Text

Insp Repts 50-254/97-27 & 50-265/97-27 on 971103-07 & 1204. Violations Noted.Major Areas Inspected:Maint & Engineering
	ML20197F574
Person / Time
Site:	Quad Cities
Issue date:	12/23/1997
From:	; NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:	;
Shared Package
ML20197F496	List: ML20197F492; ML20197F543; ML20197F574;
References
	50-254-97-27, 50-265-97-27, NUDOCS 9712300247
	Download: ML20197F574 (13)
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U.S. NUCLEAR REGULATORY COMMISSION

REGION lli

Docket No:

50-254;50 265

Lloonse No:

DPR-29; DPR 30

Report No:

50-254/97027(DRS); 50-265/97027(DRS).

Licensee:

Commonwealth Edison Company

Facility:'

Quad Cities Nuclear Power Station Unit 1 and 2

Location:

22710 206th Avenue North

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Cordova,IL 61242

Dates:

November 3-7,1997 and December 4,1997

Inspector:

M. Holmberg, Reactor Inspector

Approved by:

J.- A. Gavula, Chief

Engineering Specialists Branch 1

Division of Reactor Safety

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9712300247 971223

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EXECUTIVE SUMMARY

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Quad Cities Nuclear Power Station, Unit 1 and 2

NRC Inspection Report 50-254/97027; 50-265/97027 -

This nonroutine inspection focused on the Unit 2 American Society of Mechanical Engineers

(ASME) Code Class 1 leakage test performed at pnwer on June 22,1997. Additionally, this

inspection included a review of ASME Code Class 1 and 2 pressure testing activities completed

in previous outages.

Maintenance

An apparent violation of 10 CFR 50, Appendix G was identified for failure to complete

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the Unit 2 ASME Code,Section XI, Class 1 pressure test prior to criticality, which

demonstrated a fundamentallack of safety focus and knowledge / understanding of the

basis for this testing and the applicable regulatory requirements (M1.1).

An apparent violation of Technical Specification (TS) 4.0.E (with five examples) was

identified for failure to meet or adequately complete ASME Code,Section XI, Class 1

and 2 pressure tests, which indicated a programmatic breakdown in implementation of

these requirements (M1.1).

One violation of 10 CFR 50, Appendix B, Criterion XVil was identified for failure to

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maintain retrievable copies of completed Class 1 pressure tests performed during the

1995 refueling outages for Unit 1 and Unit 2 (M8.1).

One violation of 10 CFR 50, Appendix B, Criterion XVI was identified for failure to

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promptly initiate the corrective action process for the Unit 1 failed reactor vessei flange

inner O-ring seal (M8.2).

Enaineerina

An example of an apparent violation of 10 CFR 50.59(b)(1) was identified for failure to

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perform an adequate safety evaluation of the Unit 2 ASME Code,Section XI, Class 1

leakage test (at power). The inadequate safety evaluation screening performed for this

test was of particular concem, since operation of Unit 2 prior to completing this test

constituted an apparent unreviewed safety question (E2.1).

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Report Details

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IL Maintenan.ge _ _

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M1!

Conduct of Maintenance

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M,1,1; 08=== 1 and 2 Pr==we Testino

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Inapaction Scope (73753. 73755. 73052)

The inspector reviewed' Class 1 system pressure tests for the Units 1 and 2 refueling

outage 14 and problem identification forms (PlFs) related to ASME cme pressure

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The inspector reviewed the Unit 2 drywell access records for June 22,1997 and

= Interviewed two of the three licensee VT-2 inspection personnel involved in the VT-2 --

- examination of Class 1 systems in the drywell on June '22,1997.

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. b.

Observation and Findings

The Unit i recirculation system welds (Code Class.1 system) have a history of

significant intergranular stress corrosion cracking (IGSCC) (NRC report 50-254/96004,

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50 265/96004).- Based on the type of material and similar welding process, the Unit 2

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recirculation system welds are also susceptible to IGSCC. Further, only a sample of the'

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Unit 2 recirculation system welds are ultrasonically inspected at each refueling outage to

L detect this cracking. . A defense in depth measure to detect pressure boundary leakage

(including the potential for leakage caused by recirculation system weld cracking not

detected in ultrasonic weld inspections) are the ASME Code,Section XI, Class 1 and 2

leakage tests. Faliure to properly implement Code Class 1 and 2 leakage test

- requirements (as illustrated by examples discussed below), could result in loss of this

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defense-in-depth measure. . Further, loss of this defense-in-depth measure could pose

an increased risk for pressure boundary leakage or failure at power, which would

challenge operators and/or plant engineered safety features.

- b.1 L ; Unit 2 Cl=== 1 Pressure Testina Critical-

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10 CFR 50 Appendix 'G, IV.2(d), required " Pressure tests and leak tests of the reactor.

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. vessel that are required by Section XI of the American Society for Mechanical Engineers

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(ASME) Code must be completed before the core is critical." .The ASME Code Section

XI, Table IWB-2500-1, examination category B-P,-item B15.10 required a system

leakage. test (lWB-5221) and visual VT-2. examination of the pressure retaining

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~ boundary of the reactor vessel each refueling outage.

-- On May 3T1997,~ the licensee used portions of procedure QCO3 0201-08 to perform a

inon-Code leak check of the reactor vessel and Class 1 piping (with the reactor

, shutdown)c The licensee could not credit this test toward meeting Code pressure test

requirements, as it was conducted at a lower pressure (825-900 pounds per square inch

gage (psig)) than the normal operating system pressure (1000 psig) required to meet -

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On June 22,1997, the licensee performed test procedure OCOS 0201-10 * Reactor

Vessel and Clast One Piping Leak Test at Power Operation," Revision 1 with tne

reactor at about 12 percent power, following Unit 2 criticality and startup from refueling

outage 14. This test accomplished the ASME Code,Section XI requ; red leakage test of

the reactor vessel and Class 1 system boundary for Qued Cities Unit 2. On October 1,

1997, the licensee iden9fied (documented in PlF Q1997-03694) that contrary to the

requirements of 10 CFR 50, Appendix G, this leakage test had not been completed prior

to core criticality. Fallure to complete the Code required reactor vessel and Class 1

system leakage test prior to reactor criticality is an apparent violaton of 10 CFR 50,

Appendix G, IV.2(d) (eel 50-265/97027-01(DRS)).

The liconsee staff responsible for the testing described above, demonstrated a

fundamentallack of knowledge / understanding of the regulatory requiruments as

evidenced by PIF Q1997-04157 * Station Not Notified of 10 CFR 50 Appendix G Rule

Change." In this PIF, licensee staff documented that corporate Commonwealth Edison

staff had failed to apprise the on site lleensee's staff of the applicable regulatory rule

changes and requirements, and that this had created the situation discussed above.

However, the inspecto' considered that licensee on site staff lacked appropriate safety

focus and a fundamental understanding of the intent / basis of the Code Class i system

testing requirements (e.g. that the purpose of the Class 1 pressure test was to

reestablish / verify the !ntegrity of the Class 1 system boundary disturbed by outage

maintenance work activities and as such must be completed odor to operating the

plant),

b.2

6bbreviated Unit 2 Presitgg Testina VT-2 Examinations

The inspector identified the following issues with respect to the VT 2 Inspections

conducted % the Unit 2 drywell during the June 22,1997 Class 1 system leakage test.

The VT-2 examinations were conducted in a very short time; one VT 2 inspector

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completed 93 examination areas in 25 minutes, a second VT-2 inspector

completed 50 examination areas in 12 minutes, and a '5lrd VT 2 inspecto;

completed 69 examination areas in 25 minutes.

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The two VT 2 examination personnelInterviewed by the inspe: tor had not

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utilized mirrors nor system drawings in conducting the VT 2 examinations within

the drywell.

This issue is considered an unresolved item (URI), pending review of information that

demonstrates how all accessible surfaces and insulation joints of Class 1 piping systems

were examined to the VT-2 inspection requirements in the elapsed times and with the

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methods used during this test (URI 50 265/97027-02(DRS)).

b.3

Incomntete Unit 2 Pressure Testina VT-2 Examinations

Technical Specification 4.0.E, required in part * Inservice inspection of ASME Code

Class 1,2,3 components....shall be performed in accordance with Section XI of the

ASME Boiler and Pressure Vossel Codo..." (he following sections discuss examples of

nonconformance with applicable ASME Code,Section XI requirements and, as such,

constitute violations of TS 4.0.E.

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b.3.1 The 1989 Edit;on of the ASME Code Section XI, IWA 5242(a) * Insulated Components *

required that *For other components, visual examination VT 2 may be conducted

without the removal of insulation by examining the accessible and exposed surfaces and

joints of the insulation. Essentially vertical surfaces of insulation need only be examined

at the lowest eleva' ion where leakage may be detectable." IWA 5242(b) * Insulated

Components * required that "When examining insulated components, the examination of

surrounding area (including floor areas or equipment surfaces located underneath the

components) for evidence of leakage, or other areas to which such leakage may be

channeled, shall be required."

The inspector determined the following information based on interviews with the

licensee VT P. qualified inspector that performed Code VT 2 examinations of the Unit 2

reactor vessel head area on June 22,1997. The VT 2 Intpection of the reactor vessel

head area in the refueling cavity was performed by making a visual inspection from the

two accese hatches in the fourth level of the drywell. No physical walk down of the

refueling cavity floor had been performed to examine the lower edge of the vertical

insulation wall surrounding the vessel head. Hence, for this examination, approximately

one hali the circumference of the vertical insulation wall at radial locatione (described in

Attachment A) had not been inspected. Additionally, the access port in this vertical

insulation wall had not been removed to allow direct VT 2 inspections of the head. The

inspector identified that this head area examination was inadequate, in that the

examination failed to include the lower edge and floor areas of the refueling cavity at

radial locations along the vertical head insulation wall and the lic9nsee failed to utilize an

access port in the insulation wall to perform a direct inspection of the vessel head (IWA-

5242(a) & (b)). Failure to meet Code requirements for the VT 2 examination of the Unit

2 reactor vessel head area,is considered an example of an apparent violation of TS 4.0.E (eel 50 265/97027-03(a)(DRS)).

b.3.2 The 1989 Edition of the ASME Code Section XI, IWA 5243 * Components With Leakage

Collection Systems' required that * Where leakages from components are normally

expected and collected (such as valve stems, pump seals, or vessel flange gaskets) the

visual examination VT-2 shall be conducted by verifying that the leakage collection

system is operative."

The inspector identified that test procedure QCOS 0201 10? Reactor Vessel and Class

Orie Piping Leak Test at Power Operation," completed on June 22,1997, was

inadequate, because the licensee failed to perform the Code required VT-2 examination

of the Unit 2 reactor vessel head flange joint. Specificall , the VT-2 examination

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porto.med failed to verify the absence of leakage from !ne vessel flange, as monitored

and collected by the reactor pressuru vessel flange sealleakage detection system.

Further, the VT 2 examination failed to verify th* this system was operative (lWA 5243).

Failure to complete the Code VT-2 examination of the Unit 2 reactor vessel flange is

considered an example of an apparent violation of TS 4 A 5 (eel 50 265/97027-

03(b)(DRS)).

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b.4

incomotete Unit i Pressure Test VT-2 ExaminatiOD

The Inspector identified that test procedure OCOS 0201-08 * Reactor Vessel and Class

One Piping Leak Test," completed on May 3,1996, was inadequate, because the

licensee failed to perform a Code required VT-2 examination of the Unit i reactor vessel

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head flanga Joint. Specifically, the VT 2 examination failed to verify the absence of

leakage from the vessel flange, as monitored and collected by the reactor pressure

vessel flange seat leakage detection system. Further, the VT-2 examination failed to

verify that this system was operative (IWA 5243). Failure to complete the Code VT 2

examination of the Unit i reactor vessel flege is considered an example of an apparent

violation of TS 4.0.E (eel 50-254/97027-03(c)(DRS)).

b.5

MJssed Pressure Testina of a Class 2 System for Unit 1 and 2

The ASME Code Section XI Table IWC-2500 Category C H required a pressure test

(IWC-5221) and VT 2 examination of the pressure retalrdng boundary of this system

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within the Code inspection period. The NRC safety evaluation issued September 15,

1995 approved a licensee Code relief request PR-02 * Examination Category C H,

Prassure Testing of the Reactor Pressure Vessel Head Fla7ge Seal Leak Detection

System.' This Code relief, allowed VT 2 examination of this system to detect gross

leakage with the reactor vesselin a floodup ':ondition (e.g. during a refueling outage)in

lieu of the Code leakage test (lWC-5221).

On November 7,1997, the licensee identified (in PlF Q1997-04295) that a leakage test

and VT 2 examination of the Unit 1 and 2 reactor pressure vessel head flange sealleak

detection system (a Code Class 2 system) h:d not been completed during the required

Code period as discussed below,

b5.1 For the Unit 1 ASME Code Section XI, third interval first period, which ended on

February 18,1996, the licensee failed to perform a leakage test on the vessel head

f'ange sealleak detection system nor had relief request PR-02 requirements been met

(Table IWC 2500). Failure to perform a Code Class 2 system leakage test (lWC-5221

or Code relief PR-02) of the Unit 1 reactor pressure vessel head flange sealleak

detection system within the required Code period is considered an example of an

apparent violation of TS 4.0.E (eel 50 254/97027-03(d)(DRS)).

b5.2 For the Unit 2 ASME Code Section XI, third interval first period, which ended on

March 10,1996, the licensee failed to perform a leakage test on the vessel head flan 00

sealleak detection system nor had relief request PR-02 requirements been met (Table

IWC-2500). Failure to perform a Code Class 2 system leakage test (IWC 5221 or Code

relief PR-02) of the Unit 2 reactor pressure vessel head flange sealleak detection

system within the required Code period is considered an example of an apparent

violation of TS 4.0.E (eel 50 265/97027-03(e)(DRS)).

c.

Conclusions

An apparent violation was identified that per*ained to the Unit 2 ASME Section XI Code

Class 1 pressure test performed at power, which demonstrated a fundamental lack of

licensee staff safety focus and knowledge / understanding of the basis for this testing and

ti.e applicable regulatory requirements (M1.1).

A defense in depth measure to detect potential pressure boundary leakage (e.g.

recirculation system weld cracking)is the ASME Code,Section XI, Class 1 and 2

leakage test. An apparent violation was identified with fic examples that pertained to

missed or inadequate completion of ASME Ccde Section XI requirements for Class 1

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and 2 pressure tests, which Indicated a programmatic breakdown in implementation of

these requirements. This issue is of concem, because the failure to adequately

implement Code testing requirements for Class 1 and 2 systems could result in an

inr:reased risk for pressure boundary leakage or failure at power, which would challenge

operators and/or plant engirieered safety features.

M8

Miscellaneous Maintenance issues

M8.1 Unicidayable Test Records

The inspector requested copies of Class i leakage tests performed during the Unit 1

and Unit 21995 refueling outages (Q2R13 and Q1R13). The licensee reportedly found

checklists associated with this testing, but was unable to locate copies of the comphted

Class 1 pressure tests and, on November 7,1997, issued PIF Q1997-04263 to

document this issue.10 CFR 50, Appendix B Criterion XVil * Quality Assurance

Records' requires in part that sufficient records shall be maintained to fumish evidence

of activities affecting quality. These records include records ofinspections and tests.

Further these records shall be identifiable and retrievable. As of December 4,1997, the

licensee had not located these records. Fellure to maintain retrievable records of these

completed Class 1 pressure tests is consbered an example of a violation of 10 CFR 50,

Apper<ix B, Criterion XVil (VIO 50 254(265)/97027-04(DRS)).

M8.2 UD'd Reactor Vessel Flange 0-Ring Fallure

Quad Cities UFSAR Section 5.2.5.4.2 stated that "A pressure switch will alarm if failure

of the inner 0 ring takes place on the reactor vessel." The inspector identified, that the

licensee had operated from August 31,1996, thiough November 6,1997, with the Unit i

reactor vessel head flange leak detection system pressurized above the alarm setpoint

(640 psig), vice at drywell pressure. This condition Indicated that a failure of the reactor

vessel flange inner 0-ring had occurred on or about August 31,1990 and as of

November 5,1997, the licensee had failed to initiate the corrective action process (e.g.

Issue a PlF) for this potential condition adverse to quality. Further, on November 6,

1997, the licensee attempted to open it.e system drain valve (AO 1-0220-51) to relieve

the trapped pressure (710 psig) and the drain valve failed to open. On November 6,

1997, the licensee issued PlF Q1997-04261 to document this condition and initiate the

corrective action process. However, failure to promptly initiate the corrective action

process (e.g. Issue a PlF) for the failed reactor vessel flange inner 0-ring seal is

considered a violation of 10 CFR 50, Appendix B, Criterion XVI (VIO

50 254/97027-05(DRS)).

111. Engineering

E2

Engineering Support of Facilities and Equipment

E2.1

GAEs 1 Leakage Tegt Safetv Evaluation

a.

IDEDCCliQajQQDe (37700)

The inspector reviewed the completed safaty evaluation screenings for the Unit 2 Class

1 system leakage test performed on June 22,1997.

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The inspector reviewed the Quad Cities Updatad Final Safety Analysis Report (UFSAR);

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Section 5.2.4.7 " System Leakage and Hydrostatic Pressure Tests,"

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Observations and Findinas

b.1

Inadeauate Ursit 2 Safety Evaluation

10 CFR 50.59(b)(1) states, in part, that the licensee shall maintain records of changes

and these ret ords must include a written safety evaluation which provides the basis for

the determination that the change does not involve an unreviewed safete question.

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The inspector identified that as of Novemb1r 5,1991 the licensee had failed to perform

an adequate safety evaluation of procedure QCOS 0201 10 * Reactor Vessel and Class

One Piping Leak Test at Power Operation," Rcvision 1, to determine if performance of

this procedure after core criticality and plant startup constituted an unreviewed safety

question. Specifically, the "10 CFR 50.59 Screenir,g for Procedure Changes, Tests, &

Experiments' completed for procedure QCOS 0201 10 Revision 0, authorized June 2,

1997 and for QCOS 0201 10 Revision 1 authorized on Juns 6,1997 were inadequate.

in the 50.59 screening for Revision 1 of procedure QCOS 0201 10, the licensee stated

"The SAR does not describe leak testing of the reactor vessel with the reactor at power.

The SAR will not be affected by this revision." The licensee review of this procedure

lacked an evaluation of the change to the UFSAR Section 5.2.4.7 requirement that

stated ' System leakage and hydrostatic tests are conducted in accordance with IWB-

5000 [of Section XI the ASME Code),.. "(e.g. prior to plant startup). Specifically, IWB

5210(a) required pressure retaining components to be tested in accordance with Table

IWB-2500-1 Examination Category B-P, and Note 5 of this table required that "The

system leakage test (IWB 5221) shall be conducted prior to plant startup following each

refueling outage." Further, operatbn of Unit 2 prior to completing this test constituted an

apparent unreviewed safety question, since tne probability for a loss of coolant accident

(LOCA) had been increased as discussed below. Failure to perform an adequate safety

evaluation for this test procedure is an example of an apparent violation of 10 CFR 50.59(b)(1) (eel 50-265/97027-06(DRS)).

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The licensee had not completed a Code Class 1 system leakage test prior to the Unit 2

startup from the refueling outage, in which the Class 1 system boundary had been

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breached for maintenance and repair work. Therefore, the disturbed / breached Class 1

boundaries which were not fully tested posed an increased risk for leakage or failure

(LOCA) with the plant at power. Unit 2 operated for approximately seven days before

the Class 1 system leakage test was completed on June 22,1997. Since the likelihood

of an accident (LOCA) had been potentially increased during this time frame, an

appropriate safety evaluation would have necessarily concluded that this activity

constituted an unreviewed Srfety ouestion,

c.

Conclusions

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The inspector identified an apparent violation that pertained to the failure to periorm an

adequate safety evaluation (screening) for the Unit 2 Code Class i system leakage test

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at power. This inadequate screening is of particular concem, since operation of Unit 2

prior to completing this test constituted an apparent unreviewed safety question.

V. Management Meetings

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Exit Meeting Summary

The inspector presented the inspection results to members of licensee management at

the conclusion of the inspe ction on November 7,1997 and in a final phone exit held

Decernber 4,1997. The licensee acknowledged the findings oresented and did not

identify any of the potential report input discussed as propr;etary,

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PARTIAL LIST OF PERSONS CONTACTED

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Commonwealth Ealson

L. Pearce

Site Vice President

D. Cook

Station Manager

M. Wayland

Maintenance Manager

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C. Peterson

Regulatory Affairs Manager

T. Peterson -

Regulatory Assurance

- R. Baumer

Regulatory Assurance

T. Wojcik

Engineer

J. Amold

Support Engineer

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R. Fairbank

Engineer

J. Tricorichson

System Engineer

D. Gibson

. Work Control

J. Stortz

Shift Manager

F. Famularl

Quality Assurance Manager

M. Stiener

Quality Assurance

B. Rybak

Licensing

NRC

K. Walton

Resident inspector

C. Miller

Senior Resident Inspector

M. Ring

Branch Chief, Reactor Projects Branch 1

J. Jacobson

Deputy Division Director, Division of Reactor Safety

INSPt!CTION PROCEDURES USED

IP 73753

INSERVICE INSPECTION

lP 73755

INSERVICE INSPECTION - DATA REVIEW AND EVALUATION

IP 73052

INSERVICE INSPECTION - REVIEW OF PROCEDURES

. IP 37700

DESIGN CHANGES AND MODIFICATIONS

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ITEMS OPENED, CLOSED or DISCUSSED

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50 265/97027-01(DRS)

eel

Failure to complete the Unit 2 Code Class 1

leakage test prior to reactor criticality.

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50-265/97027-02(DRS)

URI

Abbreviated Unit 2 VT 2 examinations, may not

meet Code requirements.

50 265/97027-03(a)(DRS)

eel

Failure to perform an adequate VT 2 examination

of the Unit 2 reactor head.

50 265/97027-03(b)(DRS)

eel

Failure to pedorm an adequate VT 2 examination

of the Unit 2 reactor vessel flange.

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50 254/97027-03(c)(DRS)

eel

Failure to perform an adequate VT-2 examination

of the Unit i reactor vessel flange.

50 254/97027-03(d)(DRS)

eel

Failure to perform a Code leakage test of the Unit 1

head flange leakage detection system.

50 265/97027-03(e)(DRS)

eel

Failure to perform a Code leakage test of the Unit 2

head flange leakage detection system.

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50 254/97027-04(DRS);

VIO

Failure to maintain retrievable records of Code

50 265/97027-04(DRS)

Class 1 pressure tests for both Units.

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50-254/97027 05(DRS)

VIO

l' allure to promptly implement the corrective action

process for the failed Unit i vessel head flange

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inner 0 ring seal.

. 50-265/97027-06(DRS)

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Failure to perform an adequate safety evaluation

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for the Unit 2 Class i leakage test at power,

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LIST OF DOCUMENTS REVIEWE0

QCOS 0201-08 * Reactor Vessel and Class One Piping Leak Test' Revision 8 performed in part

on May 23,1997, during the Unit 2 refueling outage 14.

QCOS 0201 10 * Reactor Vessel and Class One Piping Leak Test at Power Operation,'

Revision 1 performed on June 22,1997, following the Unit 2 refueling outage 14.

OCOS 0201-08 * Reactor Vessel and Class One Piping Leak Test' Revision 7 performed May 3,

1996, following the Unit i refueling outage 14.

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PlF Q1997-03694 * Leak Test Done With Reactor Critical," issued on October 1,1997.

PlF Q1997 04157 ' Station Not Notified of 10 CFR 50 Appendix G Rule Change," issued on

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October 30,1997.

PlF Q1997-04228 'RPV Flange VT 2 During Q2R14," issued on November 5,1997.

PlF Q1997-04260 * Reactor Head Seal Leakoff System Operation Does Not Agree With FSAR,"

issued on Novembar 7,1997.

PIF Q1997-04261 * Reactor Head Seal System in Alarm,' Issued on November 7,1997.

PlF Q1997-04266 "Q2R14 VT 2 Exam of Rx [ Reactor) Flange,' lasued on November 7,1997.

PlF Q1097-04263 * Class 1 Gyuem Leakage Test Records Not Retrievable For Q1R13 &

O2R13," issued November 7,1997.

PlF Q1997-04295 'Misced VT-2 Exam RX Head Seal Leak Detection System," issued

November 8,1997.

'10 CFR 50.59 Screening for Procedure Changes, Tests, & Experiments" completed for

procedure OCOS 0201 10 * Reactor Vessel and Class One Piping Leak Test at Power

Operation," Revision 0, authorized on June 2,1997 and Revision 1 authorized on June 6,1997.

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Inspection Re

1

50 254/97027: port

50 265/97027

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Attuchment A

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ML20197F574

Text

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