Information Notice 1996-41, Effects of a Decrease in Feedwater Temperature on Nuclear Instrumentation

Effects of a Decrease in Feedwater Temperature on Nuclear Instrumentation

ML031060009
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:	07/26/1996
From:	Grimes B Office of Nuclear Reactor Regulation
To:
References
IN-96-041, NUDOCS 9607220160
Download: ML031060009 (10)

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UNITED STATES

NUCLEAR REGULATORY COMMISSION

OFFICE OF NUCLEAR REACTOR REGULATION

WASHINGTON, D.C. 20555-0001 July 26, 1996 NRC INFORMATION NOTICE 96-41: EFFECTS OF A DECREASE IN FEEDWATER TEMPERATURE

ON NUCLEAR INSTRUMENTATION

Addressees

All holders of operating licenses or construction permits for pressurized

water reactors (PWRs).

Purpose

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

notice to alert addressees to the potential for operation above licensed power

as a result of a decrease in feedwater temperature event affecting nuclear

instrumentation. 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

On February 14, 1996, the licensee for the Comanche Peak Steam Electric

Station was operating Unit 2 at 95 percent rated thermal power near end-of- core life when a significant reduction in feedwater temperature occurred

because of the loss of feedwater heaters. This reduction, in turn, caused a

reduction in the reactor coolant system cold-leg temperatures. The colder

reactor coolant temperature, with a large negative moderator temperature

coefficient, caused reactor power to increase to approximately 102 percent

according to ex-core nuclear instrumentation. The nitrogen-16 (N-16)

detection system reached the overpower turbine runback setpoint (109 percent)

and initiated a turbine runback. The N-16 detection system measures N-16 activity in the primary coolant as a measure of the total power generation.

This system is a substitute for the resistance temperature detector over- temperature and over-power reactor trip functions used at other Westinghouse

PWRs. The plant stabil zed at an indicated power of approximately 97 percent

according to the ex-core nuclear instrumentation.

After approximately 90 minutes, a second similar turbine runback occurred

while restoring balance-of-plant equipment. Following this runback, reactor

power was stabilized at approximately 100 percent according to nuclear

instrumentation. During the next 30 minutes, the reactor was operated at

approximately 100 percent power as indicated by nuclear instrumentation, with

reactor coolant temperatures below normal. The licensee noted that the N-16

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IN 96-41 July 26, 1996 detection system indicated approximately 106 percent power and the computer- based plant calorimetric system indicated approximately 102 percent power.

Subsequently, the reactor power was reduced to less than 100 percent by all

indications.

Discussion

There are three aspects of this event which have generic implications. First, with a loss of secondary plant efficiency, programmed T e can no longer

reliably represent core thermal power. Second, the venturi-based input into

the computer-based calorimetric system may not be accurate with cold

feedwater. And third, the final safety analysis report had not analyzed this

transient accurately.

Following the second runback, operators noted that reactor power indicated

<100 percent according to nuclear instrumentation. Although the operators

knew that cold feedwater could cause an increase in the amount of neutron

attenuation, they believed that the nuclear instrumentation indicated

conservatively (i.e., higher than actual) because they were maintaining TA"e

approximately 1.7 eC [3 OF] above TRef. The licensee could not use the

computer-based calorimetric until some time after the second turbine runback

due to maintenance activities. Te , based on the main turbine impulse

pressure, is programmed as a functlon of turbine load and, for normal

efficiency, is a good representation of thermal power. When the unit lost the

feedwater heaters, the plant efficiency decreased. Because the main turbine

electro-hydraulic control system maintained generator output, core thermal

power increased to account for the loss of efficiency, and thus, TRef no

longer accurately represented the core thermal power.

The cold-leg temperature is a more appropriate indicator of the accuracy of

the nuclear instrumentation than programmed TY.e. As the cold-leg temperature

decreased, the amount of neutron attenuation in the downcomer area surrounding

the core increased and hence affected the amount of neutrons reaching the

detectors. The licensee analysis showed that for every 0.6 C (1 OF] of cold- leg temperature change, the nuclear instrumentation was affected by 0.6 to 0.8 percent power. A review of the second transient showed that the cold-leg

temperature was approximately 2.5 °C [4.5 OF] lower than when the detectors

were last calibrated. This corresponded to a 3 to 4 percent error, which

corresponded to the difference in the actual versus the indicated power (104 percent actual versus 100 percent indicated).

During the review, the licensee noted that the computer-based calorimetric was

4 percent lower than the actual thermal power (N-16 power monitor). The

calorimetric was based on feedwater flow measured by venturis. Although the

calorimetric calculation used feedwater temperature as an input, temperatures

significantly different than the normal 227 OC [440 OF] introduced errors into

the calculation.

Finally, the actual events involved temperature and power levels that exceeded

those in the analysis of the Decrease in Feedwater Temperature" event

presented in Chapter 15 of the licensee final safety analysis report. In that

IN 96-41 July 26, 1996 analysis, the inadvertent opening of the low-pressure heater bypass valve, coupled with the trip of the heater drain pumps, resulted in a feedwater

temperature drop of less than 19 'C (35 OF], and a corresponding power

increase of less than 10 percent. In the actual event, the feedwater

temperature dropped by approximately 111 °C (200 OF], and the licensee

calculated that reactor power would have increased by approximately 35 percent

without operator or protective actions. The licensee determined that although

the initiating events were the same, the Chapter 15 analysis did not account

for the loss of extraction steam to the high-pressure heaters, which was the

cause of the temperature difference. During the event, a level imbalance

occurred between the two heater drain tanks, which resulted in the isolation

of extraction steam.

The NRC staff review of analyses of feedwater temperature events at similar

facilities revealed that most of these analyses assumed similar initiating

events as the Comanche Peak analysis and had similar conclusions concerning

the amount of feedwater temperature drop. The licensee has reanalyzed the

event to include a 119 OC [246 OF] feedwater temperature drop and concluded

that all accident analysis parameters remained within requirements.

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

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

one of the technical contacts listed below or the appropriate Office of

Nuclear Reactor Regulation project manager.

Brian K. Grimes, Acting Director

Division of Reactor Program Management

Office of Nuclear Reactor Regulation

Technical contacts: Harry A. Freeman, RIV

(817) 897-1500

E-mail: haf~nrc.gov

Chu-Yu Liang, NRR

(301) 415-2878 E-mail: cylenrc.gov

Attachment: List Of Recently Issued HRC Information Notices

A1h4 Stir A Je6tQ

K> KJ

Attachment

IN 96-41 July 26, 1996 LIST OF RECENTLY ISSUED

NRC INFORMATION NOTICES

Information Date of

Notice No. Subject Issuance Issued to

96-40 Deficiencies in Material 07/25/96 All holders of OLs or CPs

Dedication and Procure- for nuclear power reactors

ment Practices and in

Audits of Vendors

96-09, Damage in Foreign Steam 07/10/96 All holders of OLs or CPs

Supp. 1 Generator Internals for pressurized-water

reactors

96-39 Estimates of Decay Heat 07/05/96 All holders of OLs or CPs

Using ANS 5.1 Decay Heat for nuclear power reactors

Standard May Vary Signi- ficantly

96-38 Results of Steam Generator 06/21/96 All holders of OLs or CPs

Tube Examinations for pressurized water

reactors

96-37 Inaccurate Reactor Water 06/18/96 All pressurized water

Level Indication and Inad- reactor facilities holding

vertent Draindown During an operating license or a

Shutdown construction permit

96-36 Degradation of Cooling 06/12/96 All holders of OLs or CPs

Water Systems Due to Icing for nuclear power reactors

96-35 Failure of Safety Systems 06/11/96 All U.S. Nuclear Regulatory

on Self-Shielded Irradia- Commission irradiator

tors Because of Inadequate licensees and vendors

Maintenance and Training

96-34 Hydrogen Gas Ignition 05/31/96 All holders of OLs or CPs

during Closure Welding for nuclear power reactors

of a VSC-24 Multi-Assembly

Sealed Basket

OL - Operating License

CP - Construction Permit

• ~ - K> K

IN 96-41 July 26, 1996 analysis, the inadvertent opening of the low-pressure heater bypass valve, coupled with the trip of the heater drain pumps, resulted in a feedwater

temperature drop of less than 19 *C [35 OF], and a corresponding power

increase of less than 10 percent. In the actual event, the feedwater

temperature dropped by approximately 111 *C [200 OF], and the licensee

calculated that reactor power would have increased by approximately 35 percent

without operator or protective actions. The licensee determined that although

the initiating events were the same, the Chapter 15 analysis did not account

for the loss of extraction steam to the high-pressure heaters, which was the

cause of the temperature difference. During the event, a level imbalance

occurred between the two heater drain tanks, which resulted in the isolation

of extraction steam.

The NRC staff review of analyses of feedwater temperature events at similar

facilities revealed that most of these analyses assumed similar initiating

events as the Comanche Peak analysis and had similar conclusions concerning

the amount of feedwater temperature drop. The licensee has reanalyzed the

event to include a 119 *C [246 OF] feedwater temperature drop and concluded

that all accident analysis parameters remained within requirements.

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

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

one of the technical contacts listed below or the appropriate Office of

Nuclear Reactor Regulation project manager.

Original signed by Brian K.Grimes

Brian K. Grimes, Acting Director

Division of Reactor Program Management

Office of Nuclear Reactor Regulation

Technical contacts: Harry A. Freeman, RIV

(817) 897-1500

E-mail: haf@nrc.gov

Chu-Yu Liang, NRR

(301) 415-2878 E-mail: cyl~nrc.gov

Attachment: List of Recently Issued NRC Information Notices

DOCUMENT NAME: G:\SSK2\INFONOT.C P

To receive a copy of this docunent, tndicate in the box CO~opy So attachment/enclosure EsCopy with attachment/enctosure N

• No

cops

OFFICE C BC:SRXBI BC:LPECB lI (A)DW M i

NAME CYLiang* RJones* AChaffee*

HAFreeman* ____ _

DATE 16/ 3/96 16/21/96 17/08/96 17LI/96 I

OFFILIAL MLLUM LWUF

• See previous concurrence Tech Editor reviewed & concurred on 05/28/96

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IN 96-XX

July XX, 1996 for the loss of extraction steam to ti he high-pressure heaters, which was the

cause of the temperature difference. During the event, a level imbalance

occurred between the two heater drain tanks, which resulted in the isolation

of extraction steam.

The NRC staff review of analyses of feedwater temperature events at similar

facilities revealed that most of these analyses assumed similar initiating

events as the Comanche Peak analysis and had similar conclusions concerning

the amount of feedwater temperature drop. The licensee has reanalyzed the

event to include a 119 'C [246 'F] feedwater temperature drop and concluded

that all accident analysis parameters remained within requirements.

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

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

one of the technical contacts listed below or the appropriate Office of

Nuclear Reactor Regulation project manager.

Brian K. Grimes, Acting Director

Division of Reactor Program Management

Office of Nuclear Reactor Regulation

Technical contacts: Harry A. Freeman, RIV

(817) 897-1500

E-mail: haf~nrc.gov

Chu-Yu Liang, NRR

(301) 415-2878 E-mail: cyl~nrc.gov

Attachment: List of Recently Issued NRC Information Notices

DOCUMENT NAME: G:\SSK2\INFONOT.C P

To receive a copy of this document, indicate in the box CzAopy w/o attachment/enclosure E-Copy with attachment/enclosure N

• No

OFFICE l kd BC: SRXB BC:PECB )D:DR

NAME CYLiang* RJones* AChaffee* BGrimes

HAFreeman*

DATE 6/ 3/96 6/21/96 7/08/96 7/ /96 OFFICIAL RECORD COPY

• See previous concurrence

IN 96-XX

July XX, 1996 for the loss of extraction steam to the high-pressure heaters, which was the

cause of the temperature difference. During the event, a level imbalance

occurred between the two heater drain tanks, which resulted in the isolation

of extraction steam.

The NRC staff review of analyses of feedwater temperature events at similar

facilities revealed that most of these analyses assumed similar initiating

events as the licensee analysis and had similar conclusions concerning the

amount of feedwater temperature drop. The licensee has reanalyzed the event

pursuant to Section 50.59 of Title 10 of the Code of Federal Regulations to

include a 119 'c [246 OF] feedwater temperature drop and concluded that all

accident analysis parameters remained within requirements.

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

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

one of the technical contacts listed below or the appropriate Office of

Nuclear Reactor Regulation project manager.

Brian K. Grimes, Acting Director

Division of Reactor Program Management

Office of Nuclear Reactor Regulation

Technical contacts: Harry A. Freeman, RIV

(817) 897-1500

E-mail: haftnrc.gov

Chu-Yu Liang, NRR

(301) 415-2878 E-mail: cyl~nrc.gov

Attachment: List of Recently Issued NRC Information Notices

DOCUMENT NAME: G:\SSK2\INFONOT.C P

To receive a copy of this document, indicate in the box C-Topy u/o attachment/enclosure E=Copy with attachment/enclosure N No

copy

OFFICE CONT:i kd l BC:SRXBLl BC:iPECB lI (A)iD:iDRPM I _

NAME CYLiang* RJones* AChaffee* BGrimes

l _ HAFreeman*

DATE 6/ 3/96 6/21/96 7/08/96 7/ /96 OFFICIAL KLLUKV UV X!

• See previous concurrence

IN 96-XX

July XX, 1996 for the loss of extraction steam to the high-pressure heaters, which was the

cause of the temperature difference. During the event, a level imbalance

occurred between the two heater drain tanks, which resulted in the isolation

of extraction steam.

The NRC staff review of analyses of feedwater temperature events at similar

facilities revealed that most of these analyses assumed similar initiating

events as the licensee analysis and had similar conclusions concerning the

amount of feedwater temperature drop. The licensee has reanalyzed the event

pursuant to Section 50.59 of Title 10 of the Code of Federal Regulations to

include a 119 *C [246 *F] feedwater temperature drop and concluded that all

accident analysis parameters remained within requirements.

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

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

one of the technical contacts listed below or the appropriate Office of

Nuclear Reactor Regulation project manager.

Brian K. Grimes, Acting Director

Division of Reactor Program Management

Office of Nuclear Reactor Regulation

Technical contacts: Harry A. Freeman, RIV

(817) 897-1500

Internet:haf@nrc.gov

Chu-Yu Liang, NRR

(301) 415-2878 Internet:cyl nrc.gov

Attachment: List of Recently Issued NRC Information Notices

DOCUMENT NAME: G:\SSK2\INFONOT.C P

To receive a copy of this document, Indicate in the box Conopy w/c attachment/enclosure EnCopy with attachment/enclosure N

• No

OFFICE CONT: Ekd BC: SLB BC:PECB (A)D:DRPM

NAME CYLiang* RJones* ACh)f BGrimes

l ~~HAFreeman*tVt

DATE 6/ 3/96 6/21/96 7/7/96 7/ /96 OFFICIAL RECOR COPY

• See previous concurrence

K-, /

IN 96-XX

June XX, 1996 for the loss of extraction steam to ti he high-pressure heaters, which was the

cause of the temperature difference. During the event, a level imbalance

occurred between the two heater drain tanks, which resulted inthe isolation

of extraction steam.

The NRC staff's review of analyses of feedwater temperature events at similar

facilities revealed that most of these analyses assumed similar initiating

events as the licensee's analysis and had similar conclusions concerning the

amount of feedwater temperature drop. The licensee has reanalyzed the event

pursuant to Section 50.59 of Title 10 of the Code of Federal Regulations to

include a 119 'C [246 OF] feedwater temperature drop and concluded that all

accident analysis parameters remained within requirements.

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

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

one of the technical contacts listed below or the appropriate Office of

Nuclear Reactor Regulation project manager.

Brian K. Grimes, Acting Director

Division of Reactor Program Management

Office of Nuclear Reactor Regulation

Technical contacts: Harry A. Freeman, RIV

(817) 897-1500

Internet:haffnrc.gov

Chu-Yu Liang, NRR

(301) 415-2878 Internet:cyl@nrc.gov

Attachment: List of Recently Issued NRC Information Notices

DOCUMENT NAME: G:\SSK2\INFONOT.CP

To receive a copy of this document, indicate in the box Ciropy w/dattachmeft1/enctosure EnC Ioith attachment/enclosure 1

• No copy

OFFICE CONT:jkd _l BC: SRXB EC:PECB I _ A)D:DRPM I

NAME CYLiang* RJones AChaffee BGrimes

HAFreeman* I- _

DATE 6/ 3/96 6/2j /96 6/ /96 6/ /96 OFFICIAL RECORD COPY

• See previous concurrence

IN 96-XX

June XX, 1996 detection system. The licensee believed that this system would probably not

be significantly affected by feedwater temperatures because of a different

mass flow rate determination method.

Finally, the licensee's final safety analysis report did not accurately

analyze this transient. The actual events were similar to the analysis of the

'Decrease in Feedwater Temperature event presented in Chapter 15. In that

analysis, the inadvertent opening of the low-pressure heater bypass valve, coupled with the trip of the heater drain pumps, resulted in a feedwater

temperature drop of less than 35 OF, and a corresponding power increase of

less than 10 percent. In the actual event, the feedwater temperature dropped

by approximately 200 OF, and the licensee calculated that reactor power would

have increased by approximately 35 percent without operator or protective

actions. The licensee determined that although the initiating events were the

same, the Chapter 15 analysis did not account for the loss of extraction steam

to the high-pressure heaters, which was the cause of the temperature

difference. During the event, a level imbalance occurred between the two

heater drain tanks, which resulted in the isolation of extraction steam.

The NRC staff's review of analyses of feedwater temperature events at similar

facilities revealed that most of these analyses assumed similar initiating

events as the licensee's analysis and had similar conclusions concerning the

amount of feedwater temperature drop.

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

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

one of the technical contacts listed below or the appropriate Office of

Nuclear Reactor Regulation project manager.

Brian K. Grimes, Acting Director

Division of Reactor Program Management

Office of Nuclear Reactor Regulation

Technical contacts: Harry A. Freeman, RIV

(817) 897-1500

Internet:haf@nrc.gov

Chu-Yu Liang, NRR

(301) 415-2878 Internet:cyl nrc.gov

Attachment: List of Recently Issued NRC Information Notices

DOCUMENT NAME: G:\SSK2\INFONOT.C P

To receive a copy of this docunent, indicate in the box Catopy w/o attachment/enclosure E-C with attachment/enclosure N

• No copy

OFFICE lCONT:kd l BC:SRXB l BC:PECB l (A)D:DRPM

NAME CYLiang 9 RJones AChaffee BGrimes

HAFreema r _ _

DATE /96 /96 6/ /96 6/ /96 OFFICIAL RECORD COPY