Effects of a Decrease in Feedwater Temperature on Nuclear InstrumentationML031060009 |
Person / Time |
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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 ![Entergy icon.png](/w/images/7/79/Entergy_icon.png) |
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Issue date: |
07/26/1996 |
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From: |
Grimes B K Office of Nuclear Reactor Regulation |
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To: |
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References |
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IN-96-041, NUDOCS 9607220160 |
Download: ML031060009 (10) |
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Similar Documents at 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 |
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Category:NRC Information Notice
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Mclaughlin on NRC, Regarding NRC Information Notice 2006-13: Groundwater Contamination 2020-09-03 The following query condition could not be considered due to this wiki's restrictions on query size or depth: <code> [[:Beaver Valley]] OR [[:Millstone]] OR [[:Hatch]] OR [[:Monticello]] OR [[:Calvert Cliffs]] OR [[:Dresden]] OR [[:Davis Besse]] OR [[:Peach Bottom]] OR [[:Browns Ferry]] OR [[:Salem]] OR [[:Oconee]] OR [[:Mcguire]] OR [[:Nine Mile Point]] OR [[:Palisades]] OR [[:Palo Verde]] OR [[:Perry]] OR [[:Indian Point]] OR [[:Fermi]] OR [[:Kewaunee]] OR [[:Catawba]] OR [[:Harris]] OR [[:Wolf Creek]] OR [[:Saint Lucie]] OR [[:Point Beach]] OR [[:Oyster Creek]] OR [[:Watts Bar]] OR [[:Hope Creek]] OR [[:Grand Gulf]] OR [[:Cooper]] OR [[:Sequoyah]] OR [[:Byron]] OR [[:Pilgrim]] OR [[:Arkansas Nuclear]] OR [[:Three Mile Island]] OR [[:Braidwood]] OR [[:Susquehanna]] OR [[:Summer]] OR [[:Prairie Island]] OR [[:Columbia]] OR [[:Seabrook]] OR [[:Brunswick]] OR [[:Surry]] OR [[:Limerick]] OR [[:North Anna]] OR [[:Turkey Point]] OR [[:River Bend]] OR [[:Vermont Yankee]] OR [[:Crystal River]] OR [[:Haddam Neck]] OR [[:Ginna]] OR [[:Diablo Canyon]] OR [[:Callaway]] OR [[:Vogtle]] OR [[:Waterford]] OR [[:Duane Arnold]] OR [[:Farley]] OR [[:Robinson]] OR [[:Clinton]] OR [[:South Texas]] OR [[:San Onofre]] OR [[:Cook]] OR [[:Comanche Peak]] OR [[:Yankee Rowe]] OR [[:Maine Yankee]] OR [[:Quad Cities]] OR [[:Humboldt Bay]] OR [[:La Crosse]] OR [[:Big Rock Point]] OR [[:Rancho Seco]] OR [[:Zion]] OR [[:Midland]] OR [[:Bellefonte]] OR [[:Fort Calhoun]] OR [[:FitzPatrick]] OR [[:McGuire]] OR [[:LaSalle]] OR [[:Fort Saint Vrain]] OR [[:Shoreham]] OR [[:Satsop]] OR [[:Trojan]] OR [[:Atlantic Nuclear Power Plant]] </code>.
[Table view]The following query condition could not be considered due to this wiki's restrictions on query size or depth: <code> [[:Beaver Valley]] OR [[:Millstone]] OR [[:Hatch]] OR [[:Monticello]] OR [[:Calvert Cliffs]] OR [[:Dresden]] OR [[:Davis Besse]] OR [[:Peach Bottom]] OR [[:Browns Ferry]] OR [[:Salem]] OR [[:Oconee]] OR [[:Mcguire]] OR [[:Nine Mile Point]] OR [[:Palisades]] OR [[:Palo Verde]] OR [[:Perry]] OR [[:Indian Point]] OR [[:Fermi]] OR [[:Kewaunee]] OR [[:Catawba]] OR [[:Harris]] OR [[:Wolf Creek]] OR [[:Saint Lucie]] OR [[:Point Beach]] OR [[:Oyster Creek]] OR [[:Watts Bar]] OR [[:Hope Creek]] OR [[:Grand Gulf]] OR [[:Cooper]] OR [[:Sequoyah]] OR [[:Byron]] OR [[:Pilgrim]] OR [[:Arkansas Nuclear]] OR [[:Three Mile Island]] OR [[:Braidwood]] OR [[:Susquehanna]] OR [[:Summer]] OR [[:Prairie Island]] OR [[:Columbia]] OR [[:Seabrook]] OR [[:Brunswick]] OR [[:Surry]] OR [[:Limerick]] OR [[:North Anna]] OR [[:Turkey Point]] OR [[:River Bend]] OR [[:Vermont Yankee]] OR [[:Crystal River]] OR [[:Haddam Neck]] OR [[:Ginna]] OR [[:Diablo Canyon]] OR [[:Callaway]] OR [[:Vogtle]] OR [[:Waterford]] OR [[:Duane Arnold]] OR [[:Farley]] OR [[:Robinson]] OR [[:Clinton]] OR [[:South Texas]] OR [[:San Onofre]] OR [[:Cook]] OR [[:Comanche Peak]] OR [[:Yankee Rowe]] OR [[:Maine Yankee]] OR [[:Quad Cities]] OR [[:Humboldt Bay]] OR [[:La Crosse]] OR [[:Big Rock Point]] OR [[:Rancho Seco]] OR [[:Zion]] OR [[:Midland]] OR [[:Bellefonte]] OR [[:Fort Calhoun]] OR [[:FitzPatrick]] OR [[:McGuire]] OR [[:LaSalle]] OR [[:Fort Saint Vrain]] OR [[:Shoreham]] OR [[:Satsop]] OR [[:Trojan]] OR [[:Atlantic Nuclear Power Plant]] </code>. |
K) K) 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 9 6 0 7 2 2 0l 6 0 ujo i 7 9,oi4 (R ~IE ctG
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 T Ref. 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 96-09, Supp. 1 96-39 96-38 Deficiencies
in Material Dedication
and Procure-ment Practices
and in Audits of Vendors Damage in Foreign Steam Generator
Internals Estimates
of Decay Heat Using ANS 5.1 Decay Heat Standard May Vary Signi-ficantly Results of Steam Generator Tube Examinations
Inaccurate
Reactor Water Level Indication
and Inad-vertent Draindown
During Shutdown Degradation
of Cooling Water Systems Due to Icing Failure of Safety Systems on Self-Shielded
Irradia-tors Because of Inadequate
Maintenance
and Training Hydrogen Gas Ignition during Closure Welding of a VSC-24 Multi-Assembly
Sealed Basket 07/25/96 07/10/96 07/05/96 06/21/96 06/18/96 06/12/96 06/11/96 05/31/96 All holders of OLs or CPs for nuclear power reactors All holders of OLs or CPs for pressurized-water
reactors All holders of OLs or CPs for nuclear power reactors All holders of OLs or CPs for pressurized
water reactors All pressurized
water reactor facilities
holding an operating
license or a construction
permit All holders of OLs or CPs for nuclear power reactors All U.S. Nuclear Regulatory
Commission
irradiator
licensees
and vendors All holders of OLs or CPs for nuclear power reactors 96-37 96-36 96-35 96-34 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:SRXB I 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
~1~1 -,K)IN 96-XX July XX, 1996 for the loss of extraction
steam to ti cause of the temperature
difference.
occurred between the two heater drain of extraction
steam.he high-pressure
heaters, which was the During the event, a level imbalance tanks, which resulted in the isolation 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* See previous concurrence
OFFICIAL KLLUKV UV X!
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 cause of the temperature
difference.
occurred between the two heater drain of extraction
steam.he high-pressure
heaters, which was the During the event, a level imbalance tanks, which resulted in the isolation 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 OFFICE CONT:jkd _l BC: SRXB E C: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 I oith attachment/enclosure
1
- No 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
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list | - Information Notice 1996-01, Potential For High Post-Accident Closed-Cycle Cooling Water Temperatures to Disable Equipment Important to Safety (3 January 1996)
- Information Notice 1996-01, Potential for High Post-Accident Closed-Cycle Cooling Water Temperatures to Disable Equipment Important to Safety (3 January 1996)
- Information Notice 1996-02, Inoperability of Power-Operated Relief Valves Masked by Downstream Indications During Testing (5 January 1996, Topic: Stroke time)
- Information Notice 1996-03, Main Steam Safety Valve Setpoint Variation as a Result of Thermal Effects (5 January 1996)
- Information Notice 1996-03, Main Steam Safety Valve Setpoint Variation As a Result of Thermal Effects (5 January 1996)
- Information Notice 1996-04, Incident Reporting Requirements for Radiography Licensees (10 January 1996, Topic: Brachytherapy)
- Information Notice 1996-05, Partial Bypass of Shutdown Cooling Flow from Reactor Vessel (18 January 1996, Topic: Reactor Vessel Water Level)
- Information Notice 1996-06, Design & Testing Deficiencies of Tornado Dampers at Nuclear Power Plants (25 January 1996)
- Information Notice 1996-07, Slow Five Percent Scram Insertion Times Caused by Viton Diaphragms in Scram Solenoid Pilot Valves (26 January 1996)
- Information Notice 1996-08, Thermally Induced Pressure Locking of a High Pressure Coolant Injection Gate Valve (5 February 1996, Topic: Anchor Darling, Cold shutdown justification)
- Information Notice 1996-09, Damage in Foreign Steam Generator Internals (12 February 1996, Topic: Earthquake)
- Information Notice 1996-10, Potential Blockage by Debris of Safety System Piping Which Is Not Used During Normal Operation or Tested During Surveillances (13 February 1996)
- Information Notice 1996-10, Potential Blockage by Debris of Safety System Piping Which is Not Used During Normal Operation or Tested During Surveillances (13 February 1996)
- Information Notice 1996-11, Ingress of Demineralizer Resins Increases Potential For Stress Corrosion Cracking of Control Rod Drive Mechanism Penetrations (14 February 1996)
- Information Notice 1996-11, Ingress of Demineralizer Resins Increases Potential for Stress Corrosion Cracking of Control Rod Drive Mechanism Penetrations (14 February 1996)
- Information Notice 1996-12, Control Rod Insertion Problems (15 February 1996)
- Information Notice 1996-13, Potential Containment Leak Paths Through Hydrogen Analysis (26 February 1996)
- Information Notice 1996-14, Degradation of Radwaste Facility Equipment at Millstone Nuclear Power Station, Unit 1 (1 March 1996)
- Information Notice 1996-15, Unexpected Plant Performance During Performance of New Surveillance (8 March 1996)
- Information Notice 1996-16, BWR Operation with Indicated Flow Less than Natural Circulation (14 March 1996)
- Information Notice 1996-17, Reactor Operation Inconsistent with the Updated Final Safety Analysis Report (18 March 1996)
- Information Notice 1996-18, Compliance with 10 CFR Part 20 for Airborne Thorium (25 March 1996, Topic: Brachytherapy)
- Information Notice 1996-19, Failure of Tone Alert Radios to Activate When Receiving a Shortened Activation Signal (2 April 1996)
- Information Notice 1996-20, Demonstration of Associated Equipment Compliance with 10 CFR 34.20 (4 April 1996, Topic: Brachytherapy)
- Information Notice 1996-21, Safety Concerns Related to the Design of the Door Interlock Circuit on Nucletron High-Dose Rate and Pulsed Dose Rate Remote Afterloading Brachytherapy Devices (10 April 1996, Topic: Brachytherapy)
- Information Notice 1996-22, Improper Equipment Settings Due to Use of Nontemperature-Compensated Test Equipment (11 April 1996, Topic: Brachytherapy)
- Information Notice 1996-23, Fires in Emergency Diesel Generator Exciters During Operation Following Undetected Fuse Blowing (22 April 1996, Topic: Brachytherapy)
- Information Notice 1996-24, Preconditioning of Molded-Case Circuit Breakers Before Surveillance Testing (25 April 1996, Topic: Brachytherapy)
- Information Notice 1996-25, Traversing In-Core Probe Overwithdrawn at Lasalle County Station, Unit 1 (30 April 1996, Topic: Brachytherapy)
- Information Notice 1996-26, Recent Problems with Overhead Cranes (30 April 1996, Topic: Brachytherapy)
- Information Notice 1996-26, Recent Problems With Overhead Cranes (30 April 1996)
- Information Notice 1996-27, Potential Clogging of High Pressure Safety Injection Throttle Valves During Recirculation (1 May 1996, Topic: Brachytherapy)
- Information Notice 1996-28, Suggested Guidance Relating to Development and Implementation of Corrective Action (1 May 1996, Topic: Brachytherapy)
- Information Notice 1996-29, Requirements in 10 CFR Part 21 for Reporting and Evaluating Software Errors (20 May 1996, Topic: Brachytherapy)
- Information Notice 1996-30, Inaccuracy of Diagnostic Equipment for Motor-Operated Butterfly Valves (21 May 1996)
- Information Notice 1996-31, Cross-Tied Safety Injection Accumulators (22 May 1996)
- Information Notice 1996-32, Implementation of 10 CFR 50.55a(g) (6) (II) (A), Augmented Examination of Reactor Vessel (5 June 1996, Topic: Non-Destructive Examination)
- Information Notice 1996-32, Implementation of 10 CFR 50.55a(g) (6) (ii) (A), Augmented Examination of Reactor Vessel (5 June 1996, Topic: Nondestructive Examination)
- Information Notice 1996-33, Erroneous Data From Defective Thermocouple Results in a Fire (24 May 1996, Topic: Reverse polarity)
- Information Notice 1996-33, Erroneous Data from Defective Thermocouple Results in a Fire (24 May 1996, Topic: Reverse polarity)
- Information Notice 1996-34, Hydrogen Gas Ignition During Closure Welding of a VSC-24 Multi-Assembly Sealed Basket (31 May 1996)
- Information Notice 1996-35, Failure of Safety Systems on Self-Shielded Irradiators Because of Inadequate Maintenance and Training (11 June 1996)
- Information Notice 1996-36, Degradation of Cooling Water Systems Due to Icing (12 June 1996, Topic: High winds, Ultimate heat sink, Frazil ice)
- Information Notice 1996-37, Inaccurate Reactor Water Level Indication and Inadvertent Draindown During Shutdown (18 June 1996, Topic: Reactor Vessel Water Level)
- Information Notice 1996-38, Results of Steam Generator Tube Examinations (21 June 1996)
- Information Notice 1996-39, Estimates of Decay Heat Using ANS 5.1 Decay Heat Standard May Vary Significantly (5 July 1996)
- Information Notice 1996-40, Defciencies in Material Dedication and Procurement Practices and in Audits of Vendors (7 October 1996, Topic: Coatings, Troxler Moisture Density Gauge)
- Information Notice 1996-41, Effects of a Decrease in Feedwater Temperature on Nuclear Instrumentation (26 July 1996)
- Information Notice 1996-42, Unexpected Opening of Multiple Safety Relief Valves (5 August 1996, Topic: Reactor Vessel Water Level)
- Information Notice 1996-43, Failures of General Electric Magne-Blast Circuit Breakers (2 August 1996)
... further results |
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