Information Notice 1992-50, Cracking of Valves in the Condensate Return Lines of a BWR Emergency Condenser System

Cracking of Valves in the Condensate Return Lines of a BWR Emergency Condenser System
	ML031210758
Person / Time
Issue date:	07/02/1992
From:	Rossi C; Office of Nuclear Reactor Regulation
To:	;
References
	IN-92-050, NUDOCS 9206290237
	Download: ML031210758 (10)
	v • d • e

UNITED STATES

NUCLEAR REGULATORY COMMISSION

OFFICE OF NUCLEAR REACTOR REGULATION

WASHINGTON, D.C.

20555

July 2, 1992

NRC INFORMATION NOTICE 92-50: CRACKING OF VALVES IN THE CONDENSATE RETURN

LINES OF A BWR EMERGENCY CONDENSER SYSTEM

Addressees

All holders of operating licenses or construction permits for boiling water

reactors (BWRs).

Purpose

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

notice to inform addressees of cracking found in valves in the condensate

return lines of the emergency condenser system at the Nine Mile Point Nuclear

Station, Unit 1. 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

Following a reactor trip at Nine Mile Point, Unit 1, on May 1, 1992, the

licensee (Niagara Mohawk Power Corporation) inspected the drywell to

investigate the cause of a recent increase of unidentified leakage in the

reactor coolant system.

The licensee found a 0.5 gpm leak coming from a

manual gate valve at a 1-inch drain line connection. The leaking gate valve, designated valve 39-02, is located in the condensate return line for the

loop 12 emergency condenser system.

The emergency condenser system has two independent loops (loops 11 and 12).

Figure 1 shows the configuration of the condensate return line in loops 11 and 12.

In the condensate return line, a manual gate valve is connected

downstream of a tilting disc check valve. At each of those two valves, two

1-inch drain lines are connected to the bottom part of the valve body with one

drain line at the upstream side and the other one at the downstream side of

the valve. The valve bodies are made of CF8M cast stainless steel.

While investigating the leakage at the manual gate valve 39-02, the licensee

removed the internal components of the adjacent check valve to perform a

visual test (VT), a radiographic test (RT), and an ultrasonic test (UT).

The

licensee visually observed cracks on the inside surfaces at both valves in

loop 12.

At gate valve 39-02, the licensee found cracks near each of the two

drain holes. At check valve 39-04, the licensee found cracks near a

downs.

.,drain hole and found evidence of cracking in the threads of the

9206290237

-

-1 IN 92-50

July 2, 1992 upstream drain.

These cracks were further examined radiographically and

ultrasonically.

The licensee found four cracks including a throughwall crack

near the drain hole upstream of'gate valve 39-02. The licensee reported the

throughwall crack to be about 3.5 inches long and oriented radially outward

from the hole.

The other three cracks were all reported to be within 0.15 to

0.35 inch of passing through the wall (1.25 inch wall thickness).

The

licensee found two cracks in the drain hole area downstream of the gate valve

(39-02), which is the unisolable side of the valve body.

The licensee

reported the deepest crack to be about 1 inch long and within 0.15 inch of

passing through the wall.

The licensee found four cracks near the drain hole

downstream of check valve 39-04 with the deepest reported to be within

0.1 inch of passing through the wall.

The licensee visually observed one

small indication on the seat ring in manual gate valve 39-02.

The licensee

also examined valves 39-01 and 39-03 in the condensate return line for loop 11 of the emergency condenser system and found two cracks near the drain hole

upstream of manual gate valve 39-01.

The largest crack was reported to be

about 1.25 inch long and 1 inch deep. The licensee reported the cracking of

the valve body in loop 11 to be less severe than that in loop 12.

The

licensee observed cracking indications on the inside surface of a butt weld

that joins the gate valve to the check valve but did not confirm these

indications by the radiographic examination. The licensee ultrasonically

examined selected piping welds inboard of the condensate return isolation

valves and found no indications.

The licensee removed a boat sample containing a 0.5-inch long crack from

manual gate valve 39-02 in loop 12.

The licensee examined the boat sample

metallographically and fractographically (using a scanning electron

microscope) and found that the crack had propagated transgranularly with very

little secondary cracking.

These features are typical of fatigue crack

propagation.

The licensee noted possible fatigue striations that were not

well-developed.

The licensee measured the delta ferrite content of the boat

sample to be about 15 percent.

Discussion

The emergency condensate system at Nine Mile Point, Unit 1, which is connected

directly to the reactor coolant system, operates by natural circulation and

acts as a backup for the main condenser to remove the reactor decay heat

following a reactor isolation.

The emergency condenser system at Nine Mile

Point, Unit 1, as shown in Figure 2, has two loops (loop 11 and loop 12) with

two condensers in each loop. During normal plant operation, the condensate

return isolation valves (39-05 and 39-06) are closed, and the steam isolation

valves (39-07, 08, 09, and 10) are open in each loop.

As shown in Figure 1, two valves, a manual gate valve and a tilting disc check valve, are located in

horizontal sections of the condensate return line.

The horizontal sections

are connected to the suction side of the recirculation piping system.

The

manual gate valves are maintenance valves and are open during normal

operation.

The licensee postulated thermal fatigue as the root cause of the cracking in

the valve bodies, upon considering the straight and transgranular cracking

morphology, the location of the cracks on the bottom surface near

IN 92-50

July 2, 1992 discontinuities, and the orientation of the cracks. However, the licensee did

not find the direct causes of the apparent thermal stratification and cycling

at the affected valves. The licensee speculated that the observed cracking

may have been caused by the leaking of the cold water from the condensate

isolation valves (39-05 and 06) and the periodic opening of the tilting disc

in the check valve. The licensee provided a limited history of the time and

temperature as evidence of thermal cycling in the loop 12 condensate return

line valve 39-06. Although the licensee also observed cracking in loop 11, it

did not observe such thermal cycling on the condensate return line during a

1-week test. The sections of the emergency condenser condensate return lines

that showed evidence of cracking are classified as American Society of

Mechanical Engineers (ASME) Code Class 1. The licensee extended its current

outage to complete acceptable code repairs because of the extent of the cracks

in the reactor coolant pressure boundary and, in particular, the cracks found

at the downstream drain line hole for valve 39-02.

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.

4es4re

Division of Operational Events Assessment

Office of Nuclear Reactor Regulation

Technical contacts: William H. Koo, NRR

(301) 504-2706

Robert A. Hermann, NRR

(301) 504-2768 Attachments:

1. Figure 1, NMP-1 Emergency Condenser System Condensate Return

Line Configuration Inside Drywell

2. Figure 2, Nine Mile Point Unit One Emergency Condenser System

Simplified Diagram

3. List of Recently Issued NRC Information Notices

AIR OPERATED-,,

GLOBE VALVE

SECTION A-A

39-06(LOOP

12$

REACTOR

MANUAL GATE VALVE

RECIRCULATION

39-01 (LOOP 11 DRYWELL

FLOW FROM

39-02 (LOOP 12)

SHELL

VESSEL ANNULUS

'

1

\\TILTING

DISC CHECK VALVE

'

39-03 LOOP 11)

39-04 LOOP 12)

I

0

-/

TO

hi

RECIRCULATION

,n

PUMP SUCTION

o

a

FIGURE 1 NMP-1 EMERGENCY CONDENSER SYSTEM

CONDENSATE RETURN LINE

CONFIGURATION INSIDE DRYWELL

-

X

-

)I

AttC

IN !

Jul'

Page

? CONDENSER

EMERGENCY C

UP TANK

MAKEUP

00 G.-

(40.000

GENCY

EDMERGi

ENSERS

39-07 39-09 RPV

39-10 39-08

39-04

39-02

\\

~RECIRCULATION/

\\

~PUMPS (2 OF 5)/

FIGURE 2

NINE MILE POINT UNIT ONE

EMERGENCY CONDENSER SYSTEM

SIMPLIFIED DIAGRAM

3achmnt 2

92-50

.y 2, 1992 e 1 of 1

A

.

-

&

Attachment 3

IN 92-50

July 2, 1992 LIST OF RECENTLY ISSUED

NRC INFORMATION NOTICES

Information

Date of

Notice No.

Subject

Issuance

Issued to

92-49

92-48

92-47

92-46

92-45

92-44

92-43

92-42

Recent Loss or Severe

Degradation of Service

Water Systems

Failure of Exide Batteries

Intent

ional Bypassing

of Automatic Actuation

of Plant Protective

Features

Thermo-Lag Fire Barrier

Material Special Review

Team Final Report Findings,

Current Fire Endurance

Tests, and Ampacity Cal- culation Errors

Incorrect Relay Used in

Emergency Diesel Generator

Output Breaker Control

Circuitry

Problems with Westing- house DS-206 and DSL-206

Type Circuit Breakers

Defective Molded Phen- olic Armature Carriers

Found on Elmwood Con- tactors

Fraudulent Bolts in

Seismically Designed

Walls

07/02/92

06/29/92

06/23/92

06/22/92

06/18/92

06/09/92

06/01/92

All holders of OLs or CPs

for nuclear power reactors

All holders of OLs or CPs

for nuclear power reactors

All holders of OLs or CPs

for nuclear power reactors

All holders of OLs or CPs

for nuclear power reactors.

All holders of OLs or CPs

for nuclear power reactors.

All holders of OLs or CPs

for nuclear power reactors.

All holders of OLs or CPs

for nuclear power reactors.

All holders of OLs or CPs

for nuclear power reactors.

OL = Operating License

CP = Construction Permit

.

l

{

IN 92-50

July 2, 1992 discontinuities, and the orientation of the cracks.

However, the licensee did