Information Notice 1986-04, Transient Due to Loss of Power to Intergrated Control System at a Pressurized Water Reactor Designed by Babcock & Wilcox

Transient Due to Loss of Power to Intergrated Control System at a Pressurized Water Reactor Designed by Babcock & Wilcox

ML031220490
Person / Time
Site:	Beaver Valley, Millstone, Hatch, Monticello, Calvert Cliffs, Dresden, Davis Besse, Peach Bottom, Browns Ferry, Salem, Oconee, Mcguire, Nine Mile Point, Palisades, Palo Verde, Perry, Indian Point, Fermi, Kewaunee, Catawba, Harris, Wolf Creek, Saint Lucie, Point Beach, Oyster Creek, Watts Bar, Hope Creek, Grand Gulf, Cooper, Sequoyah, Byron, Pilgrim, Arkansas Nuclear, Three Mile Island, Braidwood, Susquehanna, Summer, Prairie Island, Columbia, Seabrook, Brunswick, Surry, Limerick, North Anna, Turkey Point, River Bend, Vermont Yankee, Crystal River, Haddam Neck, Ginna, Diablo Canyon, Callaway, Vogtle, Waterford, Duane Arnold, Farley, Robinson, Clinton, South Texas, San Onofre, Cook, Comanche Peak, Yankee Rowe, Maine Yankee, Quad Cities, Humboldt Bay, La Crosse, Big Rock Point, Rancho Seco, Zion, Midland, Bellefonte, Fort Calhoun, FitzPatrick, McGuire, LaSalle, 05000000, Zimmer, Fort Saint Vrain, Shoreham, Satsop, Trojan, Atlantic Nuclear Power Plant, Skagit, Marble Hill
Issue date:	01/31/1986
From:	Jordan E NRC/IE
To:
References
IN-86-004, NUDOCS 8601290048
Download: ML031220490 (4)

SSINS No.: 6835 IN 86-04 UNITED STATES

NUCLEAR REGULATORY COMMISSION

OFFICE OF INSPECTION AND ENFORCEMENT

WASHINGTON, DC 20555 January 31, 1986 IE INFQRMATION NOTICE NO. 86-04: TRANSIENT DUE TO LOSS OF POWER TO INTEGRATED

CONTROL SYSTEM AT A PRESSURIZED WATER REACTOR

DESIGNED BY BABCOCK & WILCOX

Addressees

All nuclear power facilities holding an operating license (OL) or a construction

permit (CP).

Purpose

This notice is to inform recipients of a recent event at an operating pressurized

water reactor resulting from loss of power to the integrated control system.

Recipients are expected to review the information in this notice for applicability

to their facilities and consider actions, if appropriate, to preclude similar

problems from occurring at their facilities. However, suggestions contained in

this notice,4o not constitute NRC requirements; therefore, no specific action

or written response is required.

Description of Circumstances

On December 26, 1985, Rancho Seco was operating on automatic control at a

constant power level of 710 MWe (76% of licensed power). At 4:14 a.m., power

to the integrated control system (ICS) was lost. The annunciator alarm for

"Loss of ICS or Fan Power" sounded. As designed, ICS demand signals went to

midscale. The main feedwater valves closed to 50%, and the atmospheric dump

valves, turbine bypass valves, and one set of auxiliary feedwater valves opened

to 50%. The main feedwater pump speed was reduced to minimum. Low discharge

pressure at the main feedwater pump caused the motor-driven auxiliary feedwater

pump to start automatically. The net decrease in feedwater flow caused the

reactor to trip on high reactor coolant system (RCS) pressure.

After the reactor trip, the above ICS valves remained at 50% (i.e., could not

be operated from the control room) causing excessive cooling of the RCS which

was exacerbated by autostarting of the dual-drive auxiliary feedwater pump.

During-the 26 minutes required to restore ICS power, operators acted to mini- mize the resulting transient. However, difficulties were experienced with

manipulation of valves, operation of pumps, and control of various liquid

levels, pressures, and temperatures. RCS pressure decreased to a minimum of

1064 psig at 4:21 a.m. At 4:40 a.m., the lowest RCS temperature (3860 F) during

the cooling transient was reached. RCS pressure at that time was 1413 psig.

Eventually, a senior reactor operator discovered that switches which supplied

8601290048

IN 86-04 January 31, 1986 power to the ICS dc power supplies were in the off position and set them to

the on position. Although manual (i.e., hand) operation was now possible in

the control room, the valves initially received a 100% demand signal. Opera- tors quickly shut the valves. At 5:00 a.m., RCS pressure and temperature

were stabilized at 716 psig and 4330 F and maintained there for 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />. This

unusual event, which was declared at 4:30 a.m., was terminated at 8:41 a.m.

Before the event was terminated, a large number of problems were experienced, including:

t The RCS was cooled 180'F in 24 minutes violating the technical

specifications limit of 100'F in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

Recommended pressure/temperature limits for pressurized thermal shock

were exceeded; however, nil ductility temperature limit in the

technical specifications was not violated.

• Pressurizer level was low and off scale.

• After loss of ICS power, ICS controlled valves could not be manually

operated from the control room.

• One auxiliary feedwater isolation valve could not be closed.

• One auxiliary feedwater flow control valve was overtorqued using

the manual handwheel, and the manual operator failed.

• Operators had considerable difficulty determining (locally) the

position of the auxiliary feedwater flow control valves.

• One steam generator was overfilled.

t A main feedwater flow recorder in the control room failed at midscale

because of the loss of ICS power although main fee~dwater flow was

essentially zero.

• An RCS makeup pump was run without water (i.e., suction valve shut)

and severely damaged, specifically, seals for the makeup pump failed

and approximately 450 gallons of water were spilled in the auxiliary

building.

^ A containment radiation monitor was damaged because it continued to

run after the suction valve had been shut by a Safety Features

Actuation Signal.

Four senior reactor operators were present during the event. At 5:01 a.m., one

of them collapsed from exhaustion in front of a control panel. He was trans- ported by ambulance to a loc'al hospital and subsequently released in satisfactory

condition at 7:00 a.m.

IN 86-04 January 31, 1986 Discussion:

The NRC sent an incident investigation team (IIT) to Rancho Seco shortly after

the event. The licensee has agreed to hold in abeyance any work in progress

or planned (except as required by plant safety considerations) until the

licensee and the NRC have had an opportunity to develop detailed trouble- shooting plans for failed equipment. Further, the licensee has agreed to

maintain the unit in a shutdown mode until NRC concurs with the licensee that

the unit can be returned to power safely.

Review by the HIT is continuing. As additional information about the event is

obtained, this notice will be supplemented, if appropriate.

No specific action or written response is required by this information notice.

If you have any questions about this matter, please contact the Regional

Administrator of the appropriate NRC regional office or this office.

ard . r

Divisi n f Emergency Preparedness

and gineering Response

Office of Inspection and Enforcement

R. W. Woodruff, IE

(301) 492-8597 Attachment: List of Recently Issued IE Information Notices

Attachment 1 IN 86-04 January 31, 1986 LIST OF RECENTLY ISSUED

IE INFORMATION NOTICES

Information Date of

Notice No. Subject Issue Issued to

86-03 Potential Deficiencies In 1/14/86 All power reactor

Environmental Qualification facilities holding

Of Limitorque Motor Valve an OL or CP

Operator Wiring

86-02 Failure Of Valve Operator 1/6/86 All power reactor

Motor During Environmental facilities holding

Qualification Testing an OL or CP

86-01 Failure Of Main Feedwater 1/6/86 All power reactor

Check Valve Causes Loss Of facilities holding

Feedwater System Integrity an OL or CP

And Water-Hammer Damage

85- 101 Applicability of 10 CFR 21 12/31/85 All power reactor

To Consulting Firms Providing facilities holding

Training an OL or CP

85-100 Rosemount Differential 12/31/85 All power reactor

Pressure Transmitter Zero facilities holding

Point Shift an OL or CP

85-99 Cracking In Boiling-Water- 12/31/85 All BWR facilities

Reactor Mark I And Mark II having a Mark I or

Containments Caused By Failure Mark II containment

Of The Inerting System

85-98 Missing Jumpers From Westing- 12/26/85 All Westinghouse

house Reactor Protection designed PWR

System Cards For The Over- facilities holding

Power Delta Temperature Trip an OL or CP

Function

85-97 Jail Term For Former 12/26/85 All power reactor

Contractor Employee Who facilities holding

Intent

ionally Falsified an OL or CP

Welding Inspection Records

OL = Operating License

CP = Construction Permit