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Insp Rept 50-302/90-32 on 900918-1005.Violations Noted. Major Areas Inspected:Events Resulting in Breach of Reactor Bldg Containment & Lack of Leak Integrity of Encapsulation Assembly Providing Secondary Boundary to Decay Heat Sys
	ML20058B089
Person / Time
Site:	Crystal River
Issue date:	10/17/1990
From:	Crelenjak R, Holmesray P; NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:	;
Shared Package
ML20058B085	List: ML20058B081; ML20058B089;
References
	50-302-90-32, NUDOCS 9010290390
	Download: ML20058B089 (6)
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Report No:

50-302/90-32

Licensee:

Florida Power Corporation

3201 34th Street, South

St. Petersburg, FL 33733

Docket No: 50-30'2

License No: DPR-72

facility Name:

Crystal River 3

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InspectionConducte). September 18 - October 5, 1990

1nspector:

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P. lMlmes~~ ay, Wnior ResiKnt Inspector

D6te Sfgned

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Approved by:

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R.7Crlenjak, ~5ettion ChheT

Ddte Ylgned

Division of Reactor Projects

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SUMMARY

Scope:

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This special unannounced reactive inspection was conducted by the tsenior-

resident inspector to review the events resulting in a breach of ~ reactor

building (RB) containment and the events surrounding the lack of leak integrity-

of an encapsulation assembly providing a seccndary boundary to a portion of the

decay heat (DH) system.

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Results:

With the plant in MODE 1, valve DH-43, the first valve outside the RB in the

suction line from the RB sump for the "B" train of DH, was improperly operated

resulting'about 7:00 p.m. the same day.in a breach of_RB containment from about 4:

1990, to

This breach of containment is in

vi61ation of Technical Specification (TS) 3.6.1.1 which requires containment

integrity in MODES 1, 2, 3, and 4.

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The guard assemblyfencapsulating a portion of the DH piping and the first_ valve

immediately outside the RB containment wall was found by the licensee not to be

leak tight as required by Chapter 5 of the Final Safety Analysis Report (FSAR).'

-This lack of' integrity of the encapsulation is a violation of 10 CFR 50,

Appendix B, Criterion III which requires that design measures shall provide for

verifying or checking the adequacy of design.

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REPORT DETAILS

1.

Persons Contacted

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Licensee Employees

J. Albr.rdi, Manager Nuclear Site Support

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W. Bandhaver, Nuclear Operations Superintendent

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G. Boldt, V4,ce President Nuclear Production

J. Co'.oy, Nuclear Principal Mechanical Engineer

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P. V,cKee, Director, Nuclear Plant Operations

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S. Robinson, Nuclear Chemistry and Radiation Protection Superintendent

V. Roppel, Manager, Nuclear Plant Maintenance

W. Rossfeld, Manager, Nuclear Compliance

F. Sullivan, Manager, Nuclear Plant Systems Engineering

M. Williams, Nuclear Regulatory Specialist

K. Wilson, Manager, Nuclear Licensing

Other licensee employees contacted included operations, engineering

maintenance and corporate personnel,

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Attended exit intery, w

Acronyms and initialisms used throughout this report are listed in the

last paragraph.

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2.

ContainmentBreachFollowup(93702)

On September 18, 1990, with the reactor op(erating at full power, theES) train out of

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licensee took the "B" Emergency Safeguards

perform planned maintenance on DHV-40 which had a seat leak (See

Attachment 1 for a diagram of a portion of the DH system).

Prior to the

repair,.DHV-43, a containment isolation valve, was cycled to allow water

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in the pipe below DHV-40 to drain back into the RB sump.

DHV-43 was being

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operated from the MCC prior to placing a clearance tag on the breaker for

DHV-43.

This process was accomplished at approximately 4:30 a.m. and the

breaker to DHV-43 was opened when the position indicating light indicated

the valve was closed.

A drain valve and- a vent valve located in the

auxiliary building were open to allow the remainder of the piping to drain

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to the auxiliary building sump.

At about 6:45 p.m., BSV-58, a one inch demineralized water valve to the DH

pump, was remcVed by maintenance personnel to perform maintenance. When

the valve was removed air flow was observed by maintenance personnel

emanating from the piping.

The air flow was immediately stopped with

plastic and tape.

Air flow was also found emanating from the DH vent

valve which had been opened earlier.

The SS00 then ordered DHV-43 to be

manually closed and the air flow ceased at 7:00 o.m.

It was concluded

that DHV-43 had been partially open permitting air to flow from the RB

sump to ' the auxiliary building.

The operator who closed the valve

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reported that, approximately four or five turns of the handwheel were

required to fully shut the valve.

The operator could not be positive as

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to the number of turns to close the valve because some turns were made

without the handwheel being engaged on the valve stem.

DHV-43 was

declared inoperable by the licensee at about 6:45 p.m. on September 18,

1990, and TS 3.6.3.1.b action statement was entered which states,

"b.

Isolate each affected penetration within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months by use of at least one

deactivated automatic valve secured in the isolation position or, d. Be in

at least HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTDOWN within

the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months ."

DHV-43, although not an automatic valve, was

manually shut about 7:00 p.m. September 18, 1990, and had been deactivated

at about 4:30 a.m. on September 18, 1990.

The valve was declared operational and the TS action statement was exited

on September 20, 1990, af ter the valve position indicating light limit

switch was adjusted.

The limit switch activates the valve position

indicator lights on the main control board and at the MCC.

The licensee

speculates that the valve travel was interrupted (prior to the valve being

fully closed) when the operator opened the valve breaker upon illumination

of the closed valve position light.

Apparently, this closed position

light illuminated prior to the valve travel being stopped by the torque

switch (fullyclosed).

Subsequent to the event, the licensee was able to estimate the amount

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DHV-43 was open by calculating-the orifice size based. on. the recorded

level decrease from the RB sump during the period + hat the RB sump was

draining to the auxiliary building.

Based on this information, the

licensee determined that DHV-43 was about two handwheel turns open (850

turns is full open).

The licensee's calculations show that two handwheel

turns open would allow a flow of 119 SCFM of air and 81.8 gpm of water at

design pressure of 53.9 psig.

These calculations show that containment

leakage requirements specified in TS 3.6.1.2 would have been exceeded.

DHV-43 is a 14 inch, motor operated, gate valve which is normally shut

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during operation but has a post accident position of open to provide a

suction from the RB sump to the B train DH pump in recirculating phase of

the accident.

This event was reported to the NRC duty officer on September 20, 1990,

-af ter the licensee's calculations showed that a breach i ' containment had

occurred.

With the reactor operating in MODE 1, the established flow path from the

RB sump to the auxiliary building through DHV-43 and the DH system vent -

and drain valves is a breach of RB containment and a violation of TS 3.6.1.1.

This condition existed for about fourteen and one half hours.

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Violation - (50-302/90-32-01):

Breach of containment in violation of TS 3.6.1.1.

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3.

Decay Heat System Guard Pipe followup (93702)

On September 24, 1990, with the reactor at full power, the " area owner"

was making a walkdown of the "B" decay heat pit.

He noticed that the

guard encapsulation for DHV-43 wn not leak tight in that two small

diameter pipe nipples did not contain plugs.

The area owner, an

operations assistant shif t supervisor, wrote a memorandum to engineering

requesting information as to whether the nipples should be open or

plugged.

On September 25, 1990, the system engineer accompanied by a

senior reactor operator examined the guard enclosure for both DH trains

and found the plugs missing on both DH train encapsulations.

The system

engineer referred operations to the FSAR chapter 5 which states:

"The Decay Heat System Reactor Building Sump penetrations contain

only one motor operated isolation valve 4cated outside containment.

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These valves are required to open post-LOCA for switchover to

recirculation mode emergency core cooling."

" Penetrations 345 and 346 have only one containment isolation valve

each. The second barrier, which is required in order to meet the CR3

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catainment isolation design basis, is provided by an encapsulation

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aroend sach recirculation line from the containment to beyond the

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'irst 1 solation valve.

This encapsulation is leak-tight at contain-

ment design pressure and is not directly connected to the containment

sump or atmosphere.

A single passive or active failure in these

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lines or encap:ulations will not provide a path for leakage to the

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environment."

The drawing for the eeapsulation (FPC Drawing Number P0-301-621) details

the two socket weld adapters but does not show any attachment or closure

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devices.

Therefore, the encapsulations are in accordaace with the

drawings.

There is no test data, readily available, to indicate that the

encapsulation is capable of withstanding containment design pressure. The

licensee has pluggad the openings but has not tested the encapsulation.

10 CFR 50, Appendix B, Criterion III states:

"The design control measures

shall provide for verifying or checking the adequacy of design, such as by

the performnce of design reviews, by the use of alternate or simplified

calculation methods, or by the performance of a suitable testing program."

The condition existing in the plant, with the encapsulations for both

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trains of the' DH system in accordance with the drawings but not in

accordance with the design basis, indicates that the design control

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measures required by 10 CFR 50, Appendix B, Criterion til were not

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effective and represent a violation of this Criterion.

Violation (50-302/90-32-02):

Failure to provide adequate design control

measures as required by 10 CFR 50, Appendix B, Criterion 111.

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4.

ExitInterview(30703)

The inspector met with licensee representatives (denoted in paragraph 1)

at the conclusion of the inspection on October 9,1990.

During this

meeting, the inspector summarize; the scope and findings of the inspection

as they are detailed in this repu t.

The licensee representatives acknowledged the inspector's comments and did

not identify as proprietary any of the materials provided to or reviewed

by the inspectors during this inspection.

5.

Acronyms and Abbreviations

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BSV - Building Spray Valve

CFR - Code of Federal Regulations

DH

- Decay Heat System

DHV - Decay Heat Valve

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ES

- Emergency Safeguards

FPC - Florida Power Corporation

FSAR - Final Safety Analysis

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GPM - Gallons per Minute

LOCA - Loss of Coolant Accident

MCC - Motor Control Center

RB

- Reactor Building

SCFM - Standard Cubic feet per Minute

SSOD - Shift Supervisor on Duty

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TS

- Technical Specifications

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ML20058B089

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