Information Notice 1985-23, Inadequate Surveillance and Postmaintenance and Postmodification System Testing

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)

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:

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