Information Notice 1985-23, Inadequate Surveillance and Postmaintenance and Postmodification System Testing: Difference between revisions

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


COMMISSION
===UNITED STATES===
NUCLEAR REGULATORY COMMISSION


OFFICE OF INSPECTION
===OFFICE OF INSPECTION AND ENFORCEMENT===
WASHINGTON, D.C.


===AND ENFORCEMENT===
20555
WASHINGTON, D.C. 20555 March 22, 1985 IE INFORMATION


NOTICE NO. 85-23: INADEQUATE
===March 22, 1985===
IE INFORMATION NOTICE NO. 85-23:  


SURVEILLANCE
===INADEQUATE SURVEILLANCE AND POSTMAINTENANCE===
 
AND POSTMODIFICATION SYSTEM TESTING
===AND POSTMAINTENANCE===
AND POSTMODIFICATION
 
SYSTEM TESTING


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


permit (CP).
construction permit (CP).


==Purpose==
==Purpose==
: This information
:
This information notice is to alert addressees of several instances pertaining


notice is to alert addressees
to improper system modifications, inadequate postmodification system testing, and inadequate surveillance testing recently detected at the McGuire nuclear


of several instances
power facility.


pertaining
It is expected that recipients will review the information contained in this


to improper system modifications, inadequate
notice for applicability to their facilities and consider actions, if appropri- ate, to preclude similar problems from occurring at their facilities.


postmodification
However, suggestions contained in this notice do not constitute NRC requirements; there- fore, no specific action or written response is required.


system testing, and inadequate
==Description of Circumstances==
:
On November 1, 1984, Duke Power Company (DPC) informed the NRC that the four


surveillance
Rosemont differential pressure transmitters that control the closing of four


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


will review the information
original Barton reverse-acting differential pressure switches were replaced


contained
with Rosemont direct-acting differential pressure transmitters during April of


in this notice for applicability
1984).


to their facilities
As a result, the UHI isolation valves failed to close during draining


and consider actions, if appropri-ate, to preclude similar problems from occurring
of the accumulator when the water level in the UHI accumulator reached the-set


at their facilities.
point.


However, suggestions
In addition to the improper installation, the postmodification testing


contained
was limited to a dry calibration method that does not use the actual reference


in this notice do not constitute
leg of the accumulator; therefore, the installation error was not detected by


NRC requirements;
the postmodification test.
there-fore, no specific action or written response is required.Description


of Circumstances:
Consequently, the plant was operated for approxi- mately five months with the UHI isolation valves inoperable.
On November 1, 1984, Duke Power Company (DPC) informed the NRC that the four Rosemont differential


pressure transmitters
The McGuire UHI system design includes a separate nitrogen accumulator that


that control the closing of four isolation
supplies pressurized nitrogen to force the water from the UHI accumulator into


valves of the upper-head
the reactor vessel during the initial phase of a design-basis loss-of-coolant


injection (UHI) system at McGuire Unit 1 were improperly
accident (LOCA).


installed (i.e., the impulse lines were reversed when the original Barton reverse-acting
Thus, if a design-basis LOCA had occurred while the UHI


differential
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


pressure switches were replaced with Rosemont direct-acting
8503210461


differential
IN 85-23 March 22, 1985 UHI accumulator had been depleted.


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


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


valves failed to close during draining of the accumulator
exceeded and that the worst-case increase in containment pressure could have


when the water level in the UHI accumulator
resulted in exceeding the design pressure by 2 psi.


reached the-set point. In addition to the improper installation, the postmodification
A related but separate event involved the establishing of the set points for


testing was limited to a dry calibration
closing the UHI isolation valves.


method that does not use the actual reference leg of the accumulator;
===On February 14, 1984, DPC approved the===
therefore, the installation
use of a dry calibration method, which would establish the trip set point for


error was not detected by the postmodification
closing the UHI isolation valves relative to the bottom of the UHI water accumu- lator tank.


test. Consequently, the plant was operated for approxi-mately five months with the UHI isolation
However, a 24-inch nonconservative error in the trip set point


valves inoperable.
occurred at McGuire Units 1 and 2 when the responsible instrument engineer


The McGuire UHI system design includes a separate nitrogen accumulator
misinterpreted the tank measurements made by instrument technicians.


that supplies pressurized
Because


nitrogen to force the water from the UHI accumulator
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


into the reactor vessel during the initial phase of a design-basis
calibration error was finally detected on November 2, 1984, while DPC personnel


loss-of-coolant
were taking "as-found" data in response to the previous error involving the


accident (LOCA). Thus, if a design-basis
incorrect installation of the differential pressure transmitters. The conse- quences of this event would be the early isolation of the UHI water accumulator


LOCA had occurred while the UHI isolation
during a design-basis LOCA, resulting in less water being delivered to the


valves were inoperable, the UHI system would have been actuated;however, the UHI isolation
vessel than assumed in the analysis.


valves would not have closed when the water in the 8503210461 IN 85-23 March 22, 1985 UHI accumulator
A completely unrelated event involved the inoperability of two of the four


had been depleted.
overpower delta temperature reactor protection channels at McGuire Unit 2.


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


LOCA.Under such conditions, and using Appendix K assumptions, DPC's analysis indi-cated that the peak cladding temperature
affected channels responded contrary to that expected.


of 2200'F most likely would have been exceeded and that the worst-case
===This event was caused===
because an electrical jumper was not installed on two of the four overpower


increase in containment
delta temperature input logic cards.


pressure could have resulted in exceeding
===The purpose of the jumper is to ensure===
that the overpower delta temperature system provides protection for decreasing


the design pressure by 2 psi.A related but separate event involved the establishing
temperature, as might be expected on a steam line break.


of the set points for closing the UHI isolation
DPC's surveillance


valves. On February 14, 1984, DPC approved the use of a dry calibration
tests only verified that protection would be provided for increasing tempera- ture, but not for decreasing temperature. This defect was left undetected for


method, which would establish
an unknown period of time, but most likely it had existed since initial plant


the trip set point for closing the UHI isolation
startup.


valves relative to the bottom of the UHI water accumu-lator tank. However, a 24-inch nonconservative
Subsequent investigations revealed that in addition to inadequate


error in the trip set point occurred at McGuire Units 1 and 2 when the responsible
testing, there was an absence of instructions and descriptions of the required


instrument
jumpers.


engineer misinterpreted
The above examples illustrate the need for thorough reviews and detailed


the tank measurements
attention to plant surveillance and postmaintenance and postmodification tests, to ensure that they accomplish the required verification of system function.


made by instrument
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


technicians.
Regional Administrator of the appropriate NRC regional office or the technical


Because the dry calibration
contact listed below.


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


at both units for several months. The calibration
Divis


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


the incorrect
and 'ngineering Response


installation
===Office of Inspection and Enforcement===
Technical Contacts: I. Villalva, IE


of the differential
(301) 492-9007


pressure transmitters.
===H. Dance, RII===
(404) 221-5533 Attachment:


The conse-quences of this event would be the early isolation
===List of Recently Issued IE Information Notices===


of the UHI water accumulator
===Attachment 1===
IN 85-23


during a design-basis
===March 22, 1985===
LIST OF RECENTLY ISSUED


LOCA, resulting
===IE INFORMATION NOTICES===
Information


in less water being delivered
Date of


to the vessel than assumed in the analysis.A completely
Notice No.


unrelated
Subject


event involved the inoperability
Issue


of two of the four overpower
Issued to


delta temperature
85-22
85-21 Failure Of Limitorque Motor-


reactor protection
===Operated Valves Resulting===
From Incorrect Installation


channels at McGuire Unit 2.This defect was discovered
===Of Pinon Gear===
Main Steam Isolation Valve


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
===Closure Logic===
3/21/85
3/18/85
85-20
Motor-Operated Valve Failures 3/12/85


contrary to that expected.
===Due To Hammering Effect===
85-19
85-10
Sup. 1
84-18
83-70
Sup. 1
85-17
85-16
85-15


This event was caused because an electrical
===Alleged Falsification Of===
Certifications And Alteration


jumper was not installed
===Of Markings On Piping, Valves===
And Fittings


on two of the four overpower delta temperature
===Posstensioned Containment===
Tendon Anchor Head Failure


input logic cards. The purpose of the jumper is to ensure that the overpower
===Failures Of Undervoltage===
Output Circuit Boards In The


delta temperature
Westinghouse-Designed Solid


system provides protection
===State Protection System===
Vibration-Induced Valve


for decreasing
Failures


temperature, as might be expected on a steam line break. DPC's surveillance
===Possible Sticking Of ASCO===
Solenoid Valves


tests only verified that protection
Time/Current Trip Curve


would be provided for increasing
Discrepancy Of ITE/Siemens-


tempera-ture, but not for decreasing
===Allis Molded Case Circuit===
Breaker


temperature.
===Nonconforming Structural===
 
Steel For Safety-Related
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
Use


===Installation===
3/11/85
Of Pinon Gear Main Steam Isolation
3/8/85
3/7/85
3/4/85
3/1/85
2/27/85
2/22/85


Valve Closure Logic 3/21/85 3/18/85 85-20 Motor-Operated
===All power reactor===
facilities holding


Valve Failures 3/12/85 Due To Hammering
an OL or CP


Effect 85-19 85-10 Sup. 1 84-18 83-70 Sup. 1 85-17 85-16 85-15 Alleged Falsification
===All PWR facilities===
holding an OL or CP


Of Certifications
===All power reactor===
facilities holding


===And Alteration===
an OL or CP
Of Markings On Piping, Valves And Fittings Posstensioned


Containment
===All power reactor===
facilities holding


===Tendon Anchor Head Failure Failures Of Undervoltage===
an OL or CP
Output Circuit Boards In The Westinghouse-Designed


Solid State Protection
===All power reactor===
 
facilities holding
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
an OL or CP


holding an OL or CP All PWR facilities
===All Westinghouse===
PWR facilities


holding an OL or CP All power reactor facilities
holding an OL or CP


holding an OL or CP All power reactor facilities
===All power reactor===
facilities holding


holding an OL or CP All power reactor facilities
an OL or CP


holding an OL or CP All Westinghouse
===All power reactor===
facilities holding


===PWR facilities===
an OL or CP
holding an OL or CP All power reactor facilities


holding an OL or CP All power reactor facilities
===All power reactor===
facilities holding


holding an OL or CP All power reactor facilities
an OL or CP


holding an OL or CP All power reactor facilities
===All power reactor===
facilities holding


holding an OL or CP OL = Operating
an OL or CP


License CP = Construction
OL = Operating License


Permit}}
CP = Construction Permit}}


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

Latest revision as of 10:24, 16 January 2025

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
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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