ML20236K289
| ML20236K289 | |
| Person / Time | |
|---|---|
| Site: | San Onofre, Rancho Seco, 05000000 |
| Issue date: | 04/25/1986 |
| From: | Martin J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| To: | Heltemes C NRC OFFICE FOR ANALYSIS & EVALUATION OF OPERATIONAL DATA (AEOD) |
| Shared Package | |
| ML20236K265 | List:
|
| References | |
| FOIA-87-377 NUDOCS 8708070068 | |
| Download: ML20236K289 (17) | |
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NUCLEAR REGULATORY COMMIS REGION V 1450 MARIA LANE, SUITE 210
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APR 2 51986 h MEMORANDUM FOR: C. J. Heltemes, Jr., Director Office for Analysis and Evaluation of Operational Data FROM: J. B. Martin, Regional Administrator
SUBJECT:
ABNORMAL OCCURRENCE REPORT TO CONGRESS FOR FIRST QUARTER CY 1986 This transmits the Region V input to the First Quarter CY 1986 Abnormal Occurrence (AO) Report, in response to your memorandum dated April 7, 1986.
- 1. Loss of Power and Water Hammer Event (San Onofre Unit 1; November 21, 1985).
l An updated A0 writeup is attached as Enclosure 1. A summary of actions being taken (and their status) is included. An executive summary of the restart action plan (with status as of 4/15/86) is also attached (Enclosure 2) for your information.
- 2. Loss of Integrated Control System Power and Overcooling Transient (Rancho Seco; December 26, 1985).
A proposed A0 was forwarded to you by my memorandum dated 4/18/86. The restart action plan was addressed in D. Kirsch's memorandum to F. Miraglia dated 4/8/86.
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- 3. Possible Enclosure 3 Itett - Loss of Decay Heat Removal (San Onofre Unit 2; March 27, 1986).
As discussed with you and P. Bobe by P. Johnson on 4/21/86, a deceription of the loss of decay heat removal at ."an Onofre Unit 2 will be provided for your consideration as a possible Enclosure 3 item. A description of event will be sent to you via 5520 by COB 4/28/86.
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J. .' Ma rt'.u Regional Administrator
Enclosures:
As stated s
cc: H. Denton, NRR J. Taylor, IE N --
( GORDON87-377 PDR
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I LOSS OF POWER AND WATER HAMMER EVENT The following information pertaining to this event is also being reported concurrently in the Federal Register. Appendix A (see the second general criterion) of this report notes that a major degradation of essential safety-related equipment can be considered an abnormal occurrence.
Date and Place - On November 21, 1985, San Onofre Nuclear Generating Station (SONGS), Unit 1, experienced a partial loss of inplant ac electrical power while the plant was operating at 60 percent power. Following a manual reactor trip, the plant lost all inplant ac power for 4 minutes and experienced a severe water hammer in the feedwater system which caused a leak, damaged plant equipment, and challenged the integrity of the plant's heat sink. SONGS Unit 1 utilizes a Westinghouse-designed pressurized water reator. The plant is operated by Southern California Edison Company (the licensee) and is located south of San Clemente, California.
The event involved several equipment malfunctions and extensive operator actions, including operator actions outside the control room.
Nature and Probable Consequences 1
At 4:51 a.m., on November 21, 1985, the plant was operating at 60 percent ;
power, when a ground fault was detected by protective relays associated with the "C" transformer, which was supplying offsite power to one of two safety related 4160V electrical buses. The resulting isolation of the transformer caused the safety-related bus to de-energize, which tripped all feedwater and condensate pumps on the east side of the plant. The pumps on the west side of the plant were unaffected since their power was supplied from another bus which was being fed from the main generator. The continued operation of the west feedwater and condensate pumps, in combination with the failure of the east feedwater pump discharge check valve to close, resulted in over-pressurization and rupture of the east side flash evaporator (part of a low-pressure feed heater unit). The operators, as required by emergency procedures dealing with electrical systems, tripped the reactor and turbine-generator. As a result, the plant experienced its first complete loss of steam generator feedwater and inplant electrical power since it began operation.
The manual trip of the main generator caused loss of AC power to remaining in-plant loads. The subsequent 4-minute loss of inplant electric power started the emergency diesel generators (which by design did not load), de-energized all safety-related pumps and motors, significantly reduced the number of control room instrument indications available for operators to diagnose plant conditions, produced spurious indications of safety injection system actuation, and caused the NRC Emergency Notification System (ENS) phone on the operator's desk to ring spuriously. Restoration on inplant electric power was delayed by an unexpected response of an automatic cequence (that should have established conditions for remote-manual access to offsite power still t available in the switchyard). l l
Enclosure (1)
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The loss of steam generator feedwater was the direct result of the loss of AC power to the two main feedwater and one auxiliary feedwater pump motors, and i the designed 3-minute warm-up period of the steam-powered auxiliary feedwater l pump. The loss of the feedwater pumps, in combination with the failure of five feedwater check valves to close (one at the discharge of each feedwater l pump and one in the feedwater line to each of the three steam generators),
allowed loss of inventory from all three steam generators and the partial voiding of the long horizontal runs of feedwater piping within the containment i building. The subsequent automatic start of feedwater injection by the j steam-powered auxiliary feedwater pump did not result in the recovery of steam l generator level because the auxiliary feedwater being injected into the '
feedwater lines was flowing backwards through the failed check valves to the ruptured feed heater in the condensate system. Later, operators isolated the feedwater lines upstream of the failed check valves, as required by procedure, unknowingly initiating the process of refilling the feedwater lines in the containment building. Before all feedwater lines were refilled, a severe water hammer occurred that displaced one feedwater line in the containment l building by distances ranging up to approximately one foot, damaged its associated pipe supports and snubbers, broke a feedwater control valve actuator yoke, and stretched the studs, lifted the bonnet, and deformed the gasket on a 4-inch feedwater check valve. The damaged check valve developed a significant steam-water leak, the second leak in the event.
The second Icak, in combination with an earlier inadvertent re-establishment of steam generator blowdown, caused all three steam generator water levels to drop below indicating levels. Steam from all three steam generators fed the leak, because of the absence of individual main steam isolation valves.
Despite these problems, operators later succeeded in recovering level j irdication in the two steam generators not directly associated with the l feedwater piping leak. With the reestablishment of steam generator levels, l the operators safely brought the plant to a stable cold ehutdown condition, without a significant release of radioactivity to the environment (the preexisting primary to secondary leak was not exacerbated) and without significant additional damage to plant equipment.
Cause or Causes - As described in more detail above, the most significant aspect of the event involved the failure of five safety-relatei check valves in the feedwater system. These failures occurred in less than a year, without detection, and jeopardized the integrity of safety systems.
Actions Taken to Prevent Recurrence l Licensee Actions - The licensee has undertaken an extensive study (including testing programs) of the multiple failures associated with the event to determine root causes and effective corrective actions to preclude recurrence.
On April 8, 1986, subsequent to several meetings with NRC staff and the Commission, the licensee submitted a comprehensive report, reference (1),
documenting the results of their investigations to that date and providing some conclusions and corrective actions being implemented. In reference (1),
the licensee committed to provide additional information by April 30, 1986.
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The licensee concluded that the most likely cause of the cable failure which initiated the event was temperature-induced degradation due to the presence of local heat sources such as hot pipe flanges. Additionally, the licensee concluded that the failure of the five check valves was caused by (1) their proximity to turbulent flow. (2) the fact that the valves were oversized and therefore did not remain fully open in normal operation, (3) the design by which the valve disc was fastened to the valve hinge, and (4) extended reduced flow operation at 90% power which exacerbated the effects of the design deficiencies.
The licensee's actions described in the April 8, 1986 report were extensive and included examinations and corrective actions in the areas of testing, procedures development, training, maintenance, quality assurance, emergency preparedness, post-trip review and safety review programs.
The licensee committed to and is in the process of implementing a number of corrective actions including repairs and design changes which include redesign and replacement of the damaged feedwater lines, replacement of the failed check valve design with another design, and adding an additional check valve in each feedwater line.
Additionally the licensee has committed to substantial initiatives to improve plant performance. These initiatives will systematically examine the material condition of the unit and identify and correct systems and components which deviate from defined standard conditions. The licensee has elicited the a.td of recognized experts in this area and has committed to implement necessary actions prior to restart. Additional actions are being defined to maintain the material standard on an ongoing basis.
NRC Actions - The San Onofre Resident Inspectors arrived at the site shortly after being notified of the event. They observed licensee actior.s to assure the plant remained in a stable condition and began an initial investigation of the circumstances associated with the event.
On November 21, 1985, the Regional Administrator forwarded a Confirmatory Action Letter to the licensee (Ref. (2)) indicating, in part, that the licensee would not perform any additional work on equipment that malfunctioned during the event until the NRC investigation could review the licensee's proposed actions. The letter also confirmed an understanding that the plant was not to be restarted until authorized by the NRC Region V Regional Administrator or his designee.
l On November 22, 1985, responsibility for the incident investigation was assigned to a special NRC Incident Investigation Team (IIT) by the NRC j Executive Director for Operations at the request,of the Region V Regional '
Administrator, in conformance with an NRC staff-proposed Incident i Investigation Program. The Team, composed of six technical experts, was to (1) determine pertinent facts related to the event; (2) identify the probable ;
cause; and (3) make appropriate findings and conclusions to form the basis for. ;
possible follow-on actions. The Team began their investigation at the plant '
site on November 23, 1985. The equipment which malfunctioned was quarantined. 1 l
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8 The Team collected and evaluated information to determine the sequence of operator, plant, and equipment responses during the event and the causes of equipment malfunctions. The sequence of these responses was determined primarily by interviewing personnel who were at the plant during the event and '
l by reviewing plant data for the period immediately preceding and during the l event. The Team also toured the plant to examine the equipment which l malfunctioned, the equipment that was key to mitigating the transient, and the l control room instrumentation and controls. The Team also interviewed plant management personnel and NRC Region V personnel who arrived at the site soon I after the plant was stabilized about their knowledge of the plant response and operator actions. By correlating plant records with personnel statements on their actions and observations, the Team was abic to compile a description of the event.
l The results of the Team's investigation were issued in NUREG-1190 (Ref. (3)). l l Problems identified included issues specific to SONGS Unit 1 and several l
{ possible generic issues. In addition, the Team concluded that the most l
l significant aspect of the event was that five safety-related feedwater system l check valves degraded to the point of inoperability during a period of less than a year, without detection, and that their failure jeopardized the integrity of safety-related feedwater piping.
The root causes of the check valve failures have been determined by SCE and are under independent review by the NRC. Potential contributors to this problem include inadequate inservice testing (IST), inadequate design, and inadequate consideration of the effects of reduced power operations. The licensee's IST program (submitted to but not yet approved by NRC) provided for testing a sampling of the check valves each quarter, but permitted deferral of testing when plant conditions were inappropriate (e.g., plant in operation).
The testing was also intended to identify valve failure, not degradation or impending failure. The IST was therefore not effective in identifying the l check valve failures before the event occurred. Finally, reduced power operations at Unit 1 are now routine because of steam generator tube plugging ,
and sleeving, and the reduced feedwater flow may have increased the I susceptibility of check valve components to hydraulically-induced vibration.
The NRC continues to be involved in the resolution of this event and related matters. The event provided an opportunity for the NRC to learn from experience and to feed back the pertinent lessons into NRC and licensee -
activities. The Executive Director for Operations has directed NRC program 1 managers to conduct an in-depth reappraisal of the effectiveness of their programs in light of the lessons of the SONGS Unit 1 event with the view of making the NRC programs more effective. An NRC action plan has been developed through a cooperative effort of the Offices of Nuclear Reactor Regulation, l Inspection and Enforcement, and Region V.
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l This plan resulted in three basic types of actions that the staff is undertaking:
- Evaluation of licensee corrective actions and evaluations that are required for restart in accordance with the NRC's action list. Most notably, these include an assessment of the licensee's review of plant material condition and readiness for operation. i Evaluation of generic implications of the SONGS Unit 1 event through a sampling of industry experience and technical evaluations of root causes; e.g., check valve design implementation.
Evaluation of NRC requirements and positions in light of existing j implementation practices, root causes of the event, and samples of I industry practices.
These actions outline a program that evaluates the SONGS Unit I readiness for restart and assures that generic aspects are considered. i l Future reports will be made as appropriate.
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References for Enclosure (1)
(1) Letter from M. O. Medford, Manager of Nuclear Engineering, Southern California Edison Company to A. E. Chaffee, Chief, Reactor Project Branch, NRC Region V, April 8, 1986 (2) Confirmatory Action Letter from J. B. Martin, Regional Administrator, NRC Region V to R. B. Ray, Vice President, Southern California Edison Company, Docket No. 50-206, November 21, 1986.
(3) NUREG 1190 Loss of Power and Water Hammer Event at San Onofre Unit 1 on November 21, 1985, published January 1986.
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EXECUTIVE
SUMMARY
SAN ONOFRE 1 ACTION ITEM STATCS The following provides a summary status of NRC actions tequired in response to the San Onofre 1 trip and water hammer event of November 21, 1985.
Item Description Resp. Office Status
- 1. Implement / coordinate IE (Lead) actions to assure NRR reifability of safety- RV related check valves.
Plant-Specific Review of licensee's event analysis is underway based on SCE's submittal of 4/8/86; additional information is due 4/30/86. Other activities planned or in progress.
Generic adequacy of 5 planned plant visits check valve design are complete, 5 planned and IST pregrams valve vendor inspections are complete, reviews and I documentation of results in progress.
- 2. Assess need to reevaluate NRR In progress.
unresolved safet:r issue A-1 (water hammer) to address condensation-induced water hammers in feedwater piping (EDO Memo, Item 2)
- 3. Assess adequacy of SONGS 1 NRR (lead) Licensee submittal design features (other RV received 4/8/86.
than check valves). ' Design features review (including feedwster system) in progress.
RV review / inspection of A design changes is ongoing.
Enclosure (2)
2 Item Description Resp. Office Status l
- 4. Evaluate abnormal condition NRR (Lead) and post-trip reviews Generic requirements Licensee's submittal for post-trip reviews of 4/8/86 under review and data retrieval capabilities.
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Plant-specific concerns Licensee's submittal of 4/8/86 under review.
- 5. Evaluate overall adequacy RV Inspection in progress.
of SONGS 1 maintenance records
progress.
- 7. Adequacy of training RV (Lead) Licensee's submittal i NRR of 04/08/86 under I review. Inspection scheduled 5/86.
RV Adequacy of licensee Licensee's submittal procedures / training of 4/8/86 under review.
Evaluate NRC policies In progress for use of ENS Evaluate spurious Licensee's submittal ringing of ENS of 4/8/86 under review.
- 9. Evnluate existence /signifi- NRR In progress.
cance of backlog of license amendments.
- 10. Evaluate effect of long RV (Lead) Complete (Stello ltr to outage on plant components IE Commissioners,4/7/86)
- 11. Evaluate licensee actions RV Ongoing to review / upgrade overall plant material condition of SONGS-1 prior to restart I
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APR 101986 yt I
i MEMORANDUM FOR: C. J. Heltemes, Jr., Director Office for Analysis and Evaluation of ,
Operational Data ,
I FROM: John B. Martin, Regional Administrator
SUBJECT:
ABNORMAL OCCURRENCE REPORT TO CONdRESS FOR RANCHO SECO DECEMBER 26, 1985 EVENT In accordance with a phone conversation between J. L. Crews of my staff and P. E. Bobe of your office on March 31, 1986, a draf t of an Abnormal Occurrence Report is enclosed. The enclosure addrasses the December 26, 1985 event i subject, loss of ICS power and overcooling transient. I lI
[wJ(onnB. Martin,,1f 0-Regional Administra r -
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Enclosure:
As stated cc:
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1 ENCLOSURE
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LOSS OF INTEGRATED CONTROL SYSTEM POWER AND OVERC00 LING TRANSIENT l
The following information pertaining to this event is also being reported concurrently in the Federal Register. Appendix A (see the third general I criterion) of this report notes that major deficiencies in design, !
construction, use of, or management controls for licensed f acilities or material can be considered an abnormal occurrence.
Date and Place j l
On December 26, 1985, Rancho Seco Nuclear Generating Station, located in Clay, j California, about 25 miles southeast of Sacramento, experienced a loss of de I power within the integrated control system (ICS) while the plant was !
operating at 76 percent power. Following the loss of ICS de power, the I reactor tripped on high reactor coolant system (RCS) pressure followed by a -
rapid overcooling transient and automatic initiation of the safety features actuation system on low RCS pressure. The overcooling transient continued '
until ICS de power was restored 26 minutes after its" loss. The significance of the event is that a nonsafety related system failure initiated a plant transient which could have been more severe under other postulated scenarios.
I The Rancho Seco Nuclear Generating Station, operated by the Sacramento l Municipal Utility District (SMUD) is a 916-MWe Babcock & Wilcox !
(B6W)-designed pressurized water reactor. The plant received an NRC operating j license in A974. l l
l Nature and Probable Consequences l
At 4:14 a.m. on December 26, 1985, the plant was operating at 76 percent l
power, when a loss of integrated control system (ICS) de power occurred as a I result of a single failure. The loss of de power to the ICS (a nonsafety-related system) caused a number of feedwater and steam valves.co i i reposition automatically and also caused the loss of remote control of the .
affected valves from the control room. In addition, the main feedwater (MFW) I pump turbines slowed to minimum speed and the auxiliary feedwater (AFW) pumps started. The immediate result was a reactor coolant system (RCS) undercooking condition that resulted in the reactor tripping on high pressure. The reactor trip was followed by an overcooling condition that resulted in safety features actuation and excessive RCS cooldown.
l The transient was initiated by the failure of a single module in the nonsafety-related ICS (i.e. the spurious tripping of the power supply module that interrupted all +/-24 Vdc power). The most probable cause of this failure was a design weakness that apparently made the circuit susceptible to erratic operation if " contact resistance" between the 24 Vdc bus and the pcwer supply monitor were to develop, and the development of a high resistance < onnection (i.e. a bad crimp connection) in the wiring between the
+24 Vdc bus and the power supply monitor which exposed the design weakness and caused the module to trip.
. .I 2 The operators were not immediately able to restore de power within the ICS. '
As a result, nonlicensed operators were sent to isolate the affected steam and feedwater valves locally with handwheels. During the first 7 minutes of i the incident, the excessive steam and feedwater flows resulted in a rapid RCS cooldown of over 100 *F. The pressurizer emptied and a small bubble formed in the reactor vessel head + The RCS cooldown continued and the RCS depressurized to about 1064 psig and then began to repressurize. This repressurization resulted in the RCS entering the BW-designated pressurized !
thermal shock (PTS) region. The atmospheric dump valves and turbine bypass '
j valves were isolated within 9 minutes af ter the reactor trip. However, the ,
i operators experienced difficulty closing the ICS-controlled AW flow control valves. One of the flow control valves was finally shut; however, the second '
AW flow control valve was damaged and failed open. The associated ' AW manual l isolation valve was found to be stuck open. Therefore, both AW pumps continued to feed and overfill one steam generator. Since the plant has no main steam isolation vslves, water began to overflow into the main steam
- 3 lines, i
b About 26 minutes af ter the reactor trip, the operators restored power within l the ICS by reclosing two switches in an ICS cabinet. The operators were then j able to close the open AW flow control valve from the control room, which scopped the RCS cooldown, and started stabilizing.the plant. The RCS had cooled down a total of 180 'F in this 26+ minute period.
While changing a valve lineup in the suction of the pump used to supply RCS makeup (makeup pump), the last suction valve to the makeup pump was inadvertently shut. This resulted in the overheating and destruction of the makeup pump. About 450 gallons of contaminated water were spilled on the j floor. This failure did not directly affect the incident since a high i pressure injection (HPI) pump was available to supply RCS makeup. In l addition, the spilled water did not result in any significant onsite or !
offsite radioactivity re. lease or personnel dose.
s Operators later stabilized the plant and brought it to a cold shutdown l without a significant release of radioactivity to the environment and without significant additional damage to plant equipment. l the December 26, 1985 overcooling incident did not seriously threaten the integrity cf the Rancho Seco reactor veses1. However, the plant has had a number of overcooling incidents in its 12-year operating history. Each time this occurs the potential exists for_ additional operator errors and equipment
, failures that might exacerbate the event and seriously threaten reactor l integrity. Thus, the significance of this incident lies in the fact that under alternate scenarios more serious consequences could occur.
The incident at Rancho Seco was also significant because a single failurn in the integrated control system (ICS), which is a nonsafety-related system, subjected the plant to 4 undesirable overcooling transient. During the transient, the RCS cooled down 180'F in 26 minutes, the pressurizer emptied, a l bubble formed in the reactor vessel haad, the plant entered the pressurized i thermal shock region, the safety features actuation system (SFAS) actuated, and water overflowed from a steam generator into the main steam lines.
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I Cause or Causes
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l The fundamental causes for this transient were design weaknesses and I vulnerabilities in the ICS and in the equipment controlled by that system. 1 These weaknesses and vulnerabilities were not adequately compensated by other '
design features, plant procedures or operator training. These weaknesses l l and vulnerabilities were largely known to Sacramento Municipal Utility District (SMUD) and the NRC staff by virtue of a number of precursor events and through related analyses and studies. Yet, adequate plant modifications l were not made so that this event would be improbable, or so that its course or consequences would be significantly altered.
Actions Taken To Prevent Fe;urrence I
Licensee 1 l
The licensee has underteken extensive study (including controlled i l disassembly, examination and testing) of the multiple failutes associated I with the event to determine root causes and to take;cprrective actions to prevent recurrence. Some specific improvements have been id6ntified by these efforts and are being implemented prior to plant startup. T.hese are described in the licensee's February 19, 1986 su naary report to the NRC (Reference 1).
Plant Modifications
- 1. Replacement of power distribution wiring of the ICS Power Supply l Monitor, to reduce the resistance in series with the voltage being j monitored. i l
- 2. Provision of failed cloced features cad /or control tsom manual control l capability for turbine bypars valvec, atanspheric duop valves, and au:;12iary feedwater flow control valves, under circumstances of ICS a failure. .
Training Classraom and simulator craining relating to respoose to ICS power loss conditions, handling of overcooling and potential pressurized thermal shock, recovery from safety system acticas and implementation of emergency plan procedures.
Maintenance Progrgm
- 1. Repair of damaged equipment that is req ~e ired for normal and abnormal operating conditions.
- 2. Verification of acceptabla condition of equipment in the non-nuclear systems of the platr.
- 3. Development of a preventive maintenance program for non-nuclear, balance-of-plant equipment.
1 t 4 Emergenc.y Procedures
( Development of event related procedures to complement the symptom related l emergency procedures, for ICS power loss and safety feature actuation system recovery.
NRC Upon being notified of the event, the NRC Resident inspectors for the plant arrived shortly thereafter. They observed licensee actions to assure the plant remained in a stable condition and began an initial investigation of the circumstances associated with the event.
On December 26, 1985, the Regional Administrator of the h7C Region V Office forwarded two Confirmatory Action Letters to the licensee (References 2 and 3) indicating that the licensee would perform a root cause analysis prior to e return to power and would not perform any additional work on equipment that malfunctioned during the event until the NRC could evaluate the event.
! Or December 27, 1985, an NRC Augmented Inspection Tda (AIT) was sent to the l site by the R.egional Administrator and started transcribed personnel interviews on December 28. The initial results of this investigation effort indicated that the event was complex and had potentially significant generic implications.
On December 31, 1985, the responsibility for the incident investigation was expanded to a special NRC Incident Investigation Team by the NRC Executive Director for Operations at the request of the Region V Regional Administrator, in conformance with a NRC staff-proposed Incident ,
investigation Program. The Team, composed of six technien1 experts, was to j (1) fact-find as to what happened; (2) identify the probable cause as to why I it happened; and (3) make appropriate findings and conclusions to form the basis for possible follow-on actions. The Team consisted of the AIT members w supplemented by additional staff. It continued the investigation started by the AIT at the plant site. The equipment which malfunctioned was quarantined.
The Tene collected and evaluated information to determine the sequence of operator, plant, and equipment responses during the event and the causes of equipment malfunctions. The sequence of these responses was determined primarily by interviewing personnel who were at the plant during the event and by reviewing plant data for the period immediately preceding and during the event. The Team also toured the plant to examine the equipment which malfunctioned, the equipment that was key to mitigating the transient, and the control room instrumentation and controls. The Team also interviewed plant management personnel and NRC Region V personnel who arrived at the site soon after the plant was stabilized about their knowledge of the plant response and operator actions. By correlating plant records with personnel statement on their actions and observations, the Team was able to compile a picture of the event.
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t 5 I During and subsequent to their onsite activities the Team reviewed and I concurred ir specific troubleshooting plans developed by the licensee for equipment disassembly, inspection and testing. Several of there activities were witnessed by NRC inspectors. ;
The results of the Team's investigation are contained in NUREG-1195 (Reference 4). Problems identified included issues specific to Rancho Seco I and several possible generic issues. The Team concluded that design l weaknesses and vulnerabilities in the ICS and in the equipment controlled by I that system were not adequately compensated by other design features, pl6nt procedures or operator training. These weaknesses and vulnerabilities were largely known to Sacramento Municipal Utility District (SMUD) and the NRC j staff by virtue of a number of precersor events and through related analyses I and studies. Yet, adequate plant modifications were not made so that this event would be improbable, or so that its course or consequences would be significantly altered. -
j The NRC continues to be involved in the resolution of this event'and related matters. The Executive Director for Operations has1 directed (Reference 5)
NRC program managers to conduct further generic and plant specific follow-up j i
actions. Development of NRC plant specific action plans commenced while the )
IIT was on-site in January 1986, and have been expanded subsequently to include a review of the completeness of prior staff and licensee actions ,
associated with the control systems.
The Executive Director for Operations has also considered this event in a January 24, 1986 (Reference 6) request to the B&W Owners Group, to obtain an industry effort to assess the generic aspects of plant responses to transients, and methods of reduction of the number of plant trips. The Owners '
Group has been responsive to this request and on April 8, 1986 presented a draft plan of action for conducting the r.ecessary reviews.
In addition to addressing those issues which have arisen directly as a result ,
i of the December 26, 1985 cooldown transient, the NRC regional office has '
re-evaluated the status of prior Rancho Seco open inspection findings to identify matters tihich should be resolved prior to restart of the plant.
The licensee and NRR have included these in restart plans. Also, the NRC staff has encouraged the licensee to reexamine the status of all critical plant systems to assure readiness for operation and maximum reliability, so that operation of the plant may be continued with a low probability of disruption from internal causes. Some of these efforts will be observed by NRC inspectors.
l These actions outline a program that evaluates the Rancho Seco restart program, and assures that generic aspects are considered.
Future reports will be made as appropriate.
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1 L 6 References '
- 1. Letter from R. J. Rodriguez, Assistant General Manager, Sacramento I Utility District to J. B. Martin, Regional Administrator, NRC Region V .
and F. J. Miraglia Jr. , Director, PWR-B Division NRC February 19, 1986, l forwarding a summary report " Description end Resolution of issues J Regarding the December :4, 1985 Reactor Trip".
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- 2. Letter from J. B. Martin, Regional Administrator, NRC Region V to R. J. Rodriguez, Assistant General Manager, Sacramento Municipal Utility District, December 26. 1985, requesting that a root cause analysis be completed before return to power.
- 3. Letter from J. B. Martin, Regfonal Administrator, NRC Region V to l
R. J. Rodriguez, Assistant General Manager, Secramento Municipal Utility District, December 26, 1985, requesting the licensee to hold in abeyance any repair work planned on equipment that malfunctioned
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- 4. U. S. Nuclear Regulatory Commission, " Loss of Integrated Control Systez '
i Power and Overcooling Transient at Rancho Seco on December 26, 1985",
j USNRC Report NUREG-1195. published February 1986.
- 5. Memorandum, Victor Stello, Jr., Acting Executive Director for Operations l to Harold R. Denton, Director,. NRR; James E. Taylor, Director IE; I
John B. Martin, Regional Administrator, Region V, dated March 13, 1986 -
" Staff Actions Resulting from the Investigation of the December 26, 1985 Incident at Rancho Seco (NUREG-1195)".
- 6. Letter from Victor Stello, Jr., Executive Director for Operations, to Hal Tucker, Chairman of Babcock and Wilcox Owners Group, dated January 24, 1986 requesting sn evaluation of design of B6W plants for reduction of plant trips and mitigating transient response.
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