Information Notice 1985-72, Uncontrolled Leakage of Reactor Coolant Outside Containment

Uncontrolled Leakage of Reactor Coolant Outside Containment

ML031180635
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, 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, Crane
Issue date:	08/22/1985
From:	Jordan E NRC/IE
To:
References
IN-85-072, NUDOCS 8508200630
Download: ML031180635 (4)

SSINS NO.:

6835 IN 85-72

UNITED STATES

NUCLEAR REGULATORY COMMISSION

OFFICE OF INSPECTION AND ENFORCEMENT

WASHINGTON, D.C.

20555

August 22, 1985

IE INFORMATION NOTICE NO. 85-72:

UNCONTROLLED LEAKAGE OF REACTOR COOLANT

OUTSIDE CONTAINMENT

Addressees

All boiling water reactors holding an operating license (OL) or a construction

permit (CP).

Purpose

This information notice is provided to alert recipients of a significant event

involving an uncontrolled primary coolant leak outside containment.

It is

expected that recipients will review the information for applicability to

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

events from occurring at their facilities.

However, suggestions contained in

this information notice do not constitute requirements; therefore, no specific

action or written response is required.

Description of Circumstances

On June 12, 1985, a reactor scram occurred from 99% power at Oyster Creek

Nuclear Generating Station.

The scram occurred following failure of the elec- tric pressure regulator that subsequently caused a turbine bypass valve to

open.

This resulted in a reactor pressure decrease to the low pressure trip

set point, causing the main steam isolation valves (MSIVs) to shut and the

reactor to scram.

As part of the scram sequence, the scram discharge volume (SDV) vent and drain

valves are required to shut to contain the water released during a scram.

However, in this event, the two drain valves did not fully close, allowing

hot, reactor coolant to drain to the reactor building equipment drain tank.

The hot fluid flashed in the drain system creating steam that flowed up through

various drains in the 51-ft and 23-ft building levels. The steam combined with

the fumes from the blistering paint on the SDV drain piping and caused a por- tion of the reactor building deluge fire protection system to actuate and spray

down the 51-ft level.

Approximately 500 gal of reactor coolant flowed to the

drain tank before the scram system could be reset, which took approximately 38 minutes.

The fire protection deluge system actuated approximately 20 minutes

after the scram and was shut off in approximately 5 minutes.

No safety equip- ment inside the reactor building was adversely affected by actuation of the

deluge system.

8508200630

IN 85-72 August 22, 1985 Discussion:

The failure of the SDV drain valves to properly close caused the following:

1. Uncontrolled reactor coolant leakage outside containment.

2. Temperatures of the control rod drive (CRD) seals exceeding the alarm

setting.

3. Actuation of the reactor building fire protection deluge system.

4. Radioactive contamination of the 23-ft level of the reactor building.

Each SDV drain valve that failed had a different failure mechanism. The

upstream valve stem/disc travel stopped approximately 1/8 inch before fully

seating onto the valve seat.

This was caused by the valve actuator not having

the stroking length properly adjusted. The downstream valve had an improperly

sized spring in the valve actuator. It is believed that the valve initially

closed, but was then forced open when the system pressure exerted a force

below the valve seat that exceeded the spring closing force of the actuator.

The high CR0 seal temperature alarms were received intermittently after the

scram.

The alarms are an indication of abnormal flow of reactor coolant within

or out of the CR0 system.

Degradation, or possibly failure, of the seals could

occur following prlonged exposure at elevated temperatures. As a result,'

abnormal leakage could occur that might adversely affect proper rod motion or

rod scramming ability.

The reactor building fire protection system actuated on the 51-ft level of the

reactor building. Although no equipment was adversely affected by the deluge

system spray the potential existed for damaging electrical equipment and pos- sibly aggravating an already serious problem.

Although the SDV vent and drain valves are stroke tested monthly in accordance

with the inservice testing (IST) program, there were no criteria or requirements, for leak testing these valves.

Following the initial installation of the down- stream valve as part of a system backfit in 1984, no postinstallation leak rate

test of either valve against operating pressure was conducted.

Both valve

problems could have been detected by such a test.

IE Information Notice 84-35, "BWR Post-Scram Drywell Pressurization" described

an event of August 1982 at the Hatch Nuclear Plant Unit 2 where there was a

similar leakage from the SDV.

That event was also the subject of an AEOD case

study and was included in the 3rd quarter, 1983, "Report to Congress on

Abnormal Occurrences."

IN 85-72 August 22, 1985 No specific action or written response is required by this information notice.

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

Administrator of the appropriate NRC regional office or this office.

L/ S

irtor

Divisi

of Emergency Preparedness

.

and Engineering Response

David Powell, IE

(301) 492-8373 Attachment:

Attachment 1

IN 85-72

August 22, 1985

LIST OF RECENTLY ISSUED

IE INFORMATION NOTICES

Information

Date of

Notice No.

Subject

Issue

Issued to

85-71 Containment

Rate Tests

Integrated Leak

8/22/85

All power reactor

facilities holding

an OL or CP

85-70

85-69

85-68

85-42 Rev. 1

85-67

85-66

85-65

Teletherapy Unit Full

Calibration And Qualified

Expert Requirements (10 CFR

35.23 And 10 CFR 35.24)

Recent Felony Conviction For

Cheating On Reactor Operator

Requalification Tests

Diesel Generator Failure At

Calvert Cliffs Nuclear

Station Unit 1

Loose Phosphor In Panasonic

800 Series Badge Thermo- luminescent Dosimeter (TLD)

Elements

Valve-Shaft-To-Actuator Key

May Fall Out Of Place When

Mounted Below Horizontal Axis

Discrepancies Between

As-Built Construction

Drawings And Equipment

Installations

Crack Growth In Steam

Generator Girth Welds

BBC Brown Boveri Low-Voltage

K-Line Circuit Breakers, With

Deficient Overcurrent Trip

Devices Models OD-4 and 5

8/15/85

8/15/85

8/14/85

8/12/85

8/8/85

8/7/85

7/31/85

7/26/85

All material

licensees

All power reactor

facilities holding

an OL or CP

All power reactor

facilities holding

an OL or CP

Materials and fuel

cycle licensees

All power reactor

facilities holding

an OL or CP

All power reactor

facilities holding

an OL or CP

All PWR facilities

holding an OL or CP

All power reactor

facilities holding

an OL or CP

85-64 OL = Operating License

CP = Construction Permit