtiCN or raCestiv ANo wa vat AND Ex1ENT OF cAwAct (Xrr insum il..n..a arirrir sidefur med.d nf anha..nd.u..,r ei.-
See Attachment u.
riasCsAt iN;uar statt %A?vst Aas: 1AttNT 08 INJuer wMiCM f Caus THE sAsis,CF Tmis CLAIM Nn n.
wnNissis l
. :: i i s s, s
....,r..,,,c,.. u.,. J u.,,...s z., c..s,,
are h.r.d-A of cerscns, inclnm.baised en the.A' arch 28[C, c., accicient at TMI-2, v= s me to t'.e -.ature of this.clai:9 1979 r
a==. 1cyees c: tne N the clar.;nts E.d c: -.-
- a.rcies,..clM'ha but not limited to ce 3abcock & hilc x Co. anc Toleco 2.cisen C:.
E ' a 4 wi;. esses cossessine reles.nt infer.arien.
these r_ = esses as a resd t ou,alreadv adr sed cf --
The F is its Special An ids.:_-. :: ma.y, if not all, c:
"r.ree Pile Isla-d -- A P.spert to j.he CCmissicners and to the Public" (1980) 7c:7
- ! *
- t - -
-! Av :r. ? cf C;A:: C.'i ! CN.v O Av.A01 AN: :N;Vti!! C AU11; !Y *-! ACC.:!NT ASC'.E AN; AG E! TC A CI:' !.
J Y :. *.*
8... !a*.!F AC73ON AND FIN AL $lilliv.INI CF Ik!$ CLA:?*
- 2.1 * *.a~.:i C E C.A'?.LNT n This.sier:.aure.ehuuld be u. sed ur: ait ja:ure carer <p..ndences 1s. cAit CF CLAsm December S, 1980
,,. - s v-C:\\ !L PEN ALTY FOR PRESENTING CRINIINAL PES ALTY FOR PRESENTISG FR AUDL LEN F R AUDCLENT CL Al%1 CLAl%1 OR \\1 AklSG F ALSE STATE.NIENTS T c :::.-: **cl fe'e:: and pay to :he United States the sum Fine of not rnore :han 510.000 er imrriser. ment for nei -.
l cf !. P. r.* 3:Sc the sr.eunt of dar.2ges sus::ined t y the l than ! 3 ears or t e:n eSe r 62 Sint. cCE. 7JG
$re R.S. U 'G0..u]E:JI L'.S.C.231.s V.:- 'e: h:t =
- 3. ::,.
- e.:.s:+:t-.:.-
Pet L3 s f 8 T 2*. '" C 8.. -
- 1 C' t Ia.2
M FT.I's'ACY ACT NOilCE W
}.
".J
,'4..*
?-;isted ir. ar. rrdance w'i the Pas sry Act !
1.* 5 C 3 /9.c.pf f5,. pac The irfer;. mon re.,e. icd is to N i,wg n,%, ;
f... ac: arr.t a tre i.f:=.:4cn req.es.ec as tr.. *e::tr to s bch ibs.
c ac=s N
e a : se l
C. Ac.nat De See the Nemes of Syuc s of iterceds for the age.n* ::
t
= te. you are s.es:t r; the f:- fer tha aric=a:::n.
A. /.W-. ' r.: re:.es:ed tr.femauch a sc'ieted ;. u.ar.: to eme er. ore of ;
D. E"cer o.'/s.l.., is Aa:W Dxio ure is so!.n: >. Me.es er. fc.re ::
L < fo:.:=y.r ll.5 C. XI. :t U.5,C. M1 er aag.. :! U.S C. :673 er sig :8 e so;:y the ve:unied ir.fermatier. er te esecute the f:rm may re.cer 3:.r C.F.F. laa.
I cla 6:n % ahd.
INSTRUCTIONS Cornpfete oilitems-Inser1 the word NONE where' appHeoble Caisms far da=. age to or for loss or destruction of property, or for personal
- 4) is support of claims for da. cap to prope-ty which has been or ca: be kjery. mas be sig::=d by the owser of the properry dam.sged or lost or the econo:nically repaired. the clairmant should subcut at least two ite=aad m;:=c in.sv. rut perne. If, by reason of death, other disabihty or for re.as.ons deemed r.aae:nenu or esarsaacs by re. liable, duinteressed conce ns, or, if payme1:t has sa:.isfa=ary by t.be Oc er: cent, the foregeing rnquirement cannot be fulfdied.
tv.n made. the iuwd signed receipts evidencing pan::ent.
the ciai= sy be. Sad by a duly authorized agent or other legal representanve, fc/ In support of ciaims for damage to prenny sNcb is not econor.ica. v
- -oMed edience sa.
- :s'a: tory to the Goverr.:nent is s6h=atted wnh said claim reparable, or if the propeny is lost or destit yed. the clairnant shouJd s6t=.:
seraVM.ing actbon y to act ratemenu as to the origsnal cost of the propeny, the date of purchase. asd ce If:'a=as: c.ents to fue clai= for both pene:a3 irjury end preprty car.
vsJue of the progny, both before ard after the acesdent. Such state:nenu s. :..:
c*a;= for bm.h =ia: be aboss is ite:n lo of uis for:n. Separtie c!a;rns for be by dis:n crested compte t pcNna, prefersbly reputable de.a en or eff.: t s p.ru::J: isy. y a.d proprt) damage are not a::r;.abic.
fa:.iliar w,th the type of picper:y da.= aged. or by two or t= ore ce=;s..e Tse a=:e::t c!z::ed should be substa ia: d by com;etent esidetsca as b eders and should be certified as be.=g hst and correct, ft:iew Any funher tas:rvenons or information necessary in the preparst;on of ye.:r
/c I: s: ;;c-t of clai= for pnor.a!i:iury or death. the clairmant shov]d sub: nit cla;m si:1 be ferr.ahed. upon request. by the office indicated in ite:n frI on t:a a = =c: ::N:. b; the a:1endirg phyo ian, shemsg the nature and entent of revene ude.
=.4.>. the :.t:.:.
a:.: c: test of tres: sent the cerres cf pr=a.:::t disabihty, if (d1 Fad.re to cor pletely esecute this fonn or to su;;ty the requested r. sic.a*
a:3. 2.e p :;. s L ar.d the p od of hespt:.al.:ator. or inca;actation. ara:h.tng mit':.i.n two years from the date the allept>ons accrued may render your cf =
P.e. ad bil. fe; cedgaf hospital, er buria! expe scs a::sa.Uy incurred.
anvalid".
INSURANCE COVERAGE in order that Werogation clairns may be adjudicated. it is essential that the claimant provide the fo!!owing information regarding the insura..:
cosersge of his schicle or property.
- 17. DC YOJ CA84Y ACCIDENT INSURANCE? 2 YE1. IF Y15. Giv! NAME ANO A002E15 OF INSURANCE COMPANY (Nassrtber.Jfrret, city,flate,grg Zip C. d.
A.o P0aCY Nur..stR. C NO (1).', :EriCan NuCla='- I...surars, 270 Faming:Cn Ave., Fa cington Ct. Policy No. 1353 (2)
.%...cr Insur. Co. Icag G.reve, Illinois Policy No. TA1084 f
- 19. IF DEDUCTIBLE. STATE AMOUNT
- i.G C h ita.,I CE C10VCii5ti?
.-S.L - S100,000.00 Ds:ductible Kecpar - S100,000.00
- 23. tF CLAM HA51!!N FitID wi1H YOUR CARRIER. WHAT ACitON HA5 YOUR IN5UEEE TAKEN OR F40PC115 TO TAKE WiiH E!FERENCE 10 v0.2 CLA;M? tl1 as nece.tsury that you ust errain thesefacts)
Fa.- e.ts to Cate (as cd 12/2/80) yst ni S96,256,228.00 JCHL S 4 S,12S,114. 01 Penelec $48,128,113.99-r :~ :.
- t*
3:/C taintiv. F:0*!stiv Care.'GE IN5Ut a. !? Q "!1. d TE5. Oxi ' Art.! AND A : !!1 OF N5 p a.c! :::
J 31: Mour. street. s ity. State. aeJ Zip C odc e O NO PrC?ar y C.5-age - as a cVe No. 20 d.*:
-:= 41 V T-e -=,CE.
, '.. '. -" S i A CNC E.Er f I.iEhilitf I.Dds:citsrs, 919 N. FdCh,gan Ave., ChiCEgo, I".
'ee-CC. 270 TE.inO Cn AVE., Ear ~.ing CO, CT E.. _ r e st4.1.:uy
. cue.
...u
-; 90. ten C ;e r e4 titi st.=:at: *Cas es s.cx as
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1 Ca::cw s W,
. General Public Utilities Corperation, p;o dJersey Central Power & Light Ccenaay, D..
(Metropolitan Idisen Cc:pany and '
' 1 8 --.. ' Y.
- Pen.csylvania Electric Cc=pany CLA.IM 1
Docket No.
V.
1 5
.i Nuclear Reg.alatory Cc:=ission.
i i
' ---------------------------------------x General Public Utilities Ccrporation (",GP U" ) and its eperating subsidiaries, Jersey Central Pcwer & 'ight'Cc pany
(*JCP&L"), Metrcyclitan Idisen Cc pany
(" Met-Id") and Pennsyl-vania Ilectric Cc pa.y ("Penelec"), bring this clais against the
- Nuclear Regulatery Cen=ission ("NRC") alleging as follows
j 1
1.
This is a claim arising cut of the March 28, 1979 J
accident at the nuclear electric generating facility known as Three Mile Island Unit No. 2
(" xI-2").
Claimants seek damages against the NRC under the Federal Tert Clains Act, 28 U.S.C.
iS 2671 et see.
-~
Jurisdictica 2.
Jurisdictica is based en the Federal Tort Claims
!Act, 28 U.S.C.
S 13 4 6 (b), and regulation 10 C.F.R. $ 14.1 (1975).
i
- t.This is a clain against the United States fer =cney da= aces for injury to and Icss of property caused by the necligent and wrcngful acts and c=issiens of esployees of the NRC while acting within the secpe of their e.plcy:ent.
This claim is filed befere the NRC fer dispcsitien in accordance with the pr:visiens of IE U.S.C. ! 2675 and 10 C.T.R.
5 14.1 _e t see.
The 7arties 3.
Claisi..: G;U is inecrpcrated i. Pennsylvania a.d 37 6 s
N rn t o r /-D W u-
h has its principal place of business in New Jersey. G7U is an l
l
, invester-evned public utility holding ecapany, operating pur-f
- scant to the Public Utility Holding Cc pany Act of 1935, 15 U.S.C.,
5 79 et s'ee., and owning all of the co=non stock of three
- j operating electric co=pany subsidiaries, claimants,JCPsL, 4 Met-Ed and Penelee. As used herein, GPU refers to GPU and all 1
i
- of its operating subsidiaries.
4.
Claksant JC?sL is incorporated in New Jersey and
'has its principal place of business in New Jersey. JCPsL sells electrical energy to retail custemers in nc:th-ce'ntral, east-central, northwestern and western New Jersey and to cther electric cespanies and entities fer resale.
5.
Clai= ant Met-Id is incorporated in Pennsylvania I
4and has its principal place of business in Pennsylvania. Met-Id l e
selle electrical energy to retail customers in aastern, east-central and southeastern Pe..nsylvania and to other electric ce=panies and entities fer resale.
6.,
Clattant Panelec is inccrporated in Pennsylvania and has its principal place of business in Pennsylvania.
7enelee sells electrical energy to retail custeners located in
. vestern, northern and south-central Pennsylvania and to other electric cespanies and entities for resale.
7.
JCPsL, Met-Id and Penelee are co-owners of the nuclear electric generating facility known as Three Mile Island Unit No. 2 (":M:- " ), which is lecated in Lcndenderrv 7:vnship, Pennsylvania. Met-Id owns an undivided 50% interest in TMI-2, and JCP&L and Penelee each own an unditided 25% interest in IMI-2.
Me:-Id is the cperater of 2M0-2.
The Atemic Inergy
- =nissienissuedan}pers:ingLi:ense,27R-72, tome:-Idf M:-2 on 7ehrtary 5, 1975. The n;: lear s:sa: supply system u
in 7:::-2, including the nuclear rese :: and substantially.all cf
~
2 O
9
%l 4
the engineered safety systems that ecntrel the nuclear reacter, were supplied by The labecek & Wilccx Ccmpany ("B&W").
t 8.
All of the =ajer electric generation, transmission ;
and distsibution facilities of the claimants are physically 4
e
- interconnected. The operations of these electric facilities are j centrally coordinated within GPh to function as a ' single, j
l l
- integrated electric utility system known as the CPU System. The i i
energy generated by TMI-2, when operating, is ce= ineled with i
the energy generated throughcut the GPU System and is t: Ens =itted}.
[
throughcut the GPU Systa= and distributed to retail-customers er !
i scid to other electric cc ;anies Lnd entities for resale.
9.
The NRC is a federal executive agency, established i
j hy the Inergy Rec:gani:atien Act of 1974, PL 93-438, 68 Stat.
t 1233, 42 U.S.C.
S 5214 et see., as a successor agency to the l
i.
Atomic Inergy Cc =ission.
As used herein, NRC refers to the present agency, its predecessor, the Atc=ic Inergy Cc==ission, and all present and fermer divisiens, effices, employees and agents of the NRC.
By statute, the NRC is charged with the l
1 establishment of " standards and instructicns to gevern the pos-
\\
I-sessien and use cf spe:ial nuclear :aterial, source caterial,
' and b.ve.roduct =aterial as the Cc==ission =ay deem necessarv er i
desirable to prceste the ec==ca defense and security or to p ctect he lth or to =inimize danger to life or property." 42 L U.S.C. 5 2201(b).
10.
The
- RC has the authcrity and duty to regulate
/ 4" '
the desien and coeration of ec=mercial nuclear oower plants O'
/
}-
within the L*nited States. In the 1974 Inergy Reorgani:atica Act,
. u],
yf l
('/ ',,,
s: ra, C:ngress autheri:ed the *:RC te " pres:rihe such regulatiens gyn l A,4,
?
.:: ceders as it may det: ne:essary.
- ccrern any acti.ity
...h,,,
j"' f auth: i:ed.:::suant :: :his chapter, including standards and
- estricti:ns gcVer.ning :he design,1ccatien, and eperatien cf 3
I 1
e
O O_..
,i, O
V facilities us'ed in the ccnduct cf such activity.
42 U.S.C..5 2201(i). The NRC has prenulgated regulatiens setting.
I a
- forth mandatcry agency operating procedures which are set forth a
s in the Code of Federal Regulations, the NRC Regulaton Guide, the
- t
$,NRC Manual, the office of Inspection and Enforcement Manual, the g
]StandardReviewPlanandotherguides,nanualsandpublications j
I-(
. which, in relevant part, are described below.
11.
In the operational exercise of its statutory and a
-7 regulatory duties, the NRC g2duced GPU and Met-Ed to rely and GPU and Met-Ed did rely upon the NRC to warn cf defects in equi _pment, analyses, procedures and trai-ing affecting the operatica of TMI-2 of which the NRC was or should have been aware. The NRC, in
- ne cperatienal exercise of its statutory and regulatory duties,
&~
induced) GPU and Met-Id to rely and CPU and Met-Ed did rely upon
- ss-the NRC to review with due care the equipment, analyses, pro-cedures.and training fer nuclear plant cperation sub=itted to the SRC by nuclear equipme.t venders and nuclear plant licensees.
T)e March 28, 1979 Accident at TMI-2 4
12.
Cn March 28, 1979, beginning at 4:00 A.M., while
[
CMI-2 was cperating at about 975 cf full pcwer, the turbine generator shut down er " tripped" due to sudden loss of feed-
.y
- water./ Under NRC regulations, such an unscheduled turbine g,
generator trip is an "a.nticipated cperational occurrence" which u
,,,')3 '...
is reruired to be pla.ned fer in the design of a nuclear plant.
As with any such shutdce., the re..cval cf heat f cs the pri.ary leep by the secendar.y loop was reduced substantially. Within seconds,.the centinuing buildup of heat in the primary loop raised the pressure in the reart:r coelant syste=.
- . turn, this caused a re'ief valve en tha 3:essuri:er (the "pilet-c;erated relief valve") to cpe., as it was desig.ed to de, in 4
e t
~' - Q -
Q l
I c:dar te relieve the ex:ers pressure.
13.
Several secends after the pilet-eperated relief
) valve had cpened, the reacter shut down or "scra==ed," causing
)
(the pressure in the reacter coolant systed to drop to within a
sf its nor=al range. At that point, the pilot-operated relief
-l-j ivalve should have_c_losed.
i
, u 14.
In fact, the pilot-operated relief valve ir. prep-4 t
t
{}
erly failed to close_and, becausecfalackofinstrumentationtol ot*
jv af"'
- ' indicate clearly either the cpen pcsitien cf the valve or the r
8
/
' existence cf flow thrcugh the valve, the c erators at 7.v.I-2
. fi 4
t s-
-l J
' vere unaware that the valve had failed to close. Thereafter,
(,
', 7 ( jf significa..t quantities =f ecciant water and steam escaped through 794 p
the stuck-cpen valve, and a "1 css-cf-ce lant accident" was in Y (. A/~
3 s
,-;p cgress.
Ob
} (v
- 3 15.
As =cre ecelant water and steam escaped, the J
ipressure in the reacter ecciant syste= continued to d:cp.
Within app:cximately two minutes, the.:: essure fell te a level a
8
.at which an en3ineered safety system be an t.rovidine. hic.h-i f
_ pressure injection of water into the reacter coolant system to
. replace the lest ecciant and ensure that the nuclear core was
-icevered and p ctected by coclant.
s 16.
App:cxi=ately five =inutes af ter the 4 :00 A.M.
4-
.tr
. [ j/s#*"
turbine generatcr trip, the CM -2 cperaters substantially reduced he
'the high-pressure injectica cf replacement coolant into the re-0V ct *
~->
. p, aeter, in at:::da..ce with IEW-supplied IL=its and precautions, 4
(,.
~ rocedures and training, which the NF.C had reviewed as described t
{)p.
herein.
17.
As a resul ef the Icss cf coolant thrcugh the s:::k-: pen pil:t-yerated relief val.e and the lack cf replace-the..u
- ear fuel c::e everheated, severely
- e.: ::cla.:,
[,l. 6 f
~},
f damaging the p:c e -ive cladding en the nuclear fuel and schstan-
[
b"
-~
- W
- bl1, c-yl..,
CO L
I p (V t
.l
..h.
G t
tially destrcying pertiens ef the nuclear fuel core.
Radicactive
. =aterial from the ruptured nuclear fuel ecre spread thrcughcut i
- the surrounding reacter equipnent, further darsaging and contamin- !
g I ating large portiens of the nuclear stea:n supply system and other j g
}
.t. equip:nent and structures in the TMI-2 centainment building and i
i.' in the adjacent fuel auxiliary and intermediate buildings.
1
/
W THE NRC'S NIGI.IGINT PI?.TO?_'G_NCI A.ND C".!SSICNS CF ITS CPI?'sTION;* TC;CTIONS i
18.
Prior to the March 28, 1979 accident at TMI-2,
~
the NRC bcth had reasen.to know and actually knew'that there
~
/.
W were defects in the equipment, analyses, procedures and training 7
~ _..
supplied by Esw for TMI-2.
Netwithstanding the statutcry and (t. ',~
ree-claterv duties of the NRC to warn nuclear plant licensees of J
.such defects, and notwithstanding the reliance by GPU and Met-Ed ;
en the NRC fer the dissemination of such warnings, the NRC~~
~~~~
s nee.lic.entiv f ailed to warn GPU or Met-Id of such defects in TMI-2.
That f ailure to warn by the NRC was a proximate cause
,of the March,28, 197 9 accident.
l(
19.
Fursuant to NRC regulaticns, the NRC Cf fice of 1w
(
,[~
- spectica and Inforcenent is specifically required to inspect e
t
[
nuclear plant licensees to " ascertain the status of ccepliance
\\
g s-5P
- *s
,$+ -
with NRC requirements including rules, regulations, orders and
'v.
'g ts#
A. d
,5 license provisiens," and to "!ilnvestigate incidents, accidents,
\\
</
a~
/
f.gG,i allegaticns, and cther unusual circu= stances involving satters in the nuclear industry which say he schiact to NRC jurisdic-
/-
+. h r tien to ascertain the facts and to take er recommend appropriate s
o s
, [ f.
xg y
y
- 1, /
acticns."
NRC Manual Ch. 0127 (1973).
q E
/8 20.
the ::RC dissemi-
"_he NRC regulatiens ma.date that l%
y [d t '.
..a e a=cng licensees cf. :* ear pcwer p' a.ts infer:atica derived Il
()$
free cperatin; e::;er:.e.ce at all.. clear pla.ts in the.T..itad y3 p/
S u tes, 1=clufin-ea u en cc=pt...t fan:.s e=d yrccee re I
6
y g
4
- hanges. 10 C.T.R. S 1.64 (1977); ?:RC ::anual Ch. 0127 (1972).
The NRC requires licensees to rep::t unscheduled incidents er events which involve variaticas frc= regulaticas, technical specificatiens or license conditions.
10'C.T.R. $ 21 (1977); NRC I
(Regulatory Guide 1.16 (1975). The reports, called Licensee Ivent.
?
s Reports, are sub=itted to the NRC of fice of Inspection and-Inforcement to aid the NRC in obtaining corrective actica at the reperting plant and in preventing a si=ilar occurrence at other auclear plants. The directer of each division within the office c' Inspecticn and Infc ce=ent is required to "le] valuate licensee event repcrts and Regional reperts to identify generic p chle=s and to deter =ine the significance of indicidual incidents.
NRC Manual Ch. 0127 (1978).
21.
The NRC office of Inspection and Inforce=ent is respensible for evaluating licensee and NRC respcases to inci-r, dents or accidents to " assure adequacy of the overall response
(
- o the incident er accident."
The Regicnal Oirector must review significant, events, allegations and investigatory findings r
for =atters having generic applicability. Regienal Inspecticn and Infer:ezent Oirect :s are required to re.iew all reports
=andated by NRC regulations, including all Licensee Ivent i
Reports. NRC Manual Ch. 0127 (1975);
Inspection and Inforcement.
Manual Ch. 1110-051 (1978).
22.
As the primary recipient cf plant operating data,
't the NRC Cffice ef.:nspe :icn and Inf:::ement is required by
)
regulaticas to analyze and disse =inate i=pertant safety informa-
- ica to cther NRC offices and to all nuclear plant licensees.
h*
- 7.C Manual Ch. 3127 (1975). Cne of the pri..cipal nethefs used o.
by :he ::KC Cffi:e cf ::spe::i:n and Inf:::s int f: adtising licensees cf irper-an; safety =atters is th:: ugh the issuance f 3:11e: ins sad Circulars. Ee:ause cf the i pcrtance cf these I
7 I
i
~
i I
t b_
Q G
.ctices fer varning l'icensees of p:ssihle defects a.d safety problems, NRC offices cther than Inspe=ticn and Inforcenent, t
- such'as Nuclear Reactor Regulation and Nuclear Materials Safety
'and Safeguards, also recc=end the issuance of Bulletins and s
- C1.rculars on particular subjects. Inspection and Inforce nent i
l 3
j
[ManualCh. 1125-052 (1978). NRC regulations direct all NRC j
' staff to be alert to a.ny infor=ation which has potential safety i
f significance. Che regulatiens require every ne.ber of the NRC staff "to be alert to the energence of infermation -- f cm cutside sturces er within the staff -- which is n'ew, potentially inportant, and potentially relevant te one er scre pending i
proceedings." Inspection and Infc:cenent Manual Ch. 1530 (1978).
.c a
23.
N_RC regulations i=pese_a.du_tv. en the NRC Office
- of Inspection and InforcemenA_to_issu_e Eulletins regarding l
~
i
'. matters of " safety, safeguards and envire = ental significance" for nuclear plants and to require that licensees take specific
~
a:ticas as a result of safety-related design inadequacies, equipnent defects, operating inadequacies, =alfunctions, n.er a5y q
other failu.res of a generic.qature that have occurrec',._Af,._a_simi-lar facili y er cperatic..__ A Bulletin requires licensees to inspect for and correct the inadequacies described in the sulletin. Che Inspection and Inforcenent Manual requires the issuance of sulletins when an event c: condit' ion is generic s
f and i=pertant to safety. Inspection and Infercerent Manual Ch. 1125-031, 11:5-041 (197E).
W 24.
G7U and Met-Id relied en the NRC to conply with O
the cen;;ehe.sive requirenents of data cellection, analysis and aL*
,/
,l#
/,'
['
disseninatien' set forth in statutas a.nd regulati:ns. G70 and z.
,v*
\\.p y-t Me -Id eliai :n the "RC :: ir sue warnin;s as,reqL red 'rv_j:pC i
i
.u
/. '...
regulatic.s. Met-Id.aintai..ed a f::
1'.
.:::;:a= f:: -he re ie..
~
a
/ f
.r*[]
Of C:=unicatiE".s IC" the CIfbee O ns[e Iben and In[:! e".eSt f
e
.a te de e:-ine whether any adverse : nditien repe ted by the NRC
.}.T V'
i l
I i
%,)
Y required ccrrective a:ti:n at 7::I-2.
Met-Id, the'cperater cf j
.':I-2, pre:;tly disserinated infer atien f := NRC Eulletins
.j vithin y.et-Id and required pro =pt replies and app:cpriate
[
- action.
7*I NRC's NEGLIGINT FAILURE 8
d 2
70 GIVE WARNING 3ASED ON l;
't THE OAVIS-EISSE INCIDENT l
i 25.
In Septe=ber 1977, a 1 css-of-coolant accident
/.t/ r I
b p
occurred at the B&W-sue.s_ lied Davis-3 esse.__I nuclear pewer lant e
l Af '
[v4 ~
'j of the Tcledo Idison Co:pany.
That accident closely paralleled f
the events which o: curred 1B.menths later at T.MI-2.
26.
Following the Septe=ber 1577 incident at Oavis-(
[kp ()
g S e s s.e, the *;f.C_.ge_ licentiv failed to perfern its dutv (a) to e
/
1 adequatel.v to investigate a.-d ascertain the facts, (b)
- t. o__tak e t
r/',
h
. and recc= end app c..pri.a_te actio.n. and (c) to warn Met-Id and i
6 s
I
,/f other licensees of E&W-supplied nuclear plants of defects in T
equip =ent, a.alyses, procedures and traininp which the NRC had 9
/
/'
C' discovered or should have discovered as a result.cf the 0 avis-Sesse incident. These neglicent failures contravened NRC fij
~.
/
duties i: sed bv sta:: e and rec.ulatic s and were inconsistent I
b q'J ' g '-
with duties previcusly undertaken by the ';IO.
The NRC thus
/
(J negligently performed and negligently c=itted t'o perform opera-
~
tional functicns mandated by statute, NRC reeulafions and past agency practice.
I' a proper warning had been given by the NRC, the TMI-2 accident en March 23, 1979 M have bee'[ avoided.
27.
C: Sep:Enher 24, 1977, while the Davis-3 esse plant was operating at 91 cf full power, a sudden Icss of b#
j{ /
g-feedwater caused a turbine generater trip. When the pilot-j\\/
.A
-(,
cperated relief valve subseque-tly =pened and failed to close, g
/ \\
- he avis-Iesse pla..: e :perienced a less-f- cclant accident.
dr ",
< (/
As rea:::: =c lan press;:e d::;;ei, tha high-pressure i..je::ica cf replactre..: coelant activated automatically. The Davis-lesse 9
e 9
[]
V
- u. ",
6-l $r y
M" L!
- l. 2 nuclear stea: supply systa= desig. did not have a direct 1.di-h@
- cator of whether the pilot-eperated relief valve was cpen c:
t
\\..,
closed or whether there was a flow of coolant through the l
' relief valve. Just as happened later at 7.'C-2, the water level i
j in the pfessurnTe - -r began to rise,~ misleading the operate-= hto C
{1 lconcludingthat there was no less-of-coolant accident in pro-J b
gress. Acting pursuant to NRC-reviewed li=its and precautions,
),
procedures and training, the ope;aters at Davis-Besse then d
ft.
terminated the high-pressure injection of replacenent coolant 3
into the reactor coolant syste=.
l
"?_ g 28.
I nediately following the Septa =ber 1977 Davis-1 2 esse incident, the NE: hegan an investigation, as required by 10 C.T.R.
S 1.64 (1977), which included cperater interviews and J V
.:eviews of p2 ant cperating data, equipment and cperator action.
/
.f /
29.
The NRC conducted a..other investigatica 'of the h
[
september 1977 Davis-5 esse incident during 1978, which resulted O
h' in the NRC i=p*.amenting a revised Operating Procedure for Davis-g j)
,'3 esse, described more fully below at paragraph 34.
~
.J
. {d k
n 30.
As a resul of the investigatics and analyses of facts, which regulatic..s re uired the NRC te perfor= fcilowing the Davis-3 esse incident, the NRC knew or shculd have known the
[!
,s following:
r
' a.,
(a)
There were defects in equipment application t
r i
g
[ Y
.' and instru=entation of the Davis-3 esse plant, including exces-sive reliance en the I;W-supplied pile -cperated relief valve te
, f-
/
cpen and close and a lack of instru=e.tatien to indicate the n
/,'
valve position:
g
~
(l (b)
There vere defects in the transie..: analyses t
previcusly su;; ied by 3C and reviewed by the 37C, i.c*.udi.; a 7.,
/' h
failure te analy:e adspately p:te.tia* hrea%s 1. the cccla.:
/
syste= as small as a s:::k-:,:en pilet-cperated relief valve and 1C 9
4 s
I
w/
V I
I e failure te analy:e adequa:ely pcte. tial breaks 1ccated a: the l
t s
.tcp of the pressurizer:
=
4 j
(c) There were defects in the li=its and pre-i 3
l
, cautions, procedures and training reviewed,by the NRC which I
'i=preperly directed plant cperators not to permit the pressud -
l
~(
Izer to becese filled with water or "go solid;"
I (d) There were defects in the operating and le=ergencyprocedures,includingprocedureswhichi= properly i
l i
'per=itted pre =ature termination of high-pressure injection i
1 l
'.before the cperators had identified and arrested a less-of-2 t
I lccclant accident; le) t*nanticipated bciling cf the water in the
.reacter cecla.t syste5. at ravis-3 esse had caused a rise in
. ',pressurirer water level which misled plant cperators into con-
) luding that there was no loss of water frc= the reactor coolant c
t 1
systes.
J e
I 31.
Each of the defects and cperating prcble=s set'
.A ferth in paragraph 30 were generic p chle=s affecting TMI-2 4
t f
aand other B&'f-supplied nuclear plants because those plants cen-g 3
y
. 'tained similar equipment and instru=entation and relied upon i
J
, lsi=ilar precedures and analyses. NRC regulaticas required the f
i i
A y
ICc= mission to "[nctify] licensees regarding generic problems
.f M
i r
/
't jso as to achieve app cpriate precautienary er corrective action.".i t
s 1
('e J(' {"s i
10 C.T.R.
S 1.64 (1977). Nevertheless, the NRC negligently t
i
.h
(,/
failed to netify licensees, including Met-Ed, of these " generic i
s N
K}'
d,'.
,proble=s," which it knew about er should have knew. ahcut as a V..
- result of the Davis-3 esse incident. That failure was a oreximate n
cause of the accide.t at 0::-2 en :' arch 28, 1979.
\\}
w 32.
The EC negligently disserinated :: nuclear plan:
\\[yy
(.g.
V j.
,Yk
- ice. sees, including Me:-Id, su-aries cf *. ice.see Ivan. Repc :s regardi..g.he : avis-3 esse i.. cide..
which failed : varn -hat the r
0,
(
/ /
11 W
/>
m -
1
o o
operaters at : avis-3 esse had prematurely terni..ated high-pres sure injection befere de termini.n_g whet _h_er_ _a les..s_-E._. c_oolant
.9. accident was in progress. Toledo Idison Cc:p., " Licensee Ivent N7-32-77-16," Decket 50-346, october 1977.
. Report:
t J3.
The NRC negligently disscuinated to nuclear plant i 3
t l
- licensees, including Met-Id, a sc=sary of an erroneous supple-r re 8
j/
5 r',j jnental Licensee Ivent Report on the Davis-sesse incident._ That
,j d'..tl,) '.
Ireport errcneously concluded that "[olperator action was timely i
/
,/
and p cper throughout the sequence of events."
As a result, the
- g
- - --the Davis-3 esse op6:ator action NRC failed to Carn Met-Id that
~ --
.J 4
had aggravated the less-of-ccciant accident by terminating high-pressure injectica of coolant. Toledo Idisen Ccrp., " Licensee
,Ivent Report: NP-32-77-16 Sup.:lement," Locket 50-3463 November 51977.
9 i
34.
Mere than a year after the Davis-Besse incident, the NRC implemented new cperating procedfres for Oavis 3 esse.
. ~..
p to prevent a recurrence of the Septa =ber_19,77 accident. These procedures stated:
"NdYI: Pricr to securing EPI lhigh-pressure injectien), insure that a leak does not exist in the pressurizer such as a safety valve er an elect:c agnetic
[i.e., pilot-operated) relief valve stuck cpan.
A mini =um g p.
l1
.] '
decay heat flow of 2800 gym is required l
pric: to securing high-pressure injection.
",}
If the leak has been isolated, the high-i pressure injection pump can be shut devn e'
typ g'y after FCS !:eacter coolant system) pres sure I
incretses abeve the shutoff head of the pump. "
Davis-5 esse No. 1, I=e:gency. Procedure I? 12 02. 06, "Lcss cf 1
Reacter Coclant and Reacter Ccclant Pressure." The NRC neeli-
' G, ' f-eentiv f ailed tc direct the implenentatien cf this new c:* erating v
W, Y JP
~
p: :elure by licensees cf cthe: EsW-supplici nuclea: plants, l
(y #
g a
w lJ
\\"
in:Luding Mat-ti.
i.-
J t'
4^
f($
25.
n additien releasing 1-:: plete, er:enecus Ed
.,s.s I
,f a..d misleadi..g Licensee Itent Fep::ts :egarding the 1 css-cf-p,-
3 e1. '
C
/ L F' ig JY v
12 i
e e
O
~ ~~
U
~
,j c clant accident at Oavis-Hesse, the NRC Cffice of Inspecticn and Infor:anen: negligently failed to issue anv Bulletin or
- Circular warning licensees of 3&W-supplied nuclear plants, I
including Met-Ed, of the deficiencies in 'the equipnent, analyses, t.
t t
,. procedures and ::aining which the NRC had discovered or should l
t e
i
.4
! have discovered as a result of the september 1977 Davis-Pesse I
t.
I incident. As a result of its investigations of the Davis-Besse s
- incident, the NRC knew that the e uipment and_,Ipeetaral defi-
-~
ciencies were generic to E&W-supplied plants and that the sub-s_
stituted operating procedure was L:portant to the safe operatica
,i cf the plant in that it instructed operaters to take steps 1
g which would avoid cere une:very. Thus, the NRC knew that a L-
, Bulletin was mandated by NRC regulaticas, see paragraphs 22-23, e
sucra. Nevertheless, the NRC_peeligen.j@y_f alled_to_ issue a i
Eulletin.
36.
Sectica 208 of the Iner y Rec:gani:ation Act of
[
1974, as amended, requires the NRC to determine which incidents and events represent _Ahacrual occc:rences and to repcrt those r.
j Abner =al Occurrences to Congress. The NRC must disseminate inic::ation relating to an Ahn===al C::urrence to the p$blie
. ithin 15 days af ter the NRC has learned of its occurrence.
w
,Inspectica and Inforcement Manual Ch. 1110.
"Abnc mal occu:-
rences" include
" Design or safety Analysis Deficiency, Persennel Irror, or Procedural c: Adninistra-ive Inadequacy:
- 1. Disecvery of a na c; c nditien no: specifi-l cally censidered in the safety Analysis Report (SAR) c: te:hnical.specifica:icns that require i=nediate remedial action.
- 2. Persennel er:c: c p:c:edural deficiencies which resul: in 1:ss cf plant capability
- perfern essential ssfety functi ns such that a p :entia* release cf radicactivity in excess f 10 CI? Par: ICO puidelines ceuld result frem a p:stulated ::ansient
== a::ifent (e.g.
less cf energency c::e ::: ling sys an, 1:ss cf c:::::1 ::d systen)."
13 4
9
.a
Q V
s
- spe= tics a.d Inic:cenent ::a.tal Ch. 1110, Appendix A.
The SRC knew that the procedure which nisled cperacers at :: avis-5 esse
,pra.aturely to tersinate high-pressure injection was a "proced-I
, ural deficiene[y)," as defined by the Manual.
h 37.
In violation of XRC regulations (Inspection and
~~
~
J
- Inforce:nent Manual Ch. 1110, Appendix A), the NRC negligently j
~ failed te classify the Septa.ber 1977 Davis-5 esse incident as an
?
l h)
Abnor=al occc::ence in its subsequent quarterly or annual report e
Q f
g to Congress, thereby failing to warn licensees of other EsW-J/
supplied nuclear power plants, including Met-EE,' of the defects
($ ' ' /]
f, and cperating proble.as revealed by this event which required
/
/
p in::ediate remedial actica at sinilar plants, such as CMI-2.
b 38.
In addition to tl a f ailure of the NRC to we.rn d'
[
Y
' licensees of B&W-supplied nuclear plants of defects and p chless,
/
of wnich the NRC was aware as a result of the Davis 3 esse incid-
/
ent, the NRC negligently f ailed to act in ether ways to inves-t
/
tigate, discover and warn licensees of defects of which the NRC should have been aware as a result of the Davis-Besse incident.
9 Chese negligent failures by the NRC have been docunented by the URC in a fcur vclune report which the NRC approved, published
, and released in January 1980 to the public, entitled "Chree Mile I
- Island -- A Report to the Cennissioners and to the Public"
, o-
.I r"
(hereinaf ter "Scecial Incuiry"). Che NRC, in its Scecial Incui-v, y
(/\\.,p.).j, J p s that r
(a) NRO staff persennel inec :ectly advised the Ib, V /%
v*
8-14' NRC Adviscry Consittee on Reactor Safety (ACRS) that the cense-1,. i y
}.
quences of a 1 css-ef-ecolant accife.t, such as.had cecurred at
,-[
2 avis-3 esse, did net need :: se examinad f:: a reacter cperat'n; 5'
s as C:::-l w:21d te :.. Mar:h 2*,
1979 -- be:ause at full p:ws:
- f the le ;::hatility cf such a. e.a..: ec::: ing at full pcwer.
fs e:ial n-ciry, Vel. II, Part 1 at 15 0 u
9 a
I,
O O
[~
4 l
(b) While the NRC reccy.i:ed that it sh:uld ex-l amine the basis fer the de:isien by the c.:eraters at Oavis-
?) 3 esse to termi.. ate high-pressure injectica, the NRC failed to I
? direct its inspectors to resolve this issue.
(Special Ineairy, 1
3 4
Vol. II, Part 1 at 152)
I l-39.
Even though the NRC knew cr should have known l that there was an unreasc.iably high rate of failure of pilot-v.
t t
' eperated relief valves supplied by various manufacturers fo i
nuclear plants, the NRO erreneously cencluded that the failure
~
ef the pile:-cperated relief valve at avis-3 esse in Septenbar 1977 had nc saf ety i=plications for other..uclear plants cen-
/
g F*
taining pilet-cperated relief valves designed by different t
=anufacturers. Another failure of the pilet-cperated relief I
valve occurred at Davis-Besse in Octcher 1577, while the NRC
+
was investigating the September 24 incident. The NRC, in its
')
./
/
./ ' Special Inceirv, attits that the ::RC k.ew tf.at "similar pieces
/
e cf equip =ent with ec= parable prehabilities. of failure and simila:
w failure scdes were installed en c her 3rn' plants and, in sone
\\
9 cases en all pressuri:ed water reacters."
(Special Incuiry,
.. /
,s
..,(
vol. 2, Part 1 at 136)
[
v.
40.
The final report of the NRC Inspection and Infor: ament inspecters in Regica III, where : avis-3 esse is located, failed to identify the generic implicaticas of the
- avis-3 esse incident, including the misleading rise in pressur-
)
i:e: wa e: 'evel, the i..::: ect c.: erat:: resp:nse :: pressuri:e:
J (l
level and the misleading limits and precautiens, procedures and
- aini..;, reviewed by the ::RC, whi:h had directed that er:enecus
- pert::: :ss;:nsa.
41.
7:i:: t: the septerbar..* 7 Orvis-!*sse incife..:,
-he ::R: knew :: with fue care sh :*d have kn:wn f c= c:he:
re;;;;s whien it had received that its previcus evaluatiens ef 15 9
9 I
.I
(,
j p) 1&W eruipment, a..aly s e s, pr::edures and ::aini..g ve:e inadequate.
i gje
.As the NRC special Incuiry admits, the NEC cnitted to heed these
/
(jf early "precursers," just as the Censission later failed to I
9.'
/ 'f/
- , respend with due care to the Davis-3 esse ' incident, as described 3K l
6
( at paragraphs 25-4 0, suora. These earlier precursors included:
I
{/
(a). In 1971, the Atomic Energy Commission, the l,
.prede: esser agency to the NRC, was specifically advised that a snell-break less-cf-ecelant accident at the top of a pressurizer as was to eccur at TMI-2 en March 25, 1979 -- ceuld create
-j nisleading signals, thereby interfering with high'-pressure l
injectica of c:clant. Alth ugh the' NEO was thus en notice t
j that it should analy:e =isleading signals cf water level'eaused by such an accident, the NRC negligently f ailed to perform that analysis or require suppliers of nuclear equipment, uch as 5&W, to perfer= that analysis.
(special Incuiry, vol. II, Part 1 at 139-40)
(b)
In 1975, the SEC c:npleted a c: prehensive I
report en nuclear reacter safety, "Che Reactor safety Study (WASE-14 00)," which c:ncluded that small-break Icss-ef-coclant a::idents -- se h as the f ailure cf a,:i* ct-cperated relief valve to close -- were among the highes: p cbability risks in
'a Aucletr plant.
(See ial In uiry, vcl.
II,~ 7 art 1 at 142)
- Yet, w
the NRC f ailed to analy:e or require nuclet: equipment suppliers, such as 3&W, to p;cvide adequate analyses of r=all breake -
m (c)
In 1977, the NEC su':stantially ignered a report prepared by Carlyle Michelsen, a censultant to its Advis-cry C:n=ittee en Feae::: Safeguards, which put the NRC en notice tha: neither the anal *-traak analyses supplied by nuclear e uip-nent supp*iers.:: the :::pute: :dels then used te preft:-
rea::::-:::* an -systa-isha ti:: were valid fe: analy:ing reall-h:eah icss-f-c: clan: 1::idents.
(E e:ial Inruiry, vel. II,
. art at 144-46) em e
9 I
O O
l i
- : e,-. 5... C,.. _ _.. ~.
-.us FIRTC?..(';CI A**D CM:5 5!O'*S OT ITS CPIR.*sTIOSA's Tt';CT*C::S I
f LQ _
I
_J e
42.
GPU and. Met-Id relied on the NRC to issue warnings.
/
Ct L
of defects in equipment, analyses, procedures and training in l
~
J:d MA'
! accordance with the NRC's statutory and regulatory duties.
_e4 43.
If the NRC had exercised due care in investigating !
.A i
{
. ~ ~
the Davis-Besse incident and analyzing other precursors, and if
/
the NRC had issued correct warnings of generic problems in S&W equipment, analyses, p ccedures and training, GPU and Met-Id
,! M would have had the equipment, instrumentation, piecedures and j
G, M
- aining reas nably needed :o avcid the accident en March 28, t
'/
1979.
44.
The negligent failure by the NRC to issue Bullet-
,,.^^--
ins, Abnormal occurrence Reperts and other warnings required by p
statute and NRC regulations was a p cxi= ate cause of the March 25, 1979 acciden: at T:1I-2.
NRC' S NIO'IGINT !! F*.I?*.INTATICN CT PL'IIW RICUIRI 'II TS cJ f,,,f:r--
k 45.
A p cximate cause cf the March 23, 1979 accident
(~' -
at TMI-2 was the f ailure cf the NRC to review with due care, c:
W gj -
I, c.6a.
in accordance with statutes and regulations, the equit.-.ent, analyses, p::cedures a.d training supplied by 3&W fc THI-2.
46.
Pursuant to statutcry and regulatcry authority, 1, M* '
the NRC issues licenses for the censtructica and cperation of
,,b s each c==ercial nuclear power plan: in the United States..
6--^-
f.
/ jz,..
/,. 4...~ - s.
A (4 2 U.S.C. s 2133(b)). The !!RC j
"is respcasible f:: sanaging safety reviews g.
cf applicatiens fe: cens:::::len per=its and l MI /,e y'L i,. L#M cperating *1:e..ses f:: rea::::s and ava*.ua-ti:.s f s:s..fard ;'.a..: designs; svaluates
{
s:..ni:al s;e:1ft:ati::s
- snd* perf:.:.s
- s:..1:sl sviaws sn! analyses :f ne:hani:al, s:::: ::al, a..! na *:ia's en-i.eeri..g as;s:::
cf rese :: sys.s:s, c::e p erf:=a.:e, auxiliary sys:ces, cc. ::1 systs:, me:hani:al c: pene..::,
res:::: s ::::::es, and ;:vs: sys e s."
l a C.r.R. s 1. n us m.
l e
6 e
9 S
l i
O W
]
I s
.I 1
The NRC Offi:e cf N clear 7.eact::
r
.egulatien " reviews applicatiens [for licenses) and issues lice..ses.
. and evaluates the health, safety, and environmental aspects" of a plant prior to the ap; oval of a Preli..ina y Saf ety Analysis i
- 2. Report or a Final Safety Analysis Report, which incorporate the t
6 e
. equip =ent vendors' analyses, evaluations and descripticas of t
I
- cperatica of all co=ponents and syste=s.
10 C.F.R. 5 1.61 (1977),*
i 5 50.34 (1978).
47.
The NRC Office of Nuclear Regulation is required by statute to
"[rleview the safety and safeguards cf all such f acilities, =aterials, and activities, and such review functiens shall include, but not be limited sc menit: ring, testing and rece:-
mending upgrading of syste=s designed to pre-vent substantial health cr safety ha:ards.
42 U.S.C.
S SB43 (b).
1g 48.
Applica*nts fer nuclear plant construction permits i
must subsit fc NRC review and app cval " principal design criter-ia" fer the p cposed facility.
10 C.T.R. S 50.34 (1975). These principal design criteria " establish the necessary design, fabri-catica, ccnstruction, testing, a.d perfer ance require.ents fer
- s rue:ures, syste=s, and cen.:ene.:s imper: ant to safety...."
10 C.T.R.
S 50, Appendbc A.
The NRC has ;;c=ulgated General Design Criteria and has a duty to review equipment and designs for confor=ance to the General Design Criteria which " establish
~
G ininum requireme.ts fe: the principal design criteria" for v
W all cc=mercial nuclear po-e: plants; id.
The NRC has fuber promulgated and has a duty to e. force dditional design require-
/<..~
/
cents der::ibed through::: -he appe. dices :: 10 C.T.R.
S 50.
D'
,j
\\. $
49.
G;; a..i S:e -If relied :n -he *:RC review cf nY S.
l fh e ci;:e.., a.alyser, ;:::einre s a.i ::11..i..; sc ; : vide f:: the l'
~
\\J N safe c; era:ica cf ??. -l.
\\
{
i
-M'
'g,.
l
- )/
/\\
b t
4
C O
s I
I l
I
- eglige.t Ee' view a.d App c tal cf
.a l
i Oc-deal 7.ererts and E&W Generie resiens M
50;. Prior to any licensing subsissic,n by Met-Id fer (payyd N
- OMI-2', the NRC has already revi'eved and neglige'ntly approved
):nu=erous tepical repc ts and generic :.edels prepared by EsW for nuclear plant design and cperation. These topical reports l
l l
3, described generic syste=s in and the operation of B&W nuclear
.i 1
j plants and were a means by which the NRC was able to review j
.fgenericfeaturesence, rather than repetiticusly for each 4
succeeding Esw plant. Licensees, such as Met-Ed, have no input T
'in the creatica cf tcpical reperts and rely on the :iKC to review
- the repcres with due care prier to appreving them fer use in J
subse-ne-- - ' ear plant licensing p:::sedings. 'In reliance I
i;upcn the pric: review and approval by de liKC of tcpical reports,'.
i, prospective lice. sees, such as Met-Id, ince perate such reports
!!by reference into the safety Analysis Report for specific t
{nuclearplants.
{,
@[gC 51.
During the licensing cf OM:-2, the ::KC ack=cv1-b
]edeed that:
f Ma.y features of the design of UMI-2 e n
' * ('
f Cf.r
-r -.
'~
/
are s'- =- o th:se we have evaluated and v
app:cved previcusly fe: cther nu= lear plants new under c=nstru:: ion c in cperatien. To
~~
the extent feasible and appr:priate we have
-)
C, 6-#-M i
relied en cur earlier reviews fer those fea-
'L fy) j tures which were shown to be substantially the same as these previcusly c=nsidered.
I Where this has been dene, the appropriate g
l section of this report identifies the facility involved.
~
,b-,
,.7 n,
!;30 Saf ety Ivaluatien Repc: for the Cperating Li:ense en TM:
~~
/v Set. V
. Unit 2 (1976).
i 52.
CPU and Me:-Id relied up:n and ine::p= sted by
,h g
refere..:a in the TM -: Tinal Safety A.t'.ysis Fepert, a..u.he: ef
,,, [
- 1&n ::piesi ra;:::s p:evi::s'.y revievtl anf.eglige.:1y app::ved c
f by the ::70. 07 a..d *: :-If ress:..shly relied c=
.he ::7.0 ::. ave b
t/
- eviewed these sch-is ri:.s vish d:e cars. These 1.cluded d
/
.C ll 5
19 (j' ~
t a
I
I l
s I
- evie sly a;;
- :ved t:;ical rep:::s, relating to small-break analysis, less-cf-c clant accide..: a.alysis and emergency ecre e
- cooling systen 'perfcznance in 3&W plants of substantially the' t
same design type as TM:-2, spec'ifically nhe B&W type 177-TA
" lowered-loop" nuclear plants.
. Prior to the licensing of Ty.I-2, 2
-, the NRC had licensed eight B&W plants, including seven which I
contained a 177-FA lowered-loop design. The earliest lowered-
['
I loop plant was oconee I, licensed in 1973.
Transient Analyses i
i
\\
) @b 53.
The NRC negligently app cved S&W transient analy-ses for IMI-2, including these fc small-h:eak less-of-ecolant 8
17 accidents and for less =f ncrmal feedwater, even though these t
analyses failed to ec= ply _with URC_.;regula.: ions. The NRC knew f
(f that transient analyses in ec=pliance with NRC regulations are necessary for preper plant design and cperation. A transient is an unintended cht.ge in power level or system cenditien in a f
(.,
nu: lea: pla.t, and includes anticipated cperational c:c ::ences such as a loss-of-ncrnal-feedwater transient, which occurred at 7:C-2cnMa[ch28, 1979.
/Jf
/*
P
/ /
54.
As set ferth in parag:sphs 55 and 56, helow, the S
f f
j
(
NEC, failed to evaluate with due care 3&W transient analyses y
y and failed to conpel 3&W, either during the TP.I-2 plant licensing
'j p ccess or as part of 3&W's prior sub=ission of topical reports, 4
/
to schmit t a=sie=t a=a1yses w. ic ce=pued s.it3 xBC,egu1ations, n
f t.
i..cludin; the standard 7eview Plan a.d the General resign Cri-
.y
.{.
/
teria. As a result, the NRC negligen:1y failed to re:uire 3&M
/
te schsit the ::ansie.: analyses.ecessary fc proper design and eperatic. ef :::-2.
f 55.
The ::K: has simi::ed, vi-h respect :: the ::a.-
y y...
y
/
sien: s.alysts su.-.*::si by 3&: ;;i:: := a..d in : :;::: ef the
/
/
11:n..si.; cf TM:-1, that the ::KC f ailed te e.fe ce cenpliance
.?if l k
W d
Y,' d I
s y
5-i i
n A.
. NN
/
\\
0 l
9.iW( *.
/ tr /
t)
~ v
v v
v 3
~
vish the req;irenents cf its standard Peviev Plan, Secticn 15.'
As the N?.0 has' stated:
l n
s "The TMI-2 accident started with a less of f
I feedvater transient and, because cd the-s ::k-cpen power cperated relief valve, a snall break loss-of-coolant accident resulted. Accc: ding to the standard j
Review Plan, such a sequence should have been analyzed in the licensing process, but it was not."
1
=
NURIG 0560, Staff Repert en the Generic Assess =ent of reedvater
- ansients in the ;WR's Oesigned by The Babcock & Wilecx Co.
(l*79) at 5-4.
56.
The NRC has ad=itted in the respects described in
- aragraph 57,'belew, that it failed to cceply with the require-s-
q) fb nents ef its 3eneral Casign criteria. As the Cen=issica has
~
q 3
(J'a-stated:
l g
"Teedvater transients are anticipated operational p.,
j@ O l
N cec ::ences (Acos), since they are expected to
/ '[~
/
oce : ene c: :::e times during the life of a
/V
..uclea plant. The basic requirements for ACO's are given in General resign Criteria (GDC) 10 and 15.
GDC-10 requires -ha spe:ified accept-p4j able fuel design limits nc: be exceeded during s-ACOs.
G30-14 and GDC-15 require that the design of the reactc; coolant pressure beundary gs, she:1d preclude abner al leakage and the design O'LM c nfiticas of the 50 =dary shecid net be exceeded
,d b/P' d ing A00's.
Additicaal requirements specified in 300-13 are:
':nst =nentasien shall be p crided to n:siter variables and systems ever their anticipated ranges.
for anticipated opera-tienal cecurrences.
. as app cpriate to assure adequate safety.
App cpriate controls shall be p;crided to =aintain these variables and systems within prescribed cperating ranges. '
G20-20 states the general requiranents for ;;ctec-tien rystems, including the folleving:
'The p:cte ica system shall be designed (1) to initiate au :na ically the :;erati:n cf ap;; priate systens including the reactivity centrel systems, to assure that specific acceptable fuel design limits are not ex:eeded as a result of anticipated cperatienal cc::::en=es.
- ..e li h: :f the ::':-2 *:: erience. it is l
e :aren: :na: a::.;:an.e ::::er:a were ne:
et."
l.
7
_r_i. a 3-i
.a ;hasis affad;.
fyy 57.
The *:7.C negligently fallad t: cenply with the 2.-
/
i
- . =la-1:.s f.,:-ihet := ;a a, avss 31 a.-
1e, ase..e, i= t.sa:
i e
I m
9 I
V d
-l 1
the feedwater ::ansien ana* yses and small-break less-cf-ecels..t-analyses schristed by I&W a..d approve d by the '::RC were inadequate
.to p cvide a, p cper b' asis for plant design,and, fcp the develep-1 g
j l'
]:ent of cpera::: training p:cgrt=s and operating procedures.
- pecifi= ally:
l 1
i (a) The NRC failed to require B&W to submit the I
i
- necessary analysis of any break si
- e smaller than 0.040 square 4
I l
l feet.
As a result, the NRC failed to require the necessary l
e.
.e
-i janalyses of breaks equivalent to the si:e of a pilot-operated f
relief valve (0.007 square feet) which had failed to elese.
l I
(b) The NRC f ailed to require B&# to submit the i
- .ecessary analysis of a small break cccurring in the steam space a the t=p =f the pressuri:ar, where the pilot-operated relief I
i
, valve is located.
J
..,1 (c) The NRC failed to requi're B&W to submit the s
necessary analysis of a pilet-cperated relief valve f ailing to
,elese, even thcush such a failure shculd have been assumed since the valve was desig..ated as non-safety grade equipment.
1 (d) The FRC failed to require S&W to submit analyses which examined scre than the initial r.inutes of a
' transient, whereas such analyses sh uld have covered the time jperioduntilastablesystemhadbeenassured.
j (e)
The KRC f ailed sc require E&W to submit analyses cf the sensitivity of the fere;:ing small-break less-of-
- 'ecela..t analyses (subparagraphs a-c, ab:ve), to reacter coolant pur.p :pera:ica c: ren-eperatien.
~~~
53.
As set forth in paragraph 41(b), su::a, the NRC V
knew, at less: at the time ta_f,_.L1_,teviewsd_.andguh' d e d in.
)
r if ! ths *7ea::o Safe y S tudy tW;.!E-14 C 3)," that small-break I'
1:ss-ef-:::lan a:cidents vere subs:t..:dailv==re likelv to
- ur in a :: lea: ;* an tha.. la: e-breah 1:s s-cf-: ela..t t
3 ee f
i 9
?
B e
l
d V
- l~
a::idents.
Yet, -he ':EC f ailed to exa.ine.he 2&W design and
- ccedures with due care to ascertain the likeliheed cf snr.11-
!hreak loss-cf-ccclant kecide.ts and their censequences even i i
)
- after the Davis-Esese incident, which was' a small-break loss-of-3 I
6 Pecolant accident.
8 1
M 59.
The NRC has admitted in post-accident reports l
1
.that the Esw analyses suh=itted to the NRC had f ailed to provide
?
a u
t e
.;necessary infer =ation needed for c;erater actica folleving a i
4 jsnall break. Generie Ivaluatien of snail 3:eak and I.ess of i
fCoelant Accide.t Sahavic: in labeeck & Wile =x Designed 177-TA
- erating 71 ants, :. 7.IG 0565 (19E0) at 1-1.
M. /
T 60.
If the ::RC had reviewed 3&W te;ical reperts and f
thI
. license sch-issicas with due care, and had required 3&W to fW
(/~
l=p cvide the transient analyses required by NRC regulaticas, the l
'W 6
March 28, 1979 accident at TMI-2 would have been avoided.
$7:ccedures
',e 61.
Pric: to the issuance cf the TMI-2 cperating
{
' license, the NRC Office of Inspecticn and Inforcement conducted
,y an ex ensive audit cf the 0::-2 procedures which were drafted by
^*
.1&W.
The audit included a review cf the p ccedures which were
[~f Ilater used by the operators during the. March 28, 1979 accident.
d' g
b
.i '
i he SRC negligently failed to identify deficiencies in these
.4
..IsW-drafted procedures and instead found t'.at the " technical r
i kf F
!centent [of the procedures) was adequate to assure satisfactory p
peric=ance of 1.:e..ded functicns."
Ins,=e:-icn and Inf:-ce ent f.
77-26, August 1977.
.Repcrt ::o.
62.
The ::RC eglige.:1y reviewed p cced::es fer
- erating
- ::-: which 1..:c::actly ;;;hil: si pe=i::ing the f
- essuri
- e
- :: ";: s:*id."
The !.MI-2 :; era-i..; ; ::sfere 5.,.., s...,.. a.,...e,- -,. :..,
...a.
e.a. u. e a.....
a
..,.... 4 s.. e... s.a.....s, I
.n.
e 9
I I
sl ',
l
.p f:ll: wing p::hibiticn:
l" i
"2.1.2 The ; essuri:er/EC System must net he fil*ed with c= lant to s= lid ec..ditiens (400 in:hes) at any tire except.as i
}
re:-uired for syste= h.ydrestatic::ests."
j d.This' pr:.cedure centained no exceptien fc=; e:nergency conditicas
't 1
- even if there were risks of core uncovery.
l i.
s
- i
- 63., The NRC knew or should hr.ve known as a result of a
~-
jitsinvestigationoftheDavis-Besseincidentwhichconfi=ed s.+
!earlierprecurscrs, that the failure of a pilot-operated relief e
1
' valve to clcse would cause the water level in the pressurizer to -
1 ~ise even though the reteter coclant system was net " going r
i g.
- sclid," see paragraphs 25-41, su::a. Nevertheless, in the
.f
.13 scaths fellcwing th Grs
)
74.
As a p cxi. ate result of the forescing, the
): arch 28, 1979 ac:ident at t I-2 cecurred and caused and vill it c:stinue to cause claimants to suffer damages and Icsses in the ifollevingrespects, together with other iters of da.. age inciden-9 i
tal theret=.
(a)
Cla_=a. s have incurred and will centinue to incur expenses for decentaminasien and debris removal --
,. 51,000,000,000.
I (b)
Claisants have incurred and will incur expen-ses fe: repair c: replacement cf damaged and defective plant and s
eq:1;. ant, refueling, upgrading f e::i:ren and syste.s, re-trai.ing cperators and additional expenses for personnel and censultants..ecessitated by the accident -- 5430,000,000.
(c)
In c:de :: seet the needs of their cust: ers f:: ale: ri: ;:ver, :lai:.a.:s ha* e had :: pt: chase and :: :1.ne
- ::::hase f:: c:he
- ill:tes 11d1:::n11 ca;acity 1..i g
i ene:;y and ha.e had -: :; era:e and c:ntinue te cperate thei:
.S $
e y
e O
e i
-.Y t'=7 e
- ess ::s:-a"i:ien: pla.:s linr.e: :han _he.
c:he:.ise v:uld i
have in ::fer :: re';'a:e de less f capari y a..d e.e:gy res lt #
i 1.; f := the :-; arch IE, 1979 accide..: a TM:-2, which caused the shutdevn of O!I-2 and prevented the restart of TMI-l - i 3.
t
.51,590,000,000.
?
s I
e (d) Claimants have lost and vill continue to j
l
- 1cse revenues based on the re.
- toval f c: the rate base of
.?-
I the capital invested in TMI-2, which revenues they would have etherwise earned during the pericd fc: which that unit is net i
in the rate hase -- 5950,000,000.
(e)
Claita.ts have had te and vill continue to in::: increased h:::cving cf capital and at higher rates of interest da. they v::1d have cthe:.ise i.::: red were it nh. fc de accident -- 540,0C0,000.
(f)
In the event dat clai ants are net able to
'restere TMI-2 to cperation, claimants vill icse all of the capital invested in T.v.:-2 -- SE 00,000, 0 0.
A schedule cf damages is atta:hed herete as
. n,. e.._a _x s.
't
?_ ? r.*.,*_t.,
o t.*.-.* *.M..*
~5.
he NRC has reccg.1:ed in its regulatiens that it has the s ecific du:y of "neti' vine. licensees ree.ard-ing ge. erie ;:chless so as to achieve app::priate precautienary c: cc :ective action. " 10 C.T.R.
i 1.f4 (1977).
76.
tihere a gevernment ac.e.:v. has a statutcrv dut}' to varn, c: undertakes to warn and therab-j 1. duces a reliance by a
- ivate party en gever.- e.t actic.., the gever._meht is liable
. _a _a...ag.a.,
. c,..c.. a a,,.. c.u.
...a,
.s.e,,a,
_._... -.. a,i,
w.
.....*.a
.e. a. e s,
e E
..a..a...
..a a.
e.
..s
...... r o r. =..-..
- '.. u...-... *..' ?...#
. - ~ ~ '... '. "..'.,
...,. e d e._*. e s, u9 e.
,e
, u. g.. e. e.n.
t _ e. c. )
4..
e.-
....a s..
n_.
(
...a
.$ I 9
M V
)
I i.
i i
1 1
I*
77.
In licensing a rea*----
'- cperatien, the *:RC 4
l l
1
,6
.' decides that s
t n..
"a reacter whese ICCS.[ emergency core coeling systa=) r.eets the criteria ill i
l control a LOCA [1 css-of-coolant accident)
J,'
- and is, thereicre, sai,e ict cperation....'
I s
8
-l I Union of Cencerned scientists v. Ate $iic Ener=v Con =ission, g
1 I
s i
!499F.2d1069,1037 (D.C. Cir. 197 4 ). A govern =ent agency such t
l' 1
4
' as the NRC which fails to exercise due care in its licensing and j i
fails to ec= ply with its regulaticas is liable under the Federal i
s crt claims Act.
Griffin v. United States, 500 T. 2d 1059 (3d Cir.
e
- 1974)
- In:hers v. Iaster. Air Lines, 373 T.2d 227 (2d Cir.), cert. :
d e..i e d, 359 U.S. 531 (1967); United Ai: lines, Inc. v. weiner, 235 T.id 379 (Sth Cir. ), ce-t. dir-issed sub nen., United Airlines
' Inc. v. United States, 379 U.S.
951 (1964); Hartz v. United
' states, 387 T.2d 870 (5th Cir. 1960).
I,
}
t 9
. f
,::..ti..ei tn.:aga 31:
I 30 I
l d
e 9
t l
l l
O Q
~
8 r
t I
ll l
XHIEIT EI, claimants ; ay fer an t.ard i.. the arcant
?
I cf 54,010,000.,000.
I Cated: New York, !!ew York
'Dece=ber 8, 1520 t
4 KAYE, SCHOLIR, TII?.'G.N, MAYS & EANDLER
.I a
I 3
i 3y (d
v David Klingsberg 425 Park Avenue New York, New York 10022 (212) 759-8400 Of Ocunsel:
Milten Handle 7.ichard C.
Seltzer BIPMcK ISRAILS & LI3I.T'.AN
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James 3. Liberman 26 3 cadway New York, New York 10004 (212) 248-6900 Atter.Eeys fer Claisants Of Ccunsel:
Jesse R. Meer I
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$1ES,000,000 5815,000,000 51,000,000,000 t
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da= aged a-A dafective i
erci;re.t, % _.di:q cf I
eq:iscent a:4 ris c.s a.-A i
tidi.i: a.1 e g e.sas f::
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necessitataf by the a::ida. t, re :=fitti.:g, ret. a.i.-l:q c a rat::s.
0 420,000,000 430,000,000
- ':: eased c:st Cf els: ri ;c.e: &:a ::
p;::?.Lse cf p s.: f:=:
c&- :.dli. ins a-A Costs Cf cpt:Eti.g c' > '.t.ts' less cest-i efficient,-b-ts.
465,000,000 1,125,000,000 1,590,000,000 less cf :wve. e en capitti 1::.asted in 00-2 reeved f:=a
- a.a base 165,C00,000 7E5,000,000 950,000,000 2.c eased c=st of ic. -Lg 15,000,000 25,000,00cY 40,000,000 Czpitti 1. vested in OC-2 200,C00,000
~1/.n..re lesses a:s calculated en the ass = ptien e.at OC-2 a.-d ':M -2 vf.ll rer a cperatica en en fello.-Lg dates:
02-1 Ja..u.ary 1, 1532 Oc-2 Ja cas.:y 1,1958 y J.ar es a.ca. d a.d pa a..: en e:e-be.: 21, 1582.
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