Information Notice 2000-09, Steam Generator Tube Failure at Indian Point Unit 2

Steam Generator Tube Failure at Indian Point Unit 2

ML011930011
Person / Time
Site:	Indian Point
Issue date:	06/28/2000
From:	Marsh L B Operational Experience and Non-Power Reactors Branch
To:
References
FOIA/PA-2001-0256 IN-00-009
Download: ML011930011 (5)

cc: Holody Urban UNITED STATES Nick NUCLEAR REGULATORY

COMMISSION

OFFICE OF NUCLEAR REACTOR REGULATION

WASHINGTON, D.C. 20555-0001 June 28, 2000 NRC INFORMATION

NOTICE 2000-09: STEAM GENERATOR

TUBE FAILURE AT INDIAN POINT UNIT 2

Addressees

All holders of operating

licenses for nuclear power reactors except those who have ceased operations

and have certified

that fuel has been permanently

removed from the reactor vessel.

Purpose

The U.S. Nuclear Regulatory

Commission (NRC) is issuing this information

notice to inform addressees

of a steam generator

tube failure at Indian Point Unit 2. NRC investigations

of the licensee's

steam generator

inspection

program are ongoing and any potentially

generic issues identified

will be communicated

in a separate generic communication.

However, the investigations

to date re-emphasize

the importance

of licensee involvement

with ongoing industry efforts to understand

and detect steam generator

degradation.

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 February 15, 2000, at 7:17 p.m., the Indian Point Unit 2 nuclear plant experienced

a steam generator

tube failure,, which required the declaration

of an Alert at 7:29 p.m., and a manual reactor trip at 7:30 p.m. The operators

identified

that the #24 steam generator

was the source of the leak and completed

isolation

of the #24 steam generator

by 8:31 p.m. At 9:02 p.m., the operator opened the high-pressure

steam dump valves and established

an excessive

primary plant cooldown rate that caused a rapid reduction

in the pressurizer

level and required the operators

to manually initiate safety injection.

The operators

reset the safety injection

at 9:21 p.m., reduced the reactor coolant system pressure to about 970 psig at 9:32 p.m., and re-commenced

a plant cooldown at 11:35 p.m. The residual heat removal (RHR) system was placed in service on February 16, 2000, at 12:38 p.m., and primary plant pressure was reduced below the #24 steam generator

pressure to terminate

the steam generator

tube leakage at 2:20 p.m. The plant cooldown continued, and the plant entered cold shutdown at 4:57 p.m. The licensee exited the Alert at 6:50 p.m. The NRC sent an Augmented

Inspection

Team (AIT) on February 18, 2000, to review the causes, safety implications, and licensee actions associated

with the event. The AIT developed

a sequence of events, determined

the risk significance

of the event, and assessed the response by the plant staff and management.

The cause of the tube failure was outside the scope of this inspection

and is currently

being reviewed separately

by the NRC. The AIT's report is presented

in Inspection

Report 05000247/2000-02, dated April 28, 2000 (Accession

Number ML003710036).

Discussion

The event was risk significant.

It involved a steam generator

tube failure that resulted in an initial primary-to-secondary

leak of reactor coolant of approximately

146 gallons per minute and required an "Alert" declaration (the second level of emergency

action in the NRC-required

emergency

response plan). The event resulted in a minor radiological

release to the environment

that was well within regulatory

limits. No radioactivity

was measured offsite above normal background

levels, and the event did not adversely

impact the public health and safety. The licensee performed

the necessary

actions to protect the health and safety of the public. Specifically, the operators

promptly and appropriately

took those actions in the emergency

operating

procedures

to trip the reactor, isolate the affected steam generator, and depressurize

the reactor coolant system. Additionally, the necessary

event mitigation

systems worked properly.

Notwithstanding

the above actions, the AIT identified

performance

problems in several broad areas that challenged

operators, complicated

the event response, delayed achieving

the cold shutdown condition, and affected the radiological

release. The problems involved operator performance, procedure

quality, equipment

performance, technical

support, and emergency

response.

Operator Performance

Some operator performance

problems were noted during the plant cooldown phase involving

the following:

While attempting

to cool down the reactor coolant system (RCS), the reactor operator initiated

an excessive

cooldown rate that exceeded procedural

and Technical

Specification

limits. The excessive

cooldown led to several conditions

that complicated

the subsequent

event response and delayed the RCS cooldown.

Operators

were slow to recognize

configuration

lineup problems that (1) prevented

successful

operation

of the auxiliary

spray system to lower RCS pressure and (2) delayed heatup of the RHR system. Procedure

Quality The procedures

adequately

guided the initial operator response;

however, several procedure

problems were identified

that delayed the cooldown and depressurizing

of the RCS. Procedure

deficiencies

affected Standard Operating

Procedures, Emergency

Operating

Procedures, and Emergency

Plan Implementing

Procedures.

Specific activities

included initiation

of RHR cooling, initiation

of component

cooling water alignment, use of auxiliary

pressurizer

spray, use of methods to monitor RCS temperature

to maintain cold shutdown conditions, and initiation

of emergency

response organization (ERO) notifications.

Station personnel

were previously

aware of the procedure

issue involving

initiation

of RHR cooling but had not corrected

the problem before this event. Equipment

Performance

The necessary

event mitigation

systems, including

the reactor protection

system, the auxiliary

feedwater

system, and the safety injection

system, functioned

properly.

However, several longstanding

equipment

performance

problems were identified

that challenged

operators

during this event: Two losses of condenser

vacuum resulted from problems with the operation

of the automatic

steam supply pressure control valve to the steam jet air ejectors, and the #22 condenser

vacuum pump. The isolation

valve seal water system became inoperable

during the event and required operator action and an entry into a Technical

Specification

Limiting Condition

for Operation

Action Statement.

A containment

entry was required to install a temporary

nitrogen supply to the pressurizer

power-operated

relief valve to compensate

for a design deficiency.

This action was required before placing the overpressure

protection

system in service.

The steam generator

leak rate monitoring

equipment

had been degraded for an extended period, and limited the amount of steam generator

leak rate information

available

to the operators

before the event. The AIT determined

that the number and duration of the equipment

problems reflected

weaknesses

in engineering, corrective

action processes, and operational

support at the station.

The licensee's

response to a number of the equipment

problems identified

during the event reflected

an acceptance

of "working around" the problem rather than fixing it. Emergency

Response The ERO took the necessary

steps to ensure the protection

of public health and safety. The operators

properly classified

the event, and the licensee implemented

a thorough peer review of the emergency

response to this event. The AIT identified

several emergency

plan and implementing

procedure

problems similar to those identified

by the licensee's

peer review team, including

the following:

The emergency

response staff was slow to activate the emergency

facilities.

• The licensee was slow to establish

accountability (i.e., identify the location)

of emergency

response personnel. The emergency

response data system -(ERDS) was inoperable

for the first several hours of the event as a result of a pre-existing

equipment

problem.

Problems were noted in the implementation

of the media response plan. Problems were identified

involving

the timeliness

and quality of technical

support provided to the operators.

The licensee developed

and was in the process of implementing

an emergency

response improvement

plan before the event. This information

notice requires no specific action or written response.

However, recipients

are reminded that they are required to consider industry-wide

operating

experience (including

NRC information

notices) when practical

when setting goals and performing

periodic evaluations

under Section 50.65, "Requirements

for monitoring

the effectiveness

of maintenance

at nuclear power plants," of Part 50 of Title 10 of the Code of Federal Regulations.

If you have any questions

about the information

in this notice, please contact the one of the technical

contacts listed below or the appropriate

Office of Nuclear Reactor Regulation (NRR) project manager.

/Ledyard B. Mars , Chief Events Assessment, Generic Communications

and Non-Power

Reactors Branch Division of Regulatory

Improvement

Programs Office of Nuclear Reactor Regulation

Technical

contacts:

Eric Benner, NRR Lawrence Doerflein, Region I 301-415-1171

610-337-5378 E-mail: eibl0.nrc..qov

E-mail: ltdOnrc.gov

Peter Eselgroth, Region I Raymond Lorson, Region I 610-337-5234

603-474-3589 E-mail: pwemnrc.aov

E-mail: rklO.nrc.aov

Attachment:

List of Recently Issued NRC Information

Notices

Attachment LIST OF RECENTLY ISSUED NRC INFORMATION

NOTICES Information

Date of Notice No. Subject Issuance Issued to 2000-08 Inadequate

Assessment

of the 5/15/2000

All holders of operating

licensees Effect of Differential

Temperatures

on Safety Related Pumps National Institute

for Occupational

Safety and Health Respirator

User Notice: Special Precaustions

for Using Certain Self-Contained

Breathing

Apparatus

Air Cylinders

Offsite Power Voltage Inadequacies

Recent Medical Misadministrations

Resulting

from Inattention

to Detail 1999 Enforcement

Sanctions

for Deliberate

Violations

of NRC Employee Protection

Requirements

High-Efficiency

Particulate

Air Filter Exceeds Mass Limit Before Reaching Expected Differential

Pressure Failure of Criticality

Safety Control to Prevent Uranium Dioxide (U0 2) Powder Accumulation

4/10/2000

3/22/2000

3/06/2000 2/25/2000 2/22/2000

2/22/2000 for nuclear power reactors All holders of operating

licenses for nuclear power reactors,non

power reactors, and all fuel cycle and materiallicensees

required to have an NRC-approved

emergency

plan All holders of operating

licenses for nuclear power reactors, except those who have permanently

ceased operations

and have certified

that fuel has been permanently

removed from the reactor All medical licensees All NRC licensees All NRC licensed fuel-cycled

conversion, enrichment, and fabrication

facilities

All NRC licensed fuel-cycled

conversion, enrichment, and fabrication

facilities

OL = operating

License CP = Construction

Permit 2000-07 2000-06 2000-05 2000-04 2000-03 2000-02