Information Notice 1993-62, Thermal Stratification of Water in BWR Reactor Vessels

Thermal Stratification of Water in BWR Reactor Vessels
	ML031070192
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, Fort Saint Vrain, Shoreham, Satsop, Trojan, Atlantic Nuclear Power Plant
Issue date:	08/10/1993
From:	Grimes B; Office of Nuclear Reactor Regulation
To:	;
References
	IN-93-062, NUDOCS 9308030245
	Download: ML031070192 (9)
	v • d • e

UNITED STATES

NUCLEAR REGULATORY COMMISSION

OFFICE OF NUCLEAR REACTOR REGULATION

WASHINGTON, D.C. 20555 August 10, 1993 NRC INFORMATION NOTICE 93-62: THERMAL STRATIFICATION OF WATER IN BWR REACTOR

VESSELS

Addressees

All holders of operating licenses or construction permits for boiling water

reactors.

Purpose

The U.S. Nuclear Regulatory Commission (NRC) is issuing this information

notice to alert addressees that loss of forced circulation through the reactor

vessel coupled with isolation from the main condenser-may allow cold water to

stratify in the bottom of the reactor vessel and cause temperatures to be

lower than allowable. It is expected that recipients will review the

information for applicability to their facilities,and consider actions, as

appropriate, to avoid similar problems. However, suggestions contained in

this information notice are not NRC requirements; therefore, no specific

action or written response is required.

Description of Circumstances

Hatch Unit 1

On August 27, 1992, at Unit 1 of the Hatch Nuclear Power Plant, a high

radiation signal from a main steamline radiation monitor initiated a Group 1 isolation. The main steam isolation valves closed and the.reactor

automatically scrammed from 100-percent power. The resulting low water level

in the reactor vessel caused the recirculation pumps to trip thereby

terminating forced circulation through the reactor vessel. There was no

circulation by the reactor water cleanup system because the licensee had

previously isolated that system for testing. ,The reactor vessel water level

was restored by the steam-driven feedwater pumps and the reactor core

isolation cooling (RCIC) system.

After steam was .no longer available to the feedwater pump turbines, water

level was maintained primarily by the RCIC system. Relatively cool water was

added to the reactor vessel by injection from ,the RCIC system into the

feedwater sparger and from the control rod drive system-into the lower region

of the reactor vessel. Initially, the operators used the temperature of the

coolant in the drain line from the bottom head of the reactor vessel to

monitor the temperature in the reactor vessel. Later, the operators realized

that the temperature in the drain line would not be meaningful because the

reactor water cleanup system had been secured and there was no flow in the

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i, IN 93-62 August 10, 1993 drain line. Because of this, the operators changed'to another point on the

reactor vessel. That location was subsequently determined to be above the

reactor vessel bottom head and not representative of the minimum temperature

in the vessel. A vendor review of the event determined that at one point the

bottom head temperatures were 8VC [15 0F] lower than were allowed by the

pressure-temperature'limits in the technical specifications. ' "

The operators could not restart the reactor recirculation pumps because the

difference in temperature between the reactor dome and the reactor bottom was

greater than the technical specification limit of 62.8 0 C [145 0 F]. After the

reactor was depressurized and while the coolant was still stratified, the

operators started one residual heat removal pump in the shutdown cooling mode.

The temperature at the drain line increased 820C [220'F] in 10 minutes from

its initial 50'C [90'F] temperature; this change exceeded the technical

specification limit of 37.80 C [100 0F] change in one hour.

Peach Bottom Unit 3 On October 15, 1992, at Peach Bottom Unit 3, a half-isolation of the primary

containment'occurred after operators had performed a surveillance-test of

low-pressure switches on the main steamline. While plant personnel were

checking the relays-to determine'the cause of the half-isolation signal, a

second half-isolation signal was received. The main steam isolation valves

closed and the reactor scrammed from 100-percent power. High-pressure coolant

injection and RCIC automatically initiated and, in,conjunction.with the-main.-

safety relief valves, were used to control water level and system pressure.

During recovery from the transient, a second reactor scram from high pressure

occurred. Because of a delay in resetting the first scram and limited flow

through the reactor drain line, thermal stratification of the reactor coolant

occurred in the vessel. The operators did not consider the temperature of the

drain line to be representative of the bottom head temperature because the

reduced flow rate through the drain line caused too low-an indicated -

temperature due to heat losses from the drain line.

The operators could not restart the recirculation pumps because-the

temperature difference between the reactor dome and the drain line was greater

than the 62.80 C [145 0 F]'allowed by the technical specifications for restarting

a recirculation pump. Therefore, the operators proceeded to depressurize and

cool the reactor'before restoring forced circulation. Although other bottom

head metal temperature indications were available, they were not actively

monitored by the operators because of procedural deficiencies and lack of

training. A subsequent review of recorded temperature data determined that

the temperature difference between the reactor dome and the bottom head-drain

line was about 135°C [240'F] and that the pressure-temperature limits for the

reactor vessel bottom head had been violated. With the reactor pressure about

4.14 MPa [600 psig], the'reactor vessel metal temperature had decreased to

about 64°C [115'F], nearly 22°C'[40°F] lower than the pressure-temperature

limit. I I

IN 93-62 August 10, 1993 Discussion

These events demonstrate that isolation of the reactor vessel from the main

condenser with loss of recirculation flow can be initiated by a variety of

causes. Isolation of the reactor vessel causes the operators to take manual

control to restore proper water level and system pressure. A restart of the

recirculation pumps may be delayed because of procedural restrictions.

General Electric has issued a number of communications to licensees regarding

the loss of forced circulation in the reactor vessel. In those

communications, General Electric addressed issues such as potential operating

difficulties, concerns about reactor vessel temperature monitoring, and the

potential for thermal stratification within the reactor vessel. Also, plant

technical specifications contain pressure-temperature limitations to ensure

the integrity and safe operation of the reactor coolant system.

Once thermal stratification occurs, any rapid circulation of water could

result in a large step change in the temperature of the water adjacent to the

reactor bottom head penetrations. This step change may violate the technical

specification limits for rate of temperature change. A temperature

differential within the reactor vessel may be reduced by increasing coolant

flow out of the bottom head drain and reducing cold water flow through the

control rod drive system, which enters the bottom region of the reactor

vessel.

Another concern is maintaining operation within brittle fracture temperature

limits. Once temperature differences develop in the reactor vessel that

restrict restoring forced circulation, operator actions that affect pressure- temperature limits are critical.

Correct and timely operator response to the above conditions depends upon

proper actions being specified in plant procedures and appropriate training

being provided to operators for those actions.

This information notice requires no specific action or written response. If

you have any questions about the information in this notice, please contact

the technical contact listed below or the appropriate Office of Nuclear

Reactor Regulation (NRR) project manager.

Brian K. Grimes, Director

Division of Operating Reactors Support

Office of Nuclear Reactor Regulation

Technical contact: J. Carter, NRR

(301) 504-1153 Attachment:

List of Recently Issued NRC Information Notices

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LIST OF RECENTLY ISSUED

NRC INFORMATION NOTICES

N) oom -4Ia

_m> Information Date of

v> Notice No. Subject Issuance Issued to

4mU)

0 0 0

No 4

93-61 Excessive Reactor Coolant 08/09/93 All holders of OLs or CPs

On Leakage Following A Seal for nuclear power reactor(

0

0o

o en Failure in A Reactor y

Coolant Pump or Reactor

o0 Recirculation Pump

-5 z 93-60 Reporting Fuel Cycle and 08/04/93 All fuel cycle and material

Materials Events to the licensees.

NRC Operations Center

93-59 Unexpected Opening of 07/26/93 All holders of OLs or CPs

Both Doors in An for nuclear power reactors.

Airlock

93-58 Nonconservatism in Low- 07/26/93 All holders of OLs or CPs

Temperature Overpressure for pressurized-water

Protection for Pressurized- reactors.

Water Reactors

93-57 Software Problems 07/23/93 All holders of OLs or CPs

Involving Digital for test and research

Control Console Systems reactors and nuclear power

at Non-Power Reactors reactors.

93-56 Weakness in Emergency 07/22/93 All holders of OLs or CPs -

Operating Procedures for pressurized water

Found as Result of reactors.

Steam Generator Tube

Rupture

93-55 Potential Problem with 07/21/93 All holders of OLs or CPs

Main Steamline Break for pressurized water

Analysis for Main Steam reactors.

Vaults/Tunnels

93-54 Motor-Operated Valve 07/20/93 All holders of OLs or CPs

Actuator Thrust for nuclear power reactors.

Variations Measured

with A Torque Thrust

Cell and A Strain Gage

OL - Operating License

CP - Construction Permit

IN 93-62 V ~August 10, 1993 Discussion