ML20149J733

Insp Rept 50-482/88-06 on 880125-29.Violation Noted.Major Areas Inspected:Events Associated W/Reactor Pressure Vessel Water Level & RHR Sys & Verification of Containment Integrity & Containment Local Leak Rate Testing

ML20149J733
Person / Time
Site:	Wolf Creek
Issue date:	02/11/1988
From:	Mckernon T, Seidle W, Skow M NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:
Shared Package
ML20149J644	List: ML20149J640 ML20149J693 ML20149J733
References
50-482-88-06, 50-482-88-6, NUDOCS 8802230195
Download: ML20149J733 (7)

See also: IR 05000482/1988006

Text

.

. .. ..

. .. . .

. _.

.. .

.. .. ..

..

.

.

,

.

APPENDIX B f

U.S. NUCLEAR REGULATORY COMMISSION

REGION IV

NRC Inspection Report: 50-482/88-06 Operating License: NPF-42

Docket: 50-482

Licensee: WolfCreekNuclearOperatingCorporation(WCNOC)

P.O. Box 411

Burlington, Kansas 66839 ,

Facility Name: Wolf Creek Generating Station (WCGS)

Inspection At: WCGS, Burlington, Kansas 1

Inspection Conducted: January 25-29, 1988

Inspectors: P_//// Fir 1

M. E. Skow, Reactor Inspector, Test Programs [ Tate

Section

/ I M ~

T. O. McKernon, Reactor Inspector, Test II!

Programs Section

<

Accompanied by W. C. Seidle, Chief, Test Programs Section on January 28-29, 1988

Approved:

' MS/ ~

W. C. Seidle Chief, Test Programs Section Date

Inspection Sumary  ;

'

, Inspection Conducted January 25-29, 1988 (Report 50-482/88-06[

Areas Inspected: Routine, unannounced inspection of events associated with

the reactor pressure vessel water level and the RHR system, verification of

containment integrity, and containment local leak rate testing.

Results: Within the three areas inspected, one violation was identified

(failure to have procedures appropriate to the circumstances, paragraphs 2

and 3).

8802230195 800217

DR ADOCK O y2

.

._ _. . . , . . .

, , . . __ _ __________ - _______ __ _ -________

1 +

.

. .

y

,

2 ,

DETAILS

y 1. Persons Contacted

• B. D. Withers, President

'

,

,-

• F. Rhodes, Vice President, Nuclear Operations ,

• R. M. Grant, Vice President, Quality

• G. D. Boyer, Plant Manager

• M. G. Williams, Superintendent, Regulatory Compliance

• C. E. Parry, Manager, Quality Assurance

!

t *A. L. Payne, Supervisor, Quality Plant Support

• J. S. Allen, ISEG Engineer

• W. M. Lindsay, Supervisor, Quality Systems

• G. W. Reeves, Superintendent Quality Control

• C. J. Hoch, Quality Assurance Technician ,

• 0. L. Maynard, Manager, Licensing ,

• J. W. Johnson, Chief of Security s

• R. K. Steinbrock, Engineer

• A. A. Freitag, Manager, Nuclear Plant Engineering, WCGS

• J. M. Pippin, Manager, Nuclear Plant Engineering

• G. J. Pendergrass, Licensing Engineer

• H. Chernoff, Licensing Engineer

• J. Hcughton, Operations Coordinator

C. G. Patric, Superintendent, Quality Evaluation

R. Sims, Technical Staff Engineer'

L. A. Gabryelski, Technical Staff Engineer

'

Y. MacTaggart, Supervisor, Resdt:! Engineering

K. Petersen, Supervisor, Licensing

M. Estes, Superintendent of Operations

W. Drogemuller, Reactor Operator

R. Ruttler, Maintenance Supervisor '

D. Jacobs, LLRT Coordinator

The NRC inspectors also interviewed other licensee employees during the ,

course of~the inspection.

• Denotes those present_duri,ig the exit interview held on January 29, 1988.

2. Reactive Inspection-Tygon Tube Reactor Pressure Vessel Level Error

Faview(90711B) l

The purpose of this pordon of the inspection was to review the Tygon tube

reactor pressure vessel (RPV) level error event. The purpose was also to

ascertain whether responsibilities have oeen assigned for the review of ,

the event. Additionally, the NRC inspectors were to verify that the

licensee's system fc,r identification and review of the ever,t was

,

functioning. .

At the time of the event, the licensee had lowered RPV cool 0nt level

to approximately 1 foot below the RPV flange in preparation for removing

,

, _ ___ _

.

'-

( i

' '

.

\

'

'

f

c 3 '

s

x

'

the RPV head. The RPV leel instrumentation system (RVLIS) hadebeen'

disconnected also, in preparation N r removing the RPV head. The. licensee

was monitoring water level in the.vessal by visually observin'9 the water

level in a Tygon tube. This tube ran frcm a connection near a rnctor

coolant pump to a connectinn near the top of the pressurizer. Another

tube ran from a tee just beyond the pressurf zer connection of the fird

tube to a connection at the RPV head vent. From the other direction of

the tee, the two tubes were vented via the pressurizer relief tank. The

licensee had performed this partial. system drain using Procedure GEN-00-007,

Revision 8, "Mode 5-RCS Drain Down."

1

The following event description was determined by the NRC inspectors based'

on interviews with licensee personnel. Durin s

attempting to disconnect a "Cono seal" (seal)g fromthe

theevent, a technician

RPV head. When the was

'

seal was disconnected, moisture and gas were. discharged from the vessel

under some pressure. The technician reconnectW the seal. At this time,

the water level in the Tygon tube used to monitor vessel water level

decreased by approximately 20 inches,sThe licensee attcmpted to

vent the vessel by connecting a vent rig downstream from the head vent

valve and opening the vent valve. No' discharge was observed and the

technician disconnected the seal. Gas was discharged a second time from

the seal, and the technician reconnected t'se' seal. A vent rig was then

connected at the RVLIS connection to the he.pd. The head was vented

through this rig. Little pressure appeared to remain, and venting

appeared to have been accomplished. The ; technician then successfully

disconnected the seal. It appeared that the water level in the tube was

in error and high relative to the vessel because of the pressure in the

vessel. While actual vessel water level did not appear to have changed

during this event, it was apparently already lower by 20 inches than the

licensee expected.

The NRC inspector noted that the procedure did not appear to require that

the Tygon tube from the vessel head vent to the Tee be mu itored during

the portion of the drain down when a level would have been visible. This

observation was discussed with the licensee. No irdicat4 u was given that

the licensee had considered and rejected monitcring the drain dcwn from

both tubes. As part of the corrective action, the licenste issued a

change on January 24, 1988, to Procedure GEN-00-007. This change. was to

vent the head via the two means described above that were utilized after

'

the initial "Cono seal" disconnection. The change appeared dequate to

prevent recurrence of this pacticu% r event. The NRC inspector found that

the change did not appear to address the effects of the blockage in

question or other potential blockages on water level indications relative

to actual vessel water level. TM failure to have a procedure appropriate ,

to the circumstances is an apparent violation (482/8806-01).

The root cause of the event had net been determined by the licensee during

this inspection period. The valve 01: the reactor vessel head vent was

suspected by the licensee as having been blocked. The licensee stated

that radiography of the suspect valve in both tM open and shut positionu

,

did not show an.y apparent defects. The efforts by the licensee to

'

l

1

[

. _ _ _ _ -

,

,

'

y , .;.g l

'

' y : ll

'

m

-

4.

determine the root cause(s) of the event appeared to be routine efforts.

-It was not evident to the NRC inspectors that management had become

hN sufficiently involved to assure a complete and timely analysis and

corrective action.

3. Reactive Inspection - CCW Water Hammer Event (907118)

The NRC inspector responded to the licensee's report of a water hamer

event in the CCW system which occurred at 11:40 a.m., January 22, 1988,

during startup of RHR "A" train. The NRC inspector performed a walkdown

inspection of the CCW system in the area of observed piping vibration,

B interviewed licensee The

Reactor Operator (RO) personnel, and reviewed operating procedures.present during th

lasting about 25 seconds. The RO stated that the CCW Valve EG HV-101 had

initially been positioned to the 10 percent open position. Upon subsequent

l

< instruction to throttle the valve, another operator pre w nt throttled

! the valve closed. When the R0 noted the valve in the closed position, the

L valve was opened. Upon noting the water hamer event, the R0 closed the

valve. The RO reported Valve EG HV-101 to have been opened for about

5 seconds. The control room operators began procedures to isolate RHR "A"

'

crain and to bring RHR "B" train on line. A review of control room alarm

'i records indicated the event lasted about 7 minutes having began about

L i 11:40 a.m. and lasting until about 11:47 a.m. at which time RHR "B" train

( was brought operational.

l

f' A system walkdown of the CCW system in the area of observed piping

vibration was conducted by licensee representatives and the NRC inspectors.

. Pipe rub was observed in the area of a whip restraint adjacent to

l Valve EG-HV-101. Indications on the pipe showed possible pipe movement of

1/2- to 3/4-inch. No other observable damage to piping, piping

supports, hangars, snubbers, fillet welds, or piping in wall penetration

areas was noted.

A review of Operations Procedure EJ-120, Revision 9, dated October 2,

1937, showed that instructions for Operating Valve EG-HV-101 were not

clear, concise, and presented in a manner which could have precluded

l erroneous R0 action and inadvertent closure of the valve. On January 22,

l

1988, the licensee revised Operations Procedure EJ-120, Revision 9, to

l

incorporate more specific instruction concerning the throttling of

! Valve EG-HV-101. Further, the revision clarified operation of RrIR Heat

j

Exchanger Bypass Valve EJ-FCV-618 to control RHR system heatup. The

l procedure revision appears adequate and should assist in precluding

l similar future occurrences.

. As followup action to the CCW water hamer event, the licensee had

t.

consulted with Westinghouse Electric Corporation, Power Systems Division,

a concerning the event. A Westinghouse written response, dated January 28,

1988, cited a 1976 study, "Thermal Transient and Fatigue Study of Residual

l Heat Exchanger," Report TM-188, as a basis for analysis. The licensee was

j ) in the process of analyzing the CCW system to ascertain areas of potential

- high stress concentrations. The licensee stated that followup NDE

s

.

._. -

3- -

.-

g.

1.;' ,

5

'

examination of the piping would be based on the analysis results, snubber

inspection, and visual inspection of piping and supports. The NRC

inspector noted that the Westinghouse response stated that, ". . . based

upon the duration of the event and without extenske evaluations thermal

loads and vibrations in the vicinity of the RHR Heat Exchanger would not

be significant and impact the structural integrity of the Heat Exchanger

or piping."

The NRC inspector noted the licensee's Technical Specification (TS)

requires the CCW system operability be verified prior to plant startup. A

further review of the licensee's followup action to verify the CCW

system's operability is warranted. The adequacy of the licensee's

corrective and followup actions to verify system operability shall remain

an open item pending further review (482/8806-02).

In sumary, the NRC inspector noted several centributing factors resulting

in the CCW water hammer event. The operating procedure lacked a quantitative

acceptance criteria governing the throttling of a valve critical to the

system evolution. The procedure did not incorporate cautionary statements

alerting the R0 to the significance of the action statement. In addition,

there appeared to be a miscommunication between the control room and the

R0 performing the procedure. These contributing factors are indicative of

inappropriate instructions and, as such, are an apparent violation of the

Code of Federal Regulations, 10 CFR 50, Appendix B, Criterion V

(482/8806-01).

4. Verification of Containment Integrity (61715)

The purpose of this inspection was to evaluate the adequacy and

implementation of the licensee's procedures designed to ensure and

maintain containment integrity and to mitigate contamination release in

the event of a loss of a coolant accident.

The NRC inspectors reviewed Procedure STS GP-001, Revision 6. "Containment

Penetration Integrity Verification." The specific completed procedures

reviewed were perfonned November 18 and 24,1987, and December 1, 1987.

The procedures appeared to have imposed a weekly verification of containment

penetration integrity during the refueling outage. The procedures

appeared to be in conformance with TS and appeared adequate.

No violations or deviations were identified during this portion of the

inspection.

5. Local Leak Rate Testing Review (61720)

During the inspcction period, the NRC inspector performed detailed reviews

of the following local leak rate test (LLRT) procedures and test results:

• STS PE-014 Revision 5 "Containment Air Locks Tests," to include

LER 87-023-00 dated June 1,1987, "Shaf t Seal on Containment Air Lock

Failure Causing Leakage Greater Than .6La"

'

~, =

,

5

g' ' ' _h

6

.

CKL PE-002, Revision 2,'"Equipment Hatch"

CKL PE-014, Revision 2 "EJ HV-24 and HV-26"

CKL PE-022, Revision 3, "BB-V148 and BB-HV8315B"

CKL PE-028. Revision 4. "EF HV-46, 48, 50," to include LER 87-033-00,

"IN0P Containment Isolation Valve EF HV-46"

CKL PE-080, Revision 4. "BG-8381'and BG-HV-8105" ,

ENG 09-004, Revision 2, "LLRT Acceptance Analysis and Trend Record"

These' reviews were performed to assess technical and administrative

adequacy and conformance to regulatory requirements,10 CFR 50,

Appendix J.

The review of LLRT procedures provided verification that the following

attributes, considered necessary for the conduct of a successful test,

were correctly addressed:

a. All applicable containment penetration boundaries (CPB) and ,

'

containment isolation valves (CIV) are subjected to local leak rate i

testing.

.

b. LLRTs are performed at containment integrated leak rate test peak

pressure, except where reduced pressure tests have received prior NRR

approval in the TS.

c. The LLRT program utilizes approved methods for testing CPB and CIV.

i

d. Penetration' leakage rates are determined using the maximum pathway

leakage,

e. The criteria and response for LLRT and combined leakage rate failure

are incorporated in the test program procedures. ,

f. The criteria and response for the leakage rate failure of components

is specifically cited in the TS. F

, 9 Repairs / modifications to CPB and CIV were preceded and followed by .

'

LLRTs on the applicable penetrations,  !

,

!

,

h. LLRT is performed at the correct frequency for CPB, CIV, and air

i locks.

i

i Within the scope of this inspection, no violations or deviations were

identified.

'

! ,

i

i

l

l

. - -- . _. - . - - _ - - . -_ _ - - .

_ --

,

,

,

' ,

7

6. Exit Interview

The NRC inspectors conducted an exit interview on January 29, 1988, with

the licensee personnel denoted in paragraph 1. The following NRC

personnel were also in attendance:

W. C. Seidle (Chairman)

B. L. Bartlett

R. G. Taylor

L. E. Ellershaw

R. A. Caldwell

At this meeting, the scope and findings of the inspection were summarized.

l

I

l

l

!

l

L

1

r

f

-