ML20238B443

Forwards Draft of AO Rept Addressing 851226 Loss of Integrated Control Sys Power & Overcooling Transient at Facility,Per 860331 Telcon Between Jl Crews & PE Bobe

ML20238B443
Person / Time
Site:	Rancho Seco, 05000000
Issue date:	04/13/1986
From:	Martin J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V)
To:	Heltemes C NRC OFFICE FOR ANALYSIS & EVALUATION OF OPERATIONAL DATA (AEOD)
References
FOIA-87-377, RTR-NUREG-0090, RTR-NUREG-90 NUDOCS 8709010217
Download: ML20238B443 (8)

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, , */  %,, UNITED STATES  !

! o NUCLEAR REGULATORY COMMISSION

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J' APR 131986 t

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MEMORANDUM FOR: C. J. Heltemes, Jr. , Director t Office for Analysis ar.d Evaluation of ,

Operational Data ,

FROM: John B. Martin, Regional Administrator l

SUBJECT:

\' t ABNORMAL OCCURRENCE REPORT TO CONGRESS FOR RANCHO SECO DECEMBER 26, 1985 EVENT In accordance with a phone conversation between J. L. Crews of my staff and P. E. Bobs of your office on March 31, 1986, a draft of an Abnormal Occurrence Report is enclosed. The enclosure addresses the December 26, 1985 event subject, loss of ICS power and overcooling transient.

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[,v o n B. Martin

, , Regional Administra r .

Enclosure:

As stated cc:

J. L. Crews P. E. Bobe >

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ENCLOSURE 4 e.

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-h LOSS OF INTEGRATED CONTROL SYSTEM POWER AND OVERC00 LING TRANSIENT t

. The following'information pertaining to this event is also being reported .

concurrently in the Federal Register. Appendix A (see the third general criterion) of this report notes that major deficiencies in design, construction, use of, or management controls for licensed facilities or material can be considered an abnormal occurrence.

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'Date and place on December 26, 1085, Rancho Seco Nuclear Generating Station, located in Clay, California, about 25 miles southeast of Sacramento, experienced a loss of de power within the integrated control system (ICS) while the plant was open.:ing at 76 percent power. Following the loss of ICS de power, the 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 pcver 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,

l. The Rancho Seco Nuclear Generating Station, operated by the Sacramento Municipal Utility District (SMUD) is a 916-MWe Babcock & Wilcox (B&W)-designed pressurized water reactor. The plant received an NRC operating license in 1974

Nature and probable Consequences At 4

14 a.m. on December 26, 1985, the plant was operating at 76 percent power, when a loss of integrated control system (ICS) de power occurred as a 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.to l b reposition automatically and also caused the loss of remote control of the .

l affected valves from the control room. In addition, the main feedwater (MFW) pump turbines slowed to minimum speed and the auxiliary feedwater (AFW) pumps started. The immediate result was a reactor coolant system (RCS) l 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.

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 power supply monitor were to develop, and the development of a high resistance connection (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.

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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 f eedwater valves locally with handwheels. During the first 7 minutes of 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 l repressurizatf.on resulted in the RCS entering the B&W-designated pressurized I thermal shock (PTS) region. -The atmospheric dump valves and turbine bypass valves were isolated within 9 minutes after the reactor trip. However, the operators experienced difficulty closing the ICS-controlled AFW flow control valves. One of the flow control valves was "inally shut; however, the second AFW flow control valve was damaged and failed open. lhe associated AFW manual isolation valve was found to be stuck open. Therefore, both AFW pumps continued to feed and overfill one steam generator. Since the plant has no main steam isolation valves, water began to overflow into the main steam lines.

About 26 minutes after the reactor trip, the operators restored power within the ICS by reclosing two switches in an ICS cabinet. The operators were then able to close the open AFW flow control valve from the control room, which stopped 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 floor. This failure did not directly affect the incident since a high  ;

pressure inj ection (HPI) pump was available to supply RCS makeup. In {

addition, the spilled water did not result in any significant onsite or '

offsite radioactivity release or personnel dose.

i Operators later stabilized the plant and brought it to a cold shutdown . j without a significant release of radioactivity to the environment and without  ;

significant additional damage to plant equipment. I The December 26, 1985 overcooling incident did not seriously threaten the integrity of the Rancho Seco reactor vessel. 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 i failures that might exacerbate the event and seriously threaten reactor integrity. Thus, the significance of this incident lies in the fact that under alternate scenarios more serious consequence could occur.

The incident at Rancho Seco was also significant because a single failure in the integrated control system (ICS), which is a nonsafety-related system, subjected the plant to an undesirable overcooling transient. During the transient, the RCS cooled down 180*F in 26 minutes, the pressurizer emptied, a bubble formed in the reactor vessel head. the plant entered the pressurized 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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Cause or Causes i 4

The fundamental causes for this transient were design weaknesses and vulnerabilities in the ICS and in the equipment controlled by that system. .

These weaknesses and vulnerabilities were not adequately compensated by other' design features, plant procedures or operator training. These weaknesses 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 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 Recurrence Licensee i

The licensee has undertaken extensive study (including controlled I disassembly, examination and testing) of the multiple failures associated with the event to determine root causes and to take corrective actions to prevent recurrence. Some specific improvements have been identified by these efforts and are being implemented prior to plant startup. These are described in the licensee's February 19, 1986 summary report to the NRC s (Reference 1).

Plant Modifications l 1. Replacement of power distribution wiring of the ICS Power Supply Monitor, to reduce the resistance in series with the voltage being monitored.

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Provision of failed closed features and/or control room manual control I capability for turbine bypass valves, atmospheric dump valves, and auxiliary feedwater flow control valves, under circumstances of ICS J i fcilure. .

1 Training Classroom and simulator training relating to response to ICS power loss conditions, handling of overcooling and potential pressurized thermal shock, recovery from scfety system actions and implementation of emergency plan procedures.

Maintenance Program i

1. Repair of damaged equipment that is required for normal and abnormal operating conditions.

2. Verification of acceptable condition of equipment in the non-nuclear systems of the plant.

3. Development of a preventive maintenance program for non-nuclear, balance-of-plant equipment.

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. 4 Emergency Procedures Development of event related procedures to complement the symptom related emergency procedures, for ICS power loss and safety feature actuation .

system recovery.

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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 NRC 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 return to power and would not perform any additional work on equipment that malfunctioned during the event until the NRC could evaluate the event.

On December 27, 1985, an NRC Augmented Inspection Team (AIT) was sent to the site by the Regional 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 re' quest of the Region V Regional Administrator, in conformance with a NRC staff-proposed Incident investigation Program. The Team, composed of six technical experts, was to (1) fact-find as to what happened; (2) identify the probable cause as to why it happened; and (3) make appropriate findings and conclusions to form the i basis for possible follow-on actions. The Team consisted of the AIT members i supplemented by additional staff. It continued the investigation started by -

the AIT at the plant site, The equipment which malfunctioned was quarantined.  ;

The Team collected and evaluated information to determine che sequence of I 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. 1 1

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. 5 During and subsequent to their onsite activities the Team reviewed and concurred in specific troubleshooting plans developed by the licensee for equipment disassembly, inspection and testing. Several of these activities were witnessed by NRC inspectors, The results of the Team's investigation are contained in NUREG-1195 l' (Reference 4). Problems identified included issues specific to Rancho Seco and several possible generic issues. The Team concluded that design weaknesses and vulnerabilities in the ICS and in the equipment controlled by that system were not adequately compensated by other design features, plant procedures or operator training. These weaknesses 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 studios. 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.

The NRC continues to be involved in the resolution of this event and related matters. The Executive Director for Operations has directed (Reference 5)

NRC program managers to conduct further generic and plant specific follow-up 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 j associated with the control systems.

The Executive Director for Operations has also considered this evenc in a  ;

January 24, 1986 (Reference 6) request to the B&W Owners Group, to obtain an i industry effort to assess the generic aspects of plant responses to '

i transients, and methods of reduction of the number of plant trips. The Owners l

' Group has been responsive to this request and on April 8, 1986 presented a draft plan of action for conducting the necessary reviews.

l In addition to addressing those issues which have arisen directly as a result '

. of the December 26, 1985 cooldown transient, the NRC regional office has .

re-evaluated the status cf prior Rancho Seco open inspection findings to identify matters which 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.

These actions outline a program that evaluates the Rancho Seco restart program, and assures that generic aspects are considered.

l Future reports will be made as appropriate.

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Refeesnces

1. Letter from R. J. Rodriguez, Assistant General Manager Sacramento Utility District to J. B. Martin, Regional Administrator, NRC Region V .

and F. J. Miraglia Jr. , Director, FWR-B Division, NRC February 19, 1986, forwarding a summary report " Description and Resolution of issues Regarding the December 26, 1985 Reactor Trip".

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, Regional Administrator, NRC Region V to R. J. Rodriguez, Assistant General Manager, Sacramento Municipal Utility District, December 26, 1985, requesting the licensee to hold in abeyance l any repair work planned on equipment that malfunctioned l i

4. U. S. Nuclear Regulatory Commission, " Loss of Integrated Control System l Power and Overcooling Transient at Rancho Seco on December 26, 1985",  !

USNRC Report NUREG-1195, published February 1986. I

5. Memorandum, Victor Stello, Jr., Acting Executive Director for Operations to Harold R. Denton, Director, NRR; James E. Taylor, Director IE; John B. Martin, Regional Administrator, Region V, dated March 13, 1986 -

" Staff Actions Resulting from the Investigation of the December 26, 1985 I

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 an evaluation of design of B&W plants for reduction of plant trips and mitigating transient response. .

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