Information Notice 1993-02, Malfunction of a Pressurizer Code Safety Valve

Malfunction of a Pressurizer Code Safety Valve

ML031080113
Person / Time
Site:	Beaver Valley, Millstone, Hatch, Monticello, Calvert Cliffs, Dresden, Davis Besse, Peach Bottom, Browns Ferry, Salem, Oconee, 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
Issue date:	01/04/1993
From:	Grimes B Office of Nuclear Reactor Regulation
To:
References
IN-93-002, NUDOCS 9212280132
Download: ML031080113 (14)

I-

UNITED STATES

NUCLEAR REGULATORY COMMISSION

OFFICE OF NUCLEAR REACTOR REGULATION

WASHINGTON, D.C. 20555 January 4, 1993 NRC INFORMATION NOTICE 93-02: MALFUNCTION OF A PRESSURIZER CODE SAFETY VALVE

Addressees

All holders of operating licenses or construction permits for nuclear power

reactors.

Purpose

The U.S. Nuclear Regulatory Commission (NRC) is issuing this information

notice to alert addressees to the failure of a pressurizer code safety

valve to maintain set pressure and reseat properly during a plant transient.

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 July 3, 1992, after an electrical transient, the reactor at Fort Calhoun

Station tripped on high primary pressure. Both power-operated relief valves

opened and valve RC-142, one of two pressurizer code safety valves, lifted

prematurely at a pressure below 16.75 MPa [2430 psia], as opposed to the

proper setpoint pressure of 17.24 Mpa [2500 psia] +/- 1 percent. The relief

valves shut automatically when the reactor coolant system pressure decreased

to 16.20 Mpa [2350 psia]. Because a safety valve was still open, the pressure

continued to decrease and RC-142 subsequently reseated at approximately

12.03 MPa [1745 psia]. The pressure then increased and RC-142 lifted again at

approximately 13.27 MPa [1925 psia]. RC-142 partially reseated, as pressure

again dropped, at approximately 6.89 MPa [1000 psia]. RC-142 continued to

leak, as indicated by the tail pipe temperature, until the plant was brought

to cold shutdown. The licensee removed RC-142 and sent it to Wyle Laboratory

(Wyle) for inspection and testing.

Discussion

The code safety valves installed at Fort Calhoun are "3-inch inlet by 6-inch

outlet", Size 3K6, Style HB-86-BP, Type E valves (Figure 1) manufactured by

the Crosby Valve and Gage Company (Crosby). In early 1980, the Electric Power

and Research Institute (EPRI) tested Crosby safety valves that have loop seals

and are subjected to back pressure. The EPRI test results indicated that the

initial discharge of the loop seal or a transition from discharging steam to

discharging water could cause the valve to chatter.

9212280132 Pa P.

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- 002098

IN 93-02 January 4, 1993 The licensee believes that RC-142 chattered during its initial lift from the

discharge of the loop seal. Apparently, the chatter loosened the locknut on

the adjusting bolt and allowed the adjusting bolt to partially back out.

Later, primary water discharged through RC-142 for approximately 5 minutes

(pressurizer level reached 100 percent) during its second lift and subsequent

partial reseat at approximately 6.9 MPa [1000 psia]. The discharging water

induced further chattering, apparently causing the adjusting bolt to back out

even further, reducing the valve lift setpoint to approximately

10.18 MPa [1477 psia].

To ensure that the adjusting bolt would not back out again, Crosby designed a

special mechanical locking device and installed it on the two valves at

Fort Calhoun. Crosby also specified a torque value of "400 foot-pounds" for

the adjusting bolt locknut. This value had not been previously specified in

procedures used by Wyle for inspecting and testing pressurizer code safety

valves.

An NRC augmented inspection team monitored licensee activities at Fort Calhoun

and Wyle Laboratory. At Wyle, the locknut for RC-142 was found to have backed

off from the top of the valve bonnet by approximately 3 to 6 mm [1/8 to

1/4 inch] and could be turned by hand. The adjusting bolt was determined to

be 19.5 flats of bolt revolution from the zero compression position of the

spring. Crosby representatives calculated that this position corresponded to

a setpoint value of approximately 10.18 MPa [1477 psia]. In March 1992, the

valve had been set to 17.24 MPa [2500 psia] +/- 1 percent at Wyle Laboratory.

When the valve internals were removed, the bellows assembly was found to have

failed on each end at the first weld after the transition weld. Also, the

disc insert was found jammed into the disc holder. The disc insert was

recessed approximately 0.05 mm [0.002 inch] below the top surface of the disc

ring. The disc ring was seated on the nozzle ring, which indicated that the

valve had not reseated properly. NRC Inspection Report 50-285/92-18 contains

additional information on this event.

On August 22, 1992, an additional problem was revealed when RC-142 again

lifted prematurely. This premature lift occurred as the reactor coolant

system pressure increased to approximately 16.53 MPa (2397 psia]. This

pressure was approximately 4 percent below the normal setpoint of

17.24 MPa [2500 psia] +/- 1 percent. However, the valve reseated properly

with no leakage detected before or after the valve lift.

Because of the premature lift, both valves were returned to Wyle Laboratory

for additional inspection and testing. Wyle inspected RC-142 and found the

valve to be in good condition with only minor nicks on the nozzle seat. When

Wyle attempted to test RC-142, certain conditions and difficulties were noted.

• RC-142 has a stainless steel nozzle and a carbon steel body.

• The initial attempt to test the valve under "cold" conditions to

simulate the normal operating conditions of the plant could not be

e I

IN 93-02 January 4, 1993 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 (NRR) project manager.

Brian K. Grimes, Director

Division of Operating Reactor Support

Office of Nuclear Reactor Regulation

Technical contacts: T. F. Westerman, Region IV

(817) 860-8145 P. A. Goldberg, Region IV

(817) 860-8168 Attachments:

1. Figure 1, Fort Calhoun Pressurizer Safety Valve

2. List of Recently Issued NRC Information Notices

IN 93-02 January 4, 1993 performed as planned because of increases in the valve internal and

external temperatures before and after the lift of the valve.

However, the testing revealed that the lift setpoint would increase with

increasing nozzle temperature and then decrease as the valve body temperature

began to increase. Licensee personnel concluded that the valves would require

testing at "hot" upper bound temperatures with saturated inlet steam near

setpoint pressure and with the valves insulated with the actual plant

insulation. The insulation used by Wyle differed somewhat in composition and

fit from the plant insulation and apparently affected the distribution of heat

within the valve. The data, obtained from the tests performed with saturated

steam, confirmed that the setpoint value initially increased due to the

thermal expansion of the valve nozzle, then decreased as the temperature of

the valve body increased. The setpoint value stabilized once the valve

temperature stabilized. This condition was also observed during the "in-situ"

Trevitest testing performed at the plant. As a result of this testing, the

licensee performed a safety analysis and determined that the reactor coolant

system could withstand an overpressure transient with a code safety valve

pressure setpoint deviation of +6 percent (1.03 MPa [150 psi]) and that a

pressure setpoint deviation of -4 percent (0.69 MPa [100 psi]) would not cause

unnecessary challenges to the safety valves. The licensee also reduced the

power-operated relief valve and reactor high pressure trip setpoint by

0.35MPa [50 psi]. The licensee is considering the removal of the loop seal as

a long-term action.

After completing the testing, Wyle reset both code safety valves to the

technical specification setpoint of "2500 psia +/- 1 percent." The valves

were then returned to Fort Calhoun and were reinstalled.

The licensee believes that the premature lift of RC-142 on August 22, 1992, resulted from using Wyle insulation during testing rather than the actual

plant insulation. The tests using Wyle insulation resulted in a lower

stabilized valve temperature than resulted from using the actual plant

insulation. The difference in the test methods resulted in an approximate

3 percent difference in setpoint. NRC Inspection Report 50-285/92-21 contains

additional information on this event.

Attachment 1 IN 93-02 January 4, 1993 (new)

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8

12

13

7

4

1 FORT CzJHOUN PRESSURIZER SAFETY VAcjaE

Attachment 2 IN 93-02 January 4, 1993 Page 1 of I

LIST OF RECENTLY ISSUED

NRC INFORMATION NOTICES

Information Date of

Notice No. Subject Issuance Issued to

93-01 Accuracy of Motor- 01/04/93 All holders of OLs or CPs

Operated Valve Diagnostic for nuclear power reactors.

Equipment Manufactures

by Liberty Technologies

92-86 Unexpected Restriction 12/24/92 All holders of OLs or CPs

to Thermal Growth of for nuclear power reactors.

Reactor Coolant Piping

92-85 Potential Failures of 12/23/92 All holders of OLs or CPs

Emergency Core Cooling for nuclear power reactors.

Systems Caused by

Foreign Material Blockage

92-84 Release of Patients 12/17/92 All Nuclear Regulatory

Treated with Temporary Commission Medical Licensees

Implants

88-23, Potential for Gas 12/18/92 All holders of OLs or CPs

Supp. 4 Binding of High-Pres- for nuclear power reactors.

sure Safety Injection

Pumps during A Design

Basis Accident

92-83 Thrust Limits for 12/17/92 All holders of OLs or CPs

Limitorque Actuators for nuclear power reactors.

and Potential Over- stressing of Motor- Operated Valves

92-82 Results of Thermo-Lag 12/15/92 All holders of OLs or CPs

330-1 Combustibility for nuclear power reactors.

Testing

92-81 Potential Deficiency 12/11/92 All holders of OLs or CPs

of Electrical Cables for nuclear power reactors.

with Bonded Hypalon

Jackets

92-80 Results of Thermo-Lag 12/07/92 All holders of OLs or CPs

330-1 Combustibility for nuclear power reactors.

Testing

OL = Operating License

CP = Construction Permit

UNITED STATES

. NUCLEAR REGULATORY COMMISSION FIRST CLASS MAIL

• . WASHINGTON, D.C. 20555-0001 POSTAGE AND FEES PAID

USNRC

PERMIT NO. G-67 OFFICIAL BUSINESS

PENALTY FOR PRIVATE USE, $300

PRINTED ON RECYCLED PAPER

-....

9 N 93-02 January 4, 1993 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 (NRR) project manager.

OWgina signed by

Brian K.Grimes

Brian K. Grimes, Director

Division of Operating Reactor Support

Office of Nuclear Reactor Regulation

Technical contacts: T. F. Westerman, Region IV

(817) 860-8145 P. A. Goldberg, Region IV

(817) 860-8168 Attachments:

1. Figure 1, Fort Calhoun Pressurizer Safety Valve

2. List of Recently Issued NRC Information Notices

• SEE PREVIOUS CONCURRENCE

• OGCB:DORS *OEAB:DORS *OEAB:DORS *ASC/OEAB:DORS *ADM: RPB

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IN 92-XX

December xx, 1992 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 (NRR) project manager.

Brian K. Grimes, Director

Division of Operating Reactor Support

Office of Nuclear Reactor Regulation

Technical contacts: T. F. Westerman, Region IV

(817) 860-8145 P. A. Goldberg, Region IV

(817) 860-8168 Attachments:

1. Figure 1, Fort Calhoun Pressurizer Safety Valve

2. List of Recently Issued NRC Information Notices

• SEE PREVIOUS CONCURRENCE

• OGCB:DORS *OEAB:DORS *OEAB:DORS *ASC/OEAB:DORS *ADM:RPB

JBirmingham:mkm JRamsey KMarcus AGautam JMain

12/04/92 09/29/92 09/29/92 09/26/92 09/29/92

• RIV/DRS *RIV/DRP *C/EMEB:DE *C/OEAB:DORS

TWesterman PHarrell JNorberg AChaffee

11/04/92 11/16/92 11/16/92 11/23/92

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IN 92-XX

December xx, 1992 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 (NRR) project manager.

Brian K. Grimes, Director

Division of Operating Reactor Support

Office of Nuclear Reactor Regulation

Technical contacts: T. F. Westerman, Region IV

(817) 860-8145 P. A. Goldberg, Region IV

(817) 860-8168 Attachments:

1. Figure 1, Fort Calhoun Pressurizer Safety Valve

2. List of Recently Issued NRC Information Notices

• SEE PREVIOUS CONCURRENCE

• OGCB:DORS *OEAB:D0RS *OEAB:DORS *ASC/OEAB:DORS *ADM:RPB

JBirmingham:mkm JRamsey KMarcus AGautam JMain

12/04/92 09/29/92 09/29/92 09/26/92 09/29/92

• RIV/DRS *RIV/DRP *C/EMEB:DE *C/OEAB:DORS

TWesterman PHarrell JNorberg AChaffee

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IN 92-XX

December xx, 1992 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 (NRR) project manager.

Brian K. Grimes, Director

Division of Operating Reactor Support

Office of Nuclear Reactor Regulation

Technical contacts: T. F. Westerman, Region IV

(817) 860-8145 P. A. Goldberg, Region IV

(817) 860-8168 Attachments:

1. Figure 1, Fort Calhoun Pressurizer Safety Valve

2. List of Recently Issued NRC Information Notices

• SEE PREVIOUS CONCURRENCE

OGCB:DORS *OEAB:DORS *OEAB:DORS *ASC/OEAB:DORS *ADM:RPB

JBirmingham:mkm JRamsey KMarcus AGautam JMain

12/ 09/29/92 09/29/92 09/26/92 09/29/92

• RIV/DRS *RIV/DRP *C/EMEB:DE *C/OEAB:DORS

TWesterman PHarrell JNorberg AChaffee

11/04/92 11/16/92 11/16/92 11/23/92 C/OGCB:DORS D/DORS

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IN 92-XX

September xx, 1992 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 (NRR) project manager.

Charles E. Rossi, Director

Division of Operational Events Assessment

Office of Nuclear Reactor Regulation

Technical contacts: T. F. Westerman, RIV

(817) 860-8145 P. A. Goldberg, RIV

(817) 860-8168 K. R. Marcus, NRR

(301) 504-1170

Attachments:

1. Crosby Pressurizer Code Safety Valve Figure

2. List of Recently Issued NRC Information Notices

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