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{{#Wiki_filter:SSINS No: 6835IN 85-23UNITED STATESNUCLEAR REGULATORY COMMISSIONOFFICE OF INSPECTION AND ENFORCEMENTWASHINGTON, D.C. 20555March 22, 1985IE INFORMATION NOTICE NO. 85-23: INADEQUATE SURVEILLANCE AND POSTMAINTENANCEAND POSTMODIFICATION SYSTEM TESTING
{{#Wiki_filter:SSINS No: 6835 IN 85-23 UNITED STATES NUCLEAR REGULATORY
 
COMMISSION
 
OFFICE OF INSPECTION
 
===AND ENFORCEMENT===
WASHINGTON, D.C. 20555 March 22, 1985 IE INFORMATION
 
NOTICE NO. 85-23: INADEQUATE
 
SURVEILLANCE
 
===AND POSTMAINTENANCE===
AND POSTMODIFICATION
 
SYSTEM TESTING


==Addressees==
==Addressees==
:All nuclear power reactor facilities holding an operating license (OL) or aconstruction permit (CP).
:
All nuclear power reactor facilities
 
holding an operating
 
license (OL) or a construction
 
permit (CP).


==Purpose==
==Purpose==
:This information notice is to alert addressees of several instances pertainingto improper system modifications, inadequate postmodification system testing,and inadequate surveillance testing recently detected at the McGuire nuclearpower facility.It is expected that recipients will review the information contained in thisnotice for applicability to their facilities and consider actions, if appropri-ate, to preclude similar problems from occurring at their facilities. However,suggestions contained in this notice do not constitute NRC requirements; there-fore, no specific action or written response is required.
: This information
 
notice is to alert addressees
 
of several instances
 
pertaining
 
to improper system modifications, inadequate
 
postmodification
 
system testing, and inadequate
 
surveillance
 
testing recently detected at the McGuire nuclear power facility.It is expected that recipients
 
will review the information
 
contained
 
in this notice for applicability
 
to their facilities
 
and consider actions, if appropri-ate, to preclude similar problems from occurring
 
at their facilities.
 
However, suggestions
 
contained
 
in this notice do not constitute
 
NRC requirements;  
there-fore, no specific action or written response is required.Description
 
of Circumstances:
On November 1, 1984, Duke Power Company (DPC) informed the NRC that the four Rosemont differential
 
pressure transmitters
 
that control the closing of four isolation
 
valves of the upper-head
 
injection (UHI) system at McGuire Unit 1 were improperly
 
installed (i.e., the impulse lines were reversed when the original Barton reverse-acting
 
differential
 
pressure switches were replaced with Rosemont direct-acting
 
differential
 
pressure transmitters
 
during April of 1984). As a result, the UHI isolation
 
valves failed to close during draining of the accumulator
 
when the water level in the UHI accumulator
 
reached the-set point. In addition to the improper installation, the postmodification
 
testing was limited to a dry calibration
 
method that does not use the actual reference leg of the accumulator;
therefore, the installation
 
error was not detected by the postmodification
 
test. Consequently, the plant was operated for approxi-mately five months with the UHI isolation
 
valves inoperable.
 
The McGuire UHI system design includes a separate nitrogen accumulator
 
that supplies pressurized
 
nitrogen to force the water from the UHI accumulator
 
into the reactor vessel during the initial phase of a design-basis
 
loss-of-coolant
 
accident (LOCA). Thus, if a design-basis
 
LOCA had occurred while the UHI isolation
 
valves were inoperable, the UHI system would have been actuated;however, the UHI isolation
 
valves would not have closed when the water in the 8503210461 IN 85-23 March 22, 1985 UHI accumulator
 
had been depleted.
 
As a result, nitrogen gas could have been injected into the reactor vessel during the course of a design-basis
 
LOCA.Under such conditions, and using Appendix K assumptions, DPC's analysis indi-cated that the peak cladding temperature
 
of 2200'F most likely would have been exceeded and that the worst-case
 
increase in containment
 
pressure could have resulted in exceeding
 
the design pressure by 2 psi.A related but separate event involved the establishing
 
of the set points for closing the UHI isolation
 
valves. On February 14, 1984, DPC approved the use of a dry calibration
 
method, which would establish
 
the trip set point for closing the UHI isolation
 
valves relative to the bottom of the UHI water accumu-lator tank. However, a 24-inch nonconservative
 
error in the trip set point occurred at McGuire Units 1 and 2 when the responsible
 
instrument
 
engineer misinterpreted
 
the tank measurements
 
made by instrument
 
technicians.
 
Because the dry calibration
 
method does not use the actual process leg of the UHI accu-mulator, this error was left undetected
 
at both units for several months. The calibration
 
error was finally detected on November 2, 1984, while DPC personnel were taking "as-found" data in response to the previous error involving
 
the incorrect
 
installation
 
of the differential
 
pressure transmitters.
 
The conse-quences of this event would be the early isolation
 
of the UHI water accumulator
 
during a design-basis
 
LOCA, resulting
 
in less water being delivered
 
to the vessel than assumed in the analysis.A completely
 
unrelated
 
event involved the inoperability
 
of two of the four overpower
 
delta temperature
 
reactor protection
 
channels at McGuire Unit 2.This defect was discovered
 
on November 26, 1984, by a DPC engineer while per-forming a posttrip review of a reactor scram in which signals of the two affected channels responded
 
contrary to that expected.
 
This event was caused because an electrical
 
jumper was not installed
 
on two of the four overpower delta temperature
 
input logic cards. The purpose of the jumper is to ensure that the overpower
 
delta temperature
 
system provides protection
 
for decreasing
 
temperature, as might be expected on a steam line break. DPC's surveillance
 
tests only verified that protection
 
would be provided for increasing
 
tempera-ture, but not for decreasing
 
temperature.
 
This defect was left undetected
 
for an unknown period of time, but most likely it had existed since initial plant startup. Subsequent
 
investigations
 
revealed that in addition to inadequate
 
testing, there was an absence of instructions
 
and descriptions
 
of the required jumpers.The above examples illustrate
 
the need for thorough reviews and detailed attention
 
to plant surveillance
 
and postmaintenance
 
and postmodification
 
tests, to ensure that they accomplish
 
the required verification
 
of system function.
 
IN 85-23 March 22, 1985 No specific action or written response is required by this information
 
notice;however, if you have any questions
 
regarding
 
this notice, please contact the Regional Administrator
 
of the appropriate
 
NRC regional office or the technical contact listed below.Dieor Divis of Emergency
 
===Preparedness===
and 'ngineering
 
Response Office of Inspection
 
and Enforcement
 
Technical
 
Contacts:
I. Villalva, IE (301) 492-9007 H. Dance, RII (404) 221-5533 Attachment:
List of Recently Issued IE Information
 
Notices
 
Attachment
 
1 IN 85-23 March 22, 1985 LIST OF RECENTLY ISSUED IE INFORMATION
 
NOTICES Information
 
Date of Notice No. Subject Issue Issued to 85-22 85-21 Failure Of Limitorque
 
Motor-Operated Valves Resulting From Incorrect
 
===Installation===
Of Pinon Gear Main Steam Isolation
 
Valve Closure Logic 3/21/85 3/18/85 85-20 Motor-Operated
 
Valve Failures 3/12/85 Due To Hammering
 
Effect 85-19 85-10 Sup. 1 84-18 83-70 Sup. 1 85-17 85-16 85-15 Alleged Falsification
 
Of Certifications
 
===And Alteration===
Of Markings On Piping, Valves And Fittings Posstensioned
 
Containment
 
===Tendon Anchor Head Failure Failures Of Undervoltage===
Output Circuit Boards In The Westinghouse-Designed
 
Solid State Protection
 
System Vibration-Induced
 
Valve Failures Possible Sticking Of ASCO Solenoid Valves Time/Current
 
Trip Curve Discrepancy
 
Of ITE/Siemens- Allis Molded Case Circuit Breaker Nonconforming
 
Structural
 
Steel For Safety-Related
 
Use 3/11/85 3/8/85 3/7/85 3/4/85 3/1/85 2/27/85 2/22/85 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 All power reactor facilities
 
holding an OL or CP All power reactor facilities
 
holding an OL or CP All Westinghouse
 
===PWR facilities===
holding an OL or CP All power reactor facilities
 
holding an OL or CP All power reactor facilities
 
holding an OL or CP All power reactor facilities


==Description of Circumstances==
holding an OL or CP All power reactor facilities
:On November 1, 1984, Duke Power Company (DPC) informed the NRC that the fourRosemont differential pressure transmitters that control the closing of fourisolation valves of the upper-head injection (UHI) system at McGuire Unit 1were improperly installed (i.e., the impulse lines were reversed when theoriginal Barton reverse-acting differential pressure switches were replacedwith Rosemont direct-acting differential pressure transmitters during April of1984). As a result, the UHI isolation valves failed to close during drainingof the accumulator when the water level in the UHI accumulator reached the-setpoint. In addition to the improper installation, the postmodification testingwas limited to a dry calibration method that does not use the actual referenceleg of the accumulator; therefore, the installation error was not detected bythe postmodification test. Consequently, the plant was operated for approxi-mately five months with the UHI isolation valves inoperable.The McGuire UHI system design includes a separate nitrogen accumulator thatsupplies pressurized nitrogen to force the water from the UHI accumulator intothe reactor vessel during the initial phase of a design-basis loss-of-coolantaccident (LOCA). Thus, if a design-basis LOCA had occurred while the UHIisolation valves were inoperable, the UHI system would have been actuated;however, the UHI isolation valves would not have closed when the water in the8503210461 IN 85-23March 22, 1985 UHI accumulator had been depleted. As a result, nitrogen gas could have beeninjected into the reactor vessel during the course of a design-basis LOCA.Under such conditions, and using Appendix K assumptions, DPC's analysis indi-cated that the peak cladding temperature of 2200'F most likely would have beenexceeded and that the worst-case increase in containment pressure could haveresulted in exceeding the design pressure by 2 psi.A related but separate event involved the establishing of the set points forclosing the UHI isolation valves. On February 14, 1984, DPC approved theuse of a dry calibration method, which would establish the trip set point forclosing the UHI isolation valves relative to the bottom of the UHI water accumu-lator tank. However, a 24-inch nonconservative error in the trip set pointoccurred at McGuire Units 1 and 2 when the responsible instrument engineermisinterpreted the tank measurements made by instrument technicians. Becausethe dry calibration method does not use the actual process leg of the UHI accu-mulator, this error was left undetected at both units for several months. Thecalibration error was finally detected on November 2, 1984, while DPC personnelwere taking "as-found" data in response to the previous error involving theincorrect installation of the differential pressure transmitters. The conse-quences of this event would be the early isolation of the UHI water accumulatorduring a design-basis LOCA, resulting in less water being delivered to thevessel than assumed in the analysis.A completely unrelated event involved the inoperability of two of the fouroverpower delta temperature reactor protection channels at McGuire Unit 2.This defect was discovered on November 26, 1984, by a DPC engineer while per-forming a posttrip review of a reactor scram in which signals of the twoaffected channels responded contrary to that expected. This event was causedbecause an electrical jumper was not installed on two of the four overpowerdelta temperature input logic cards. The purpose of the jumper is to ensurethat the overpower delta temperature system provides protection for decreasingtemperature, as might be expected on a steam line break. DPC's surveillancetests only verified that protection would be provided for increasing tempera-ture, but not for decreasing temperature. This defect was left undetected foran unknown period of time, but most likely it had existed since initial plantstartup. Subsequent investigations revealed that in addition to inadequatetesting, there was an absence of instructions and descriptions of the requiredjumpers.The above examples illustrate the need for thorough reviews and detailedattention to plant surveillance and postmaintenance and postmodification tests,to ensure that they accomplish the required verification of system function.


IN 85-23March 22, 1985 No specific action or written response is required by this information notice;however, if you have any questions regarding this notice, please contact theRegional Administrator of the appropriate NRC regional office or the technicalcontact listed below.DieorDivis of Emergency Preparednessand 'ngineering ResponseOffice of Inspection and EnforcementTechnical Contacts: I. Villalva, IE(301) 492-9007H. Dance, RII(404) 221-5533Attachment: List of Recently Issued IE Information Notices
holding an OL or CP OL = Operating


Attachment 1IN 85-23March 22, 1985LIST OF RECENTLY ISSUEDIE INFORMATION NOTICESInformation Date ofNotice No. Subject Issue Issued to85-2285-21Failure Of Limitorque Motor-Operated Valves ResultingFrom Incorrect InstallationOf Pinon GearMain Steam Isolation ValveClosure Logic3/21/853/18/8585-20Motor-Operated Valve Failures 3/12/85Due To Hammering Effect85-1985-10Sup. 184-1883-70Sup. 185-1785-1685-15Alleged Falsification OfCertifications And AlterationOf Markings On Piping, ValvesAnd FittingsPosstensioned ContainmentTendon Anchor Head FailureFailures Of UndervoltageOutput Circuit Boards In TheWestinghouse-Designed SolidState Protection SystemVibration-Induced ValveFailuresPossible Sticking Of ASCOSolenoid ValvesTime/Current Trip CurveDiscrepancy Of ITE/Siemens-Allis Molded Case CircuitBreakerNonconforming StructuralSteel For Safety-RelatedUse3/11/853/8/853/7/853/4/853/1/852/27/852/22/85All power reactorfacilities holdingan OL or CPAll PWR facilitiesholding an OL or CPAll power reactorfacilities holdingan OL or CPAll power reactorfacilities holdingan OL or CPAll power reactorfacilities holdingan OL or CPAll WestinghousePWR facilitiesholding an OL or CPAll power reactorfacilities holdingan OL or CPAll power reactorfacilities holdingan OL or CPAll power reactorfacilities holdingan OL or CPAll power reactorfacilities holdingan OL or CPOL = Operating LicenseCP = Construction Permit
License CP = Construction


}}
Permit}}


{{Information notice-Nav}}
{{Information notice-Nav}}

Revision as of 12:19, 31 August 2018

Inadequate Surveillance and Postmaintenance and Postmodification System Testing
ML031180395
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: 03/22/1985
From: Jordan E L
NRC/IE
To:
References
IN-85-023, NUDOCS 8503210461
Download: ML031180395 (4)
v • d • e


SSINS No: 6835 IN 85-23 UNITED STATES NUCLEAR REGULATORY

COMMISSION

OFFICE OF INSPECTION

AND ENFORCEMENT

WASHINGTON, D.C. 20555 March 22, 1985 IE INFORMATION

NOTICE NO. 85-23: INADEQUATE

SURVEILLANCE

AND POSTMAINTENANCE

AND POSTMODIFICATION

SYSTEM TESTING

Addressees

All nuclear power reactor facilities

holding an operating

license (OL) or a construction

permit (CP).

Purpose

This information

notice is to alert addressees

of several instances

pertaining

to improper system modifications, inadequate

postmodification

system testing, and inadequate

surveillance

testing recently detected at the McGuire nuclear power facility.It is expected that recipients

will review the information

contained

in this notice for applicability

to their facilities

and consider actions, if appropri-ate, to preclude similar problems from occurring

at their facilities.

However, suggestions

contained

in this notice do not constitute

NRC requirements;

there-fore, no specific action or written response is required.Description

of Circumstances:

On November 1, 1984, Duke Power Company (DPC) informed the NRC that the four Rosemont differential

pressure transmitters

that control the closing of four isolation

valves of the upper-head

injection (UHI) system at McGuire Unit 1 were improperly

installed (i.e., the impulse lines were reversed when the original Barton reverse-acting

differential

pressure switches were replaced with Rosemont direct-acting

differential

pressure transmitters

during April of 1984). As a result, the UHI isolation

valves failed to close during draining of the accumulator

when the water level in the UHI accumulator

reached the-set point. In addition to the improper installation, the postmodification

testing was limited to a dry calibration

method that does not use the actual reference leg of the accumulator;

therefore, the installation

error was not detected by the postmodification

test. Consequently, the plant was operated for approxi-mately five months with the UHI isolation

valves inoperable.

The McGuire UHI system design includes a separate nitrogen accumulator

that supplies pressurized

nitrogen to force the water from the UHI accumulator

into the reactor vessel during the initial phase of a design-basis

loss-of-coolant

accident (LOCA). Thus, if a design-basis

LOCA had occurred while the UHI isolation

valves were inoperable, the UHI system would have been actuated;however, the UHI isolation

valves would not have closed when the water in the 8503210461 IN 85-23 March 22, 1985 UHI accumulator

had been depleted.

As a result, nitrogen gas could have been injected into the reactor vessel during the course of a design-basis

LOCA.Under such conditions, and using Appendix K assumptions, DPC's analysis indi-cated that the peak cladding temperature

of 2200'F most likely would have been exceeded and that the worst-case

increase in containment

pressure could have resulted in exceeding

the design pressure by 2 psi.A related but separate event involved the establishing

of the set points for closing the UHI isolation

valves. On February 14, 1984, DPC approved the use of a dry calibration

method, which would establish

the trip set point for closing the UHI isolation

valves relative to the bottom of the UHI water accumu-lator tank. However, a 24-inch nonconservative

error in the trip set point occurred at McGuire Units 1 and 2 when the responsible

instrument

engineer misinterpreted

the tank measurements

made by instrument

technicians.

Because the dry calibration

method does not use the actual process leg of the UHI accu-mulator, this error was left undetected

at both units for several months. The calibration

error was finally detected on November 2, 1984, while DPC personnel were taking "as-found" data in response to the previous error involving

the incorrect

installation

of the differential

pressure transmitters.

The conse-quences of this event would be the early isolation

of the UHI water accumulator

during a design-basis

LOCA, resulting

in less water being delivered

to the vessel than assumed in the analysis.A completely

unrelated

event involved the inoperability

of two of the four overpower

delta temperature

reactor protection

channels at McGuire Unit 2.This defect was discovered

on November 26, 1984, by a DPC engineer while per-forming a posttrip review of a reactor scram in which signals of the two affected channels responded

contrary to that expected.

This event was caused because an electrical

jumper was not installed

on two of the four overpower delta temperature

input logic cards. The purpose of the jumper is to ensure that the overpower

delta temperature

system provides protection

for decreasing

temperature, as might be expected on a steam line break. DPC's surveillance

tests only verified that protection

would be provided for increasing

tempera-ture, but not for decreasing

temperature.

This defect was left undetected

for an unknown period of time, but most likely it had existed since initial plant startup. Subsequent

investigations

revealed that in addition to inadequate

testing, there was an absence of instructions

and descriptions

of the required jumpers.The above examples illustrate

the need for thorough reviews and detailed attention

to plant surveillance

and postmaintenance

and postmodification

tests, to ensure that they accomplish

the required verification

of system function.

IN 85-23 March 22, 1985 No specific action or written response is required by this information

notice;however, if you have any questions

regarding

this notice, please contact the Regional Administrator

of the appropriate

NRC regional office or the technical contact listed below.Dieor Divis of Emergency

Preparedness

and 'ngineering

Response Office of Inspection

and Enforcement

Technical

Contacts:

I. Villalva, IE (301) 492-9007 H. Dance, RII (404) 221-5533 Attachment:

List of Recently Issued IE Information

Notices

Attachment

1 IN 85-23 March 22, 1985 LIST OF RECENTLY ISSUED IE INFORMATION

NOTICES Information

Date of Notice No. Subject Issue Issued to 85-22 85-21 Failure Of Limitorque

Motor-Operated Valves Resulting From Incorrect

Installation

Of Pinon Gear Main Steam Isolation

Valve Closure Logic 3/21/85 3/18/85 85-20 Motor-Operated

Valve Failures 3/12/85 Due To Hammering

Effect 85-19 85-10 Sup. 1 84-18 83-70 Sup. 1 85-17 85-16 85-15 Alleged Falsification

Of Certifications

And Alteration

Of Markings On Piping, Valves And Fittings Posstensioned

Containment

Tendon Anchor Head Failure Failures Of Undervoltage

Output Circuit Boards In The Westinghouse-Designed

Solid State Protection

System Vibration-Induced

Valve Failures Possible Sticking Of ASCO Solenoid Valves Time/Current

Trip Curve Discrepancy

Of ITE/Siemens- Allis Molded Case Circuit Breaker Nonconforming

Structural

Steel For Safety-Related

Use 3/11/85 3/8/85 3/7/85 3/4/85 3/1/85 2/27/85 2/22/85 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 All power reactor facilities

holding an OL or CP All power reactor facilities

holding an OL or CP All Westinghouse

PWR facilities

holding an OL or CP All power reactor facilities

holding an OL or CP All power reactor facilities

holding an OL or CP All power reactor facilities

holding an OL or CP All power reactor facilities

holding an OL or CP OL = Operating

License CP = Construction

Permit


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