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:

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