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

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{{#Wiki_filter:4 UNITED STATES NUCLEAR REGULATORY
{{#Wiki_filter:4 UNITED STATES


COMMISSION
NUCLEAR REGULATORY COMMISSION


===OFFICE OF NUCLEAR REACTOR REGULATION===
OFFICE OF NUCLEAR REACTOR REGULATION
WASHINGTON, D.C. 20555 June 5, 1996 NRC INFORMATION


NOTICE 96-32: IMPLEMENTATION
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"
OF 10 CFR 50.55a(g)(6)(ii)(A),"AUGMENTED
 
EXAMINATION
 
OF REACTOR VESSEL"  


==Addressees==
==Addressees==
All holders of operating
All holders of operating licenses or construction permits for nuclear power
 
licenses or construction


permits for nuclear power reactors.
reactors.


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


Commission (NRC) is issuing this information
notice to alert addressees to certain aspects of scheduling and implementing


notice to alert addressees
the augmented reactor vessel examination required by Section


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


and implementing
It is expected that recipients will review the information for applicability


the augmented
to their facilities and consider actions, as appropriate, to avoid similar


reactor vessel examination
problems. However, suggestions contained in this information notice are not


required by Section 50.55a(g)(6)(ii)(A)
NRC requirements; therefore, no specific action or written response is
of Title 10 of the Code of Federal Regulations


(10 CFR).It is expected that recipients
required.


will review the information
Background


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


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


and consider actions, as appropriate, to avoid similar problems.
Reactor Vessel" [hereinafter referred to as Paragraph (A)], which contains new


However, suggestions
requirements for an augmented examination of reactor vessels. The rule


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


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


notice are not NRC requirements;
the American Society of Mechanical Engineers, Boiler and Pressure Vessel Code
therefore, no specific action or written response is required.Background


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


the scope of inspection
reactor vessel shell welds. Licensees with fewer than 40 months remaining in


of reactor vessels, the NRC issued, in 1992, 10 CFR 50.55a(g)(6)(ii)(A), "Augmented
the ISI interval that was in effect on September 8, 1992, may defer the


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


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


referred to as Paragraph (A)], which contains new requirements
Paragraph (A)(2) as "more than 90% of the examination volume of each weld"
[emphasis added].


for an augmented
Licensees unable to completely satisfy the requirements for the augmented


examination
reactor vessel examination must propose an alternative that would provide an


of reactor vessels. The rule requires licensees
acceptable level of quality and safety. The proposed alternative may be used


to implement, before the time required by normal updating of the inservice
when authorized by the Director of the Office of Nuclear Reactor Regulation


inspection (ISI) program, provisions
(NRR) [Paragraph (A)(5)].


in the 1989 Edition of the American Society of Mechanical
PDA LA6U2:41-E          CE-032 q7c5'5                                A


Engineers, Boiler and Pressure Vessel Code (ASME Code), Section XI, to examine "essentially
*  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


100%" of the length of all reactor vessel shell welds. Licensees
was developed to ensure the effectiveness of ultrasonic examinations through a


with fewer than 40 months remaining
performance demonstration to evaluate the adequacy of procedures, equipment, and personnel for detecting and sizing flaws during examinations. Licensees


in the ISI interval that was in effect on September
are not currently required to implement Appendix VIII.


8, 1992, may defer the augmented
==Description of Circumstances==
It became evident to the staff while it was conducting ISI reviews that some


reactor vessel examination
licensees were unaware of or uncertain about some aspects of the augmented


to the first period of the next ISI interval [Paragraph (A)(3)]. "Essentially
reactor vessel examination rule.


100%" examination
The staff learned that a small number of licensees were unaware of the rule


is defined in Paragraph (A)(2) as "more than 90% of the examination
and its requirements for some time after it was published. Licensees need to


volume of each weld"[emphasis
be aware of the schedular requirements of the rule to ensure timely


added].Licensees
implementation of its provisions. Because of the scope and extent of the


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


satisfy the requirements
of the reactor pressure vessel.


for the augmented reactor vessel examination
This information notice contains a discussion of certain areas of


must propose an alternative
misinterpretation that the staff has dealt with in the implementation of the


that would provide an acceptable
augmented reactor vessel examination rule.


level of quality and safety. The proposed alternative
Discussion


may be used when authorized
Schedular Requirements of the Rule


by the Director of the Office of Nuclear Reactor Regulation (NRR) [Paragraph (A)(5)].PDA 2:41-E LA6U CE-032 q7c5'5 A
In one instance, a licensee original 10-year ISI interval end date allowed


* AIN 96-32 June 5, 1996 The 1989 Edition of the ASME Code, Section XI, incorporated
deferral to the first period of the next interval. However, this licensee


Appendix VIII,"Performance
experienced an extended shutdown and, as permitted by Section XI, extended the


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


for Ultrasonic
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


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


Systems." Appendix VIII was developed
schedule the examination in accordance with the original 10-year ISI interval


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


of ultrasonic
examination equipment.


examinations
Essentially 100%0 Examination Standard


through a performance
Most licensees are finding that while the overall average examination coverage


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


to evaluate the adequacy of procedures, equipment, and personnel
individual welds may be substantially less than 90%. When a licensee is


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


and sizing flaws during examinations.
authorization of an alternative in accordance with Paragraph (A)(5).


Licensees are not currently
During discussions with the NRC staff regarding the review of the 10-year ISI


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


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


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


It became evident to the staff while it was conducting
the rule, the licensee did not submit a request for authorization of an


ISI reviews that some licensees
alternative to the NRC as required by the rule, until asked to do so by the


were unaware of or uncertain
NRC.


about some aspects of the augmented reactor vessel examination
uSplrit of Appendix VIII" Examination


rule.The staff learned that a small number of licensees
Section XI contains rules for evaluating the significance of flaws identified


were unaware of the rule and its requirements
through non-destructive examination. Flaws that are of such size that they


for some time after it was published.
cannot be dispositioned through comparison with code tables must be analyzed


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


need to be aware of the schedular
Flaws." Furthermore, Section XI, Paragraph IWB-3134(b), "Review by


requirements
Authorities," requires that analytical evaluations performed in accordance


of the rule to ensure timely implementation
with Paragraph IWB-3600 be submitted to the regulatory authority having


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


Because of the scope and extent of the examination, significant
One licensee administered a "Spirit of Appendix VIII" performance


planning is necessary
demonstration for the procedures, personnel, and equipment to be used for the


to address the technical, schedular, and regulatory
augmented reactor vessel examination. This type of examination essentially


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


with a comprehensive
yield more accurate and reliable inspection results. The licensee concluded


examination
that the performance demonstration resulted in examination and evaluation


of the reactor pressure vessel.This information
techniques that surpassed the conventional techniques of Section XI of the


notice contains a discussion
ASME Code and Regulatory Guide 1.150, "Ultrasonic Testing of Reactor Vessel


of certain areas of misinterpretation
Welds During Preservice and Inservice Examinations." During the augmented


that the staff has dealt with in the implementation
reactor vessel examination, the licensee identified 15 flaws in the shell


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


reactor vessel examination
reactor vessel examination, which required analytical evaluation in accordance


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


Schedular
conventional techniques of Section XI and Regulatory Guide 1.150 had been


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


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.
the remaining 3 flaws would have required analytical evaluation in accordance


However, this licensee experienced
with Paragraph IWB-3600. This licensee experience indicates that flaws of


an extended shutdown and, as permitted
sufficient size to require analytical evaluation may not be detected when


by Section XI, extended the ISI interval to complete the examinations
using conventional techniques for the augmented reactor vessel examination.


required for the interval.
Although the licensee in the above example submitted a request for


As a result, more than 40 months remained in the interval in effect on September
authorization of an alternative as the examination coverage was less than


8, 1992, and the licensee would have been required to do the examination
"essentially 100%," it did not submit the flaw evaluations, as required by the


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


The licensee requested
===Need for NRC Authorization of Alternatives===
A licensee unable to obtain the required examination coverage quoted 10 CFR


and was granted approval by NRR to schedule the examination
50.55a(g)(4) as a basis for not seeking NRC authorization of an alternative as


in accordance
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


with the original 10-year ISI interval end date to allow for proper scheduling
within the limitations of design, geometry and materials of construction of


and to ensure the availability
the components." As with relief requests for other Code components for


of examination
0-.


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


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


100%0 Examination
met.


Standard Most licensees
This information notice requires no specific action or written response. If


are finding that while the overall average examination
you have any questions about the information in this notice, please contact


coverage for reactor vessel shell welds may be more than 90%, examination
one of the technical contacts listed below or the appropriate NRR project


coverage for individual
manager.


welds may be substantially
Brian K. Grimes, Acting Director


less than 90%. When a licensee is unable to examine "essentially
Division of Reactor Program Management


100%" of each shell weld, it must seek NRC authorization
Office of Nuclear Reactor Regulation


of an alternative
Technical contacts:    Edmund J. Sullivan, NRR


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


with Paragraph (A)(5).During discussions
Eric J. Benner, NRR


with the NRC staff regarding
(301) 415-1171 Internet:ejbl@nrc.gov


the review of the 10-year ISI program plan, a licensee stated that it had obtained "essentially
Attachments:
 
1. Referenced Codes and Standards
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.
2. List of Recentl Issued NRC Information Notices


This type of examination
14_"A=k renk 6/AA                            c4 4          -.4


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


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


the technical
50.55a(g)(6)(ii)(A), "Augmented Examination of Reactor Vessel"
2. American Society of Mechanical Engineers, Boiler and Pressure Vessel


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


of Appendix VIII and would be expected to yield more accurate and reliable inspection
Components," 1989 Edition.


results. The licensee concluded that the performance
I


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


resulted in examination
NRC INFORMATION NOTICES


and evaluation
===Information                                  Date of===
Notice No.            Subject                Issuance  Issued to


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


that surpassed
tion Accumulators                        for pressurized water


the conventional
reactors


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


of Section XI of the ASME Code and Regulatory
Equipment for Motor-                      for nuclear power reactors


Guide 1.150, "Ultrasonic
Operated Butterfly Valves


Testing of Reactor Vessel Welds During Preservice
96-29          Requirements in 10 CFR        05/20/96  All holders of OLs or CPs


and Inservice
Part 21 for Reporting and                 for nuclear power reactors


Examinations." During the augmented reactor vessel examination, the licensee identified
Evaluating Software Errors


15 flaws in the shell welds and in the shell-to-flange
96-28          Suggested Guidance Relating    05/01/96  All material and fuel cycle


weld outside the scope of the augmented reactor vessel examination, which required analytical
to Development and Imple-                licensees


evaluation
mentation of Corrective


in accordance
Action


with Section XI, Paragraph
96-27          Potential Clogging of High    05/01/96  All holders of OLs or CPs


IWB-3600.
Pressure Safety Injection                for pressurized water


The licensee stated that if the conventional
Throttle Valves During                    reactors


techniques
Recirculation


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


Guide 1.150 had been used, 12 of these 15 flaws would not have even been recordable
head Cranes                              for nuclear power reactors


and only 2 of the remaining
96-25          Transversing In-Core Probe    04/30/96  All holders of OLs or CPs


3 flaws would have required analytical
Overwithdrawn at LaSalle                  for nuclear power reactors


evaluation
County Station, Unit 1
96-24          Preconditioning of Molded-    04/25/96  All holders of OLs or CPs


in accordance
Case Circuit Breakers                    for nuclear power reactors


with Paragraph
Before Surveillance Testing


IWB-3600.
96-23          Fires in Emergency Diesel      04/22/96  All holders of OLs or CPs


This licensee experience
Generator Exciters During                  for nuclear power reactors


indicates
Operation Following Unde- tected Fuse Blowing


that flaws of sufficient
OL = Operating License


size to require analytical
CP = Construction Permit


evaluation
IN 96-xx


may not be detected when using conventional
May xx, 1996 This information notice requires no specific action or written response. If


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


for the augmented
one of the technical contacts listed below or the appropriate NRR project


reactor vessel examination.
manager.


Although the licensee in the above example submitted
Brian K. Grimes, Acting Director


a request for authorization
Division of Reactor Program Management


of an alternative
Office of Nuclear Reactor Regulation


as the examination
Technical contacts:            Edmund J. Sullivan, NRR                        Eric J. Benner, NRR


coverage was less than"essentially
(301) 415-3266                                  (301) 415-1171 internet:ejs~nrc.gov                            internet:ejbI~nrc.gov


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
Technical Editor reviewed and concurred on January 23, 1996.


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


A licensee unable to obtain the required examination
matter was appropriate for an information notice.


coverage quoted 10 CFR 50.55a(g)(4)
OGC has no legal objections (editorial changes incorporated) per conversation
as a basis for not seeking NRC authorization


of an alternative
with EJSullivan on 5/13/96.


as required by Paragraph (A)(5). However, 10 CFR 50.55a(g)(4)
*See previous concurrence
states, in part, that "components.


..must meet the requirements.
DOCUMENT NAME: G:\EJB1\50 55A.IN


..to the extent practical within the limitations
To receive a copy of this document, dicate h the box: 'C' - copy without enclosures E - Copry with enclosures


of design, geometry and materials
IN'-No cnyCCO


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


of the components." As with relief requests for other Code components
NAME          EJSullivan*              lBSheron*                AEC I          'f      BKGrimes


for
EJBenner*                          IA R      R                JY


0-.IN 96-32 June 5, 1996 incomplete
DATE          1;3/2596125/9            .//9                    511~9-                     / /96 OFFICIAL RECORD COPY                V


or partial ASME Code-required
IN 96-xx


===ISI examinations, NRC authorization===
May xx, 1996 This information notice requires no specific action or written response. If
is required when all the examination


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


of Paragraph (A) are not met.This information
one of the technical contacts listed below or the appropriate NRR proj


notice requires no specific action or written response.
manager.                                                                                     p


If you have any questions
Brian K. Grimes, Acting Di


about the information
Division of Reactor Progras


in this notice, please contact one of the technical
Office of Nuclear Reactoj/                  lation


contacts listed below or the appropriate
Technical contacts:        Edmund J. Sullivan, NRR                          Eric J. Benner, NRR


NRR project manager.Brian K. Grimes, Acting Director Division of Reactor Program Management
(301) 415-3266                                    (301) 415-1171 internet:ejs~nrc.gov                              internet:ejbl~nrc.gov


===Office of Nuclear Reactor Regulation===
Technical Editor reviewed an concurred on January 23, 1996.
Technical


contacts: Edmund J. Sullivan, NRR (301) 415-3266 Internet:ejs@nrc.gov
JHConran of CRGR reviewed n January 11, 1996, and determined that subject


Eric J. Benner, NRR (301) 415-1171 Internet:ejbl@nrc.gov
matter was appropriate f an information notice.


Attachments:
OGC has no legal obje tons (editorial changes incorporated) per conversation
1. Referenced


Codes and Standards 2. List of Recentl Issued NRC Information
with EJSullivan on        13/96.


Notices 14_"A=k renk 6/AA c4 4-.4 K)- 1 Attachment
*See previous c ncurrence


1 IN 96-32 June 5, 1996 Referenced
DOCUMENT NAME G:\EJB1\S0_55A.IN


Codes and Standards 1. Title 10 of the Code of Federal Regulations
To reciv aopy fG  document, kIdicate hI Ih box: 'C' - Copy without enclosures 'E' - Copy with enclosures


(10 CFR), Section 50.55a(g)(6)(ii)(A), "Augmented
-N'
        No copy /
                                                                                r=,
  OFFICE tontacts          l        D:DE                    C:PECB              I      D:DRPM


Examination
BKGrimes


of Reactor Vessel" 2. American Society of Mechanical
NAME / EJSullivan*
            /]EJBenner*              I-
                                      BSheron*
                                      2/8/96 AEChaffee


Engineers, Boiler and Pressure Vessel Code, Section XI, "Rules for Inservice
5/ /96          ,         5/ /96 DATE / 1/25/96 1/25/96 UI-tILIAL KLLUKU      Lurl


Inspection
A44
  /
/
/


of Nuclear Power Plant Components," 1989 Edition.
Ih    .


I Attachment
IN 96-xx


2 IN 96-32 June 5, 1996 LIST OF RECENTLY ISSUED NRC INFORMATION
January xx, 1996 This information notice requires no specific action or written response.


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


Date of Notice No. Subject Issuance Issued to 96-31 96-30 Cross-Tied
one of the technical contacts listed below or the appropriate NRR prompt!
        manager.                                                                     /r


Safety Injec-tion Accumulators
Dennis M. Crutchfield, Dir etor


Inaccuracy
Division of Reactor Prog ok Management
 
of Diagnostic
 
Equipment
 
for Motor-Operated Butterfly
 
Valves Requirements
 
in 10 CFR Part 21 for Reporting
 
and Evaluating
 
Software Errors 05/22/96 05/21/96 05/20/96 All holders of OLs or CPs for pressurized
 
water reactors All holders of OLs or CPs for nuclear power reactors All holders of OLs or CPs for nuclear power reactors 96-29 96-28 Suggested
 
Guidance Relating to Development
 
and Imple-mentation
 
of Corrective
 
Action 05/01/96 All material licensees and fuel cycle 96-27 96-26 96-25 96-24 96-23 Potential
 
Clogging of High Pressure Safety Injection Throttle Valves During Recirculation
 
Recent Problems with Over-head Cranes Transversing
 
In-Core Probe Overwithdrawn
 
at LaSalle County Station, Unit 1 Preconditioning
 
of Molded-Case Circuit Breakers Before Surveillance
 
Testing Fires in Emergency
 
Diesel Generator
 
Exciters During Operation
 
Following
 
Unde-tected Fuse Blowing 05/01/96 04/30/96 04/30/96 04/25/96 04/22/96 All holders of OLs or CPs for pressurized
 
water reactors All holders of OLs or CPs for nuclear power reactors All holders of OLs or CPs for nuclear power reactors All holders of OLs or CPs for nuclear power reactors All holders of OLs or CPs for nuclear power reactors 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 (301) 415-3266 internet:ejs~nrc.gov
 
Eric J. Benner, NRR (301) 415-1171 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 CO cnyC 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 NAME / EJSullivan*
BSheron* AEChaffee
 
BKGrimes/] EJBenner*
I-DATE / 1/25/96 1/25/96 2/8/96 5/ /96 , 5/ /96///UI-tILIAL
 
KLLUKU Lurl A44 Ih .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
Office of Nuclear Reac %r Regulation


Sullivan.
Technical contacts:    Edmund J. Sullivan. NRR               Eric J. Benner , NRR
 
NRR Eric J. Benner Technical
 
contacts: Edmund J., NRR (301) 415-3266 internet:
ejs@nrc.gov
 
(301),o415-1171 internet:
ejbl~nrc.gov
 
Technical


Edi'tewed and concurred
(301) 415-3266                        (301),o415-1171 internet: ejs@nrc.gov                internet: ejbl~nrc.gov


on January 23, 1996.JHConran of matter was E reviewed on January 11, 1996, and determined
Technical Edi        'tewed and concurred on January 23, 1996.


that subject)riate for an information
JHConran of      reviewed on January 11, 1996, and determined that subject


notice.DOCUMENT NAME: without enclosures
matter was E      )riate for an information notice.


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{{Information notice-Nav}}
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Latest revision as of 04:38, 24 November 2019

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  Entergy icon.png
Issue date: 06/05/1996
From: Grimes B
Office of Nuclear Reactor Regulation
To:
References
IN-96-032, NUDOCS 9605200277
Download: ML031060052 (9)
v • d • e


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

A44

/

/

/

Ih .

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


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