Information Notice 1996-32, Implementation of 10 CFR 50.55a(g) (6) (II) (A), Augmented Examination of Reactor Vessel

Implementation of 10 CFR 50.55a(g) (6) (II) (A), Augmented Examination of Reactor Vessel

ML031060052
Person / Time
Site:	Beaver Valley, Millstone, Hatch, Monticello, Calvert Cliffs, Dresden, Davis Besse, Peach Bottom, Browns Ferry, Salem, Oconee, Mcguire, Nine Mile Point, Palisades, Palo Verde, Perry, Indian Point, Fermi, Kewaunee, Catawba, Harris, Wolf Creek, Saint Lucie, Point Beach, Oyster Creek, Watts Bar, Hope Creek, Grand Gulf, Cooper, Sequoyah, Byron, Pilgrim, Arkansas Nuclear, Three Mile Island, Braidwood, Susquehanna, Summer, Prairie Island, Columbia, Seabrook, Brunswick, Surry, Limerick, North Anna, Turkey Point, River Bend, Vermont Yankee, Crystal River, Haddam Neck, Ginna, Diablo Canyon, Callaway, Vogtle, Waterford, Duane Arnold, Farley, Robinson, Clinton, South Texas, San Onofre, Cook, Comanche Peak, Yankee Rowe, Maine Yankee, Quad Cities, Humboldt Bay, La Crosse, Big Rock Point, Rancho Seco, Zion, Midland, Bellefonte, Fort Calhoun, FitzPatrick, McGuire, LaSalle, Fort Saint Vrain, Shoreham, Satsop, Trojan, Atlantic Nuclear Power Plant
Issue date:	06/05/1996
From:	Grimes B Office of Nuclear Reactor Regulation
To:
References
IN-96-032, NUDOCS 9605200277
Download: ML031060052 (9)

4 UNITED STATES

NUCLEAR REGULATORY COMMISSION

OFFICE OF NUCLEAR REACTOR REGULATION

WASHINGTON, D.C. 20555 June 5, 1996 NRC INFORMATION NOTICE 96-32: IMPLEMENTATION OF 10 CFR 50.55a(g)(6)(ii)(A),

"AUGMENTED EXAMINATION OF REACTOR VESSEL"

Addressees

All holders of operating licenses or construction permits for nuclear power

reactors.

Purpose

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

notice to alert addressees to certain aspects of scheduling and implementing

the augmented reactor vessel examination required by Section

50.55a(g)(6)(ii)(A) of Title 10 of the Code of Federal Regulations (10 CFR).

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.

Background

Because of concerns regarding the scope of inspection of reactor vessels, the

NRC issued, in 1992, 10 CFR 50.55a(g)(6)(ii)(A), "Augmented Examination of

Reactor Vessel" [hereinafter referred to as Paragraph (A)], which contains new

requirements for an augmented examination of reactor vessels. The rule

requires licensees to implement, before the time required by normal updating

of the inservice inspection (ISI) program, provisions in the 1989 Edition of

the American Society of Mechanical Engineers, Boiler and Pressure Vessel Code

(ASME Code),Section XI, to examine "essentially 100%" of the length of all

reactor vessel shell welds. Licensees with fewer than 40 months remaining in

the ISI interval that was in effect on September 8, 1992, may defer the

augmented reactor vessel examination to the first period of the next ISI

interval [Paragraph (A)(3)]. "Essentially 100%" examination is defined in

Paragraph (A)(2) as "more than 90% of the examination volume of each weld"

[emphasis added].

Licensees unable to completely satisfy the requirements for the augmented

reactor vessel examination must propose an alternative that would provide an

acceptable level of quality and safety. The proposed alternative may be used

when authorized by the Director of the Office of Nuclear Reactor Regulation

(NRR) [Paragraph (A)(5)].

PDA LA6U2:41-E CE-032 q7c5'5 A

• AIN 96-32 June 5, 1996 The 1989 Edition of the ASME Code,Section XI, incorporated Appendix VIII,

"Performance Demonstration for Ultrasonic Examination Systems." Appendix VIII

was developed to ensure the effectiveness of ultrasonic examinations through a

performance demonstration to evaluate the adequacy of procedures, equipment, and personnel for detecting and sizing flaws during examinations. Licensees

are not currently required to implement Appendix VIII.

Description of Circumstances

It became evident to the staff while it was conducting ISI reviews that some

licensees were unaware of or uncertain about some aspects of the augmented

reactor vessel examination rule.

The staff learned that a small number of licensees were unaware of the rule

and its requirements for some time after it was published. Licensees need to

be aware of the schedular requirements of the rule to ensure timely

implementation of its provisions. Because of the scope and extent of the

examination, significant planning is necessary to address the technical, schedular, and regulatory issues associated with a comprehensive examination

of the reactor pressure vessel.

This information notice contains a discussion of certain areas of

misinterpretation that the staff has dealt with in the implementation of the

augmented reactor vessel examination rule.

Discussion

Schedular Requirements of the Rule

In one instance, a licensee original 10-year ISI interval end date allowed

deferral to the first period of the next interval. However, this licensee

experienced an extended shutdown and, as permitted by Section XI, extended the

ISI interval to complete the examinations required for the interval. As a

result, more than 40 months remained in the interval in effect on September 8,

1992, and the licensee would have been required to do the examination sooner

than expected. The licensee requested and was granted approval by NRR to

schedule the examination in accordance with the original 10-year ISI interval

end date to allow for proper scheduling and to ensure the availability of

examination equipment.

Essentially 100%0 Examination Standard

Most licensees are finding that while the overall average examination coverage

for reactor vessel shell welds may be more than 90%, examination coverage for

individual welds may be substantially less than 90%. When a licensee is

unable to examine "essentially 100%" of each shell weld, it must seek NRC

authorization of an alternative in accordance with Paragraph (A)(5).

During discussions with the NRC staff regarding the review of the 10-year ISI

program plan, a licensee stated that it had obtained "essentially 100%"

K< IN 96-32 June 5, 1996 coverage of the total volume of the reactor vessel shell welds but coverage of

less that 90% of several individual welds. Contrary to the requirements of

the rule, the licensee did not submit a request for authorization of an

alternative to the NRC as required by the rule, until asked to do so by the

NRC.

uSplrit of Appendix VIII" Examination

Section XI contains rules for evaluating the significance of flaws identified

through non-destructive examination. Flaws that are of such size that they

cannot be dispositioned through comparison with code tables must be analyzed

in accordance with Section XI, Paragraph IWB-3600, "Analytical Evaluation of

Flaws." Furthermore,Section XI, Paragraph IWB-3134(b), "Review by

Authorities," requires that analytical evaluations performed in accordance

with Paragraph IWB-3600 be submitted to the regulatory authority having

jurisdiction at the plant site (i.e., NRC).

One licensee administered a "Spirit of Appendix VIII" performance

demonstration for the procedures, personnel, and equipment to be used for the

augmented reactor vessel examination. This type of examination essentially

satisfies the technical requirements of Appendix VIII and would be expected to

yield more accurate and reliable inspection results. The licensee concluded

that the performance demonstration resulted in examination and evaluation

techniques that surpassed the conventional techniques of Section XI of the

ASME Code and Regulatory Guide 1.150, "Ultrasonic Testing of Reactor Vessel

Welds During Preservice and Inservice Examinations." During the augmented

reactor vessel examination, the licensee identified 15 flaws in the shell

welds and in the shell-to-flange weld outside the scope of the augmented

reactor vessel examination, which required analytical evaluation in accordance

with Section XI, Paragraph IWB-3600. The licensee stated that if the

conventional techniques of Section XI and Regulatory Guide 1.150 had been

used, 12 of these 15 flaws would not have even been recordable and only 2 of

the remaining 3 flaws would have required analytical evaluation in accordance

with Paragraph IWB-3600. This licensee experience indicates that flaws of

sufficient size to require analytical evaluation may not be detected when

using conventional techniques for the augmented reactor vessel examination.

Although the licensee in the above example submitted a request for

authorization of an alternative as the examination coverage was less than

"essentially 100%," it did not submit the flaw evaluations, as required by the

ASME Code, until asked to do so by the NRC.

Need for NRC Authorization of Alternatives

A licensee unable to obtain the required examination coverage quoted 10 CFR

50.55a(g)(4) as a basis for not seeking NRC authorization of an alternative as

required by Paragraph (A)(5). However, 10 CFR 50.55a(g)(4) states, in part, that "components. . . must meet the requirements. . . to the extent practical

within the limitations of design, geometry and materials of construction of

the components." As with relief requests for other Code components for

0-.

IN 96-32 June 5, 1996 incomplete or partial ASME Code-required ISI examinations, NRC authorization

is required when all the examination requirements of Paragraph (A) are not

met.

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 NRR project

manager.

Brian K. Grimes, Acting Director

Division of Reactor Program Management

Office of Nuclear Reactor Regulation

Technical contacts: Edmund J. Sullivan, NRR

(301) 415-3266 Internet:ejs@nrc.gov

Eric J. Benner, NRR

(301) 415-1171 Internet:ejbl@nrc.gov

Attachments:

1. Referenced Codes and Standards

2. List of Recentl Issued NRC Information Notices

14_"A=k renk 6/AA c4 4 -.4

K)- 1 Attachment 1 IN 96-32 June 5, 1996 Referenced Codes and Standards

1. Title 10 of the Code of Federal Regulations (10 CFR), Section

50.55a(g)(6)(ii)(A), "Augmented Examination of Reactor Vessel"

2. American Society of Mechanical Engineers, Boiler and Pressure Vessel

Code,Section XI, "Rules for Inservice Inspection of Nuclear Power Plant

Components," 1989 Edition.

I

Attachment 2 IN 96-32 June 5, 1996 LIST OF RECENTLY ISSUED

NRC INFORMATION NOTICES

Information Date of

Notice No. Subject Issuance Issued to

96-31 Cross-Tied Safety Injec- 05/22/96 All holders of OLs or CPs

tion Accumulators for pressurized water

reactors

96-30 Inaccuracy of Diagnostic 05/21/96 All holders of OLs or CPs

Equipment for Motor- for nuclear power reactors

Operated Butterfly Valves

96-29 Requirements in 10 CFR 05/20/96 All holders of OLs or CPs

Part 21 for Reporting and for nuclear power reactors

Evaluating Software Errors

96-28 Suggested Guidance Relating 05/01/96 All material and fuel cycle

to Development and Imple- licensees

mentation of Corrective

Action

96-27 Potential Clogging of High 05/01/96 All holders of OLs or CPs

Pressure Safety Injection for pressurized water

Throttle Valves During reactors

Recirculation

96-26 Recent Problems with Over- 04/30/96 All holders of OLs or CPs

head Cranes for nuclear power reactors

96-25 Transversing In-Core Probe 04/30/96 All holders of OLs or CPs

Overwithdrawn at LaSalle for nuclear power reactors

County Station, Unit 1

96-24 Preconditioning of Molded- 04/25/96 All holders of OLs or CPs

Case Circuit Breakers for nuclear power reactors

Before Surveillance Testing

96-23 Fires in Emergency Diesel 04/22/96 All holders of OLs or CPs

Generator Exciters During for nuclear power reactors

Operation Following Unde- tected Fuse Blowing

OL = Operating License

CP = Construction Permit

IN 96-xx

May xx, 1996 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 NRR project

manager.

Brian K. Grimes, Acting Director

Division of Reactor Program Management

Office of Nuclear Reactor Regulation

Technical contacts: Edmund J. Sullivan, NRR Eric J. Benner, NRR

(301) 415-3266 (301) 415-1171 internet:ejs~nrc.gov internet:ejbI~nrc.gov

Technical Editor reviewed and concurred on January 23, 1996.

JHConran of CRGR reviewed on January 11, 1996, and determined that subject

matter was appropriate for an information notice.

OGC has no legal objections (editorial changes incorporated) per conversation

with EJSullivan on 5/13/96.

• See previous concurrence

DOCUMENT NAME: G:\EJB1\50 55A.IN

To receive a copy of this document, dicate h the box: 'C' - copy without enclosures E - Copry with enclosures

IN'-No cnyCCO

OFFICE Contacts T D:DE I C:PECB ' I D:DRPM I

NAME EJSullivan* lBSheron* AEC I 'f BKGrimes

EJBenner* IA R R JY

DATE 1;3/2596125/9 .//9 511~9- / /96 OFFICIAL RECORD COPY V

IN 96-xx

May xx, 1996 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 NRR proj

manager. p

Brian K. Grimes, Acting Di

Division of Reactor Progras

Office of Nuclear Reactoj/ lation

Technical contacts: Edmund J. Sullivan, NRR Eric J. Benner, NRR

(301) 415-3266 (301) 415-1171 internet:ejs~nrc.gov internet:ejbl~nrc.gov

Technical Editor reviewed an concurred on January 23, 1996.

JHConran of CRGR reviewed n January 11, 1996, and determined that subject

matter was appropriate f an information notice.

OGC has no legal obje tons (editorial changes incorporated) per conversation

with EJSullivan on 13/96.

• See previous c ncurrence

DOCUMENT NAME G:\EJB1\S0_55A.IN

To reciv aopy fG document, kIdicate hI Ih box: 'C' - Copy without enclosures 'E' - Copy with enclosures

-N'

No copy /

r=,

OFFICE tontacts l D:DE C:PECB I D:DRPM

BKGrimes

NAME / EJSullivan*

/]EJBenner* I-

BSheron*

2/8/96 AEChaffee

5/ /96 , 5/ /96 DATE / 1/25/96 1/25/96 UI-tILIAL KLLUKU Lurl

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IN 96-xx

January xx, 1996 This information notice requires no specific action or written response.

you have any questions about the information in this notice, please cont

one of the technical contacts listed below or the appropriate NRR prompt!

manager. /r

Dennis M. Crutchfield, Dir etor

Division of Reactor Prog ok Management

Office of Nuclear Reac %r Regulation

Technical contacts: Edmund J. Sullivan. NRR Eric J. Benner , NRR

(301) 415-3266 (301),o415-1171 internet: ejs@nrc.gov internet: ejbl~nrc.gov

Technical Edi 'tewed and concurred on January 23, 1996.

JHConran of reviewed on January 11, 1996, and determined that subject

matter was E )riate for an information notice.

DOCUMENT NAME:

without enclosures 'E' with

w enclosures 'N' = No