ML20196J964

Safety Evaluation Accepting Licensee Third 10-yr Interval Inservice Insp Plan Request for Relief RI-27,rev 1

ML20196J964
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Site:	Cooper
Issue date:	12/07/1998
From:	NRC (Affiliation Not Assigned)
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NUDOCS 9812110135
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2 NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 3066H001 4

9 . . . . . ,o SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION OF THE THIRD 10-YEAR INTERVAL INSERVICE INSPECTION PLAN i REQUESTS FOR RELIEF NO. RI-27 EQB NEBRASKA PUBLIC POWER DISTRICT COOPER NUCLEAR STATION l

DOCKET NO. 50-298

1.0 INTRODUCTION

Inservice inspection of the American Society of Mechanical Engineers (ASME) Code Class 1,2, and 3 components shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel (B&PV) Code and applicable addenda as required by 10 CFR 50.55a(g),

except where specific relief has been granted by the Commission pursuant to 10 CFR 50.55a(6)(g)(i).10 CFR 50.55a(a)(3) states that attematives to the requirements of paragraph (g) may be used, when authorized by the Nuclear Regulatory Commission (NRC), if (i) the proposed attematives would provide an acceptable level of quality and safety or (ii) compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.

Pursuant to 10 CFR 50.55a(g)(4), ASME Code Class 1,2, and 3 components (including supports) shall meet the requirements, except the design and access provisions and the pre-service examination requirements, set forth in the ASME Code,Section XI, " Rules for Inservice inspection of Nuclear Power Plant Components," to the extent practical within the limitations of design, geometry, and materials of construction of the components. The regulations require that inservice examination of components and system pressure tests conducted during the first ten-year interval and subsequent intervals comply with the requirements in the latest edition and addenda of Section XI of the ASME Code incorporated by reference in 10 CFR 50.55a(b) i twelve months prior to the start of the 120-month interval, subject to the limitations and modifications listed therein. The applicable edition of Section XI of the ASME Code for the j

Cooper Nuclear Station (CNS) third ten-year inservice inspection (ISI) interval is the 1989 Edition.

9812110135 981207 PDR ADOCK 05000298 G PDR ENCLOSURE t

2 2.0 EVALUATION By letter dated October 261998, Nebsaska Public Power District (licensee) submitted its Third Ten-Year Interval Inservice Inspection Program Plan Request for Relief No. Rl-27 for CNS.

However, the licensee revised its relief and by its letter dated November 17,1998, submitted Request for Relief RI-27, Revision 1. The Idaho National Engineering and Environmental Laboratory (INEEL), has evaluated the information provided by the licensee in support of its Third Ten-Year Interval inservice inspection Program Request for Relief No.

RI-27, Revision 1, for CNS. Based on the results of the review, the staff adopts the contractor's conclusions and recommendations presented in the Technical Letter Report (TLR) attached. A review of the licensee's flaw evaluation of the subject flaw was completed by the staff and it concluded that the licensee's evaluation was acceptable in NRC Safety Evaluation dated November 23,1998.

The information provided by the licensee in support of the request for relief from Code requirements has been evaluated and the basis for disposition is documented below.

Request for Relief No. RI 27, Rev 1: ASME Code,Section XI, Paragraph IWB-2430(a) states that when examinations performed in accordance with Table IWB-2500-1 revealindications exceeding the acceptance standards of Table IWB-3410-1, additional examinations shall be performed during the same outage. The additional examinations shallinclude the remaining welds, areas, or parts included in the inspection item listing and scheduled for this and the subsequent period.

Pursuant to 10 CFR 50.55a(a)(3)(ii), the licensee proposed to perform additional examinations on two other main steam nozzle-to-vessel welds that are scheduled for examination during the second period. This attemative is in lieu of performing additional examinations on the remaining nine RPV nozzle-to-vessel welds during the current (1998 refueling outage). The licensee stated:

. . the District requests that the additional examinations of nozzle to vessel welds during RFO-18 be limited to the two (2) Main Steam nozzle to vessel welds. In addition, as noted above, the nine (9) nozzle to vessel weld examinations scheduled for the second period will be completed in RFO-19 along with the reexamination of Main Steam Nozzle to Vessel weld N3A. This includes the reexamination of the two (2) additional Main Steam nozzle to vessel welds examined during RFO-18. Since the hardship is significant with no compensating increase in safety, relief is requested in accordance with 10 CFR 50.55a(a)(3)(ii)."

Paragraph IWB-2430(a) requires that additional examinations be performed when initial examinations reveal indications that exceed the acceptance standards of Table IWB-3410-1.

The additional examinations shall be performed during the same outage and include the remaining welds, areas, or parts in the inspection item listing and scheduled for this ano the subsequent period. The licensee has identified an indication in main steam nozzle-to-vessel i Weld N3A that exceeds the acceptance standards, and proposed to perform the additional j examinations based on selection from welds, areas, or parts of similar material and service, as l

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allowed by the 1991 Addenda of Section XI. For the main steam nozzle-to-vessel Weld N3A, the only welds with similar material and service are the other main steam nozzles.

The current Code (1989 Edition) requires that the nine item B3.10 RPV nozzle-to-vessel welds scheduled for examination during the second period be examined during the current outage, .

and again in the second period. Later editions of the Code (1991 Addenda) were revised to l clarify that additional examinations need only be performed on welds, areas, or parts that would '

be subject to a possible common service degradation mechanism. However, the licensee has  !

determined that the subject mid-wall indication found in Weld N3A is a fabrication flaw and would not exhibit a common service degradation mechanism. Regardless, the licensee is proposing to examine two additional main steam nozzle-to-vessel welds during the current outage. These nozzles are also scheduled for future examination during the second period.

l Performing the additional examinations this outage on the nine item B3.10 welds that are

scheduled for examination in the second period, will require hydrolyzing the nozzles to reduce l dose, erecting and removing scaffolding (as needed), removing and reinstalling insulation and i

shield blocks, and cleaning the surface. These activities, in addition to the time required for examinations, would add several days to the outage and result in an estimated 12 person-rem of personnel radiation exposure. Imposition of these additional examinations would cause a hardship on the licensee.

The licensee has examined nine RPV nozzle-to-vessel welds during the first period. Further examinations were performed during the current outage on two additional main steam nozzle-to-vessel welds originally scheduled for examination in period two. These two main steam l nozzles and main steam nozzle-to-vessel Weld N3A will be reexamined during the second

period along with the remaining item B3.10 nozzle-to-vessel welds originally scheduled for examination during the second period.

The Code also requires that main steam nozzle-to-vessel Weld N3A be reexamined during the l next three (3) inspection periods before reverting to the original schedule of successive

! examinations [lWB-2420(c)) to validate the fracture mechanics analysis that was performed.

I The staff concluded that because the main steam nozzles are subjected to the same service conditions, the measures proposed by the licensee would detect existing patterns of degradation, and provide reasonable assurance of the structuralintegrity of these nozzles.

Requiring the licensee to examine, during the current outage, additional nozzles that do not j experience similar service conditions, and are already scheduled for examination during the ]

next period, would not provide a compensating increase in the level of quality and safety. In '

addition, the staff performed a review in NRC Safety Evaluation dated November 23,1998, of the licensee's flaw evaluation of the subject flaw and it concluded that the licensee's evaluation of the flaw was acceptable. Therefore, the staff concludes that the licensee's proposed attemative is authorized pursuant to 10 CFR 50.55a(a)(3)(ii) for the current interval.  ;

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3.0 CONCLUSION

The staff concluded that the licensee's proposed attemative to examine two additional main  ;

steam nozzle-to-vessel welds during the current outage, in addition to the nine RPV nozzle-to-l vessel welds that were examined during the first period, provides reasonable assurance of '

continued inservice structuralintegrity at CNS. In addition, the staff performed a review in NRC Safety Evaluation dated November 23,1998, of the licensee's flaw evaluation of the subject

flaw and it concluded that the licensee's evaluation of the flaw was acceptable. The staff concludes that the licensee's proposed alternative is authorized pursuant to 10 CFR 50.55a(a)(3)(ii).

Attachment:

As stated PrincipalContributor: T. McLellan Date: December 7, 1998

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TECHNICAL LETTER REPORT ON THIRD 10-YEAR INTERVAL INSERVICE INSPECTION REQUEST FOR RELIEF RI-27 EQB NEBRASKA PUBLIC POWER DISTRICT COOPER NUCLEAR STATION DOCKET NO. 50-298

1. INTRODUCTION By letter dated October 26,1998, the licensee, Nebraska Public Power District, submitted Request for Relief RI-27, seeking relief from the requirements of the ASME Code,Section XI, for the Cooper Nuclear Station (CNS) third 10-year inservice inspection (ISI) intental. The Idaho National Engineering and Environmental Laboratory (INEEL) staff's evaluation of the subject request for relief is in the following section. By letter dated November 17,1998, the licensee submitted Request for Relief RI-27, Revision 1, addressing comments made by the INEEL and NRC in a telephone conference with the licensee on November 9,1998. )

2. EVALUATION The information provided by Nebraska Public Power District in support of the request for relief from Code requirements has been evaluated and the basis for disposition is documented below. The Code of record for the CNS, third 10-year ISI interval, which began February 1996 is the 1989 Edition of Section XI of the ASME Boiler and Pressure Vessel Code.

Recuest for Relief No. RI-27. Revision 1. Paraoraoh IWB-2430fal Additional Examination i Reauirements for Flaws in the RPV Main steam Nozzle-to-Vessel Weld l

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i 2 Code Reauirement: Paragraph IWB-2430(a) states that when examinations performed in l accordance with Table IWB-2500-1 revealindications exceeding the acceptance standards of Table IWB-3410-1, additional examinations shall be performed during the  ;

same outage. The additional examinations shall include the remaining welds, areas, or parts included in the inspection item listing and scheduled for this and the subsequent period.

f Licensee's Prooosed Attemative: In accordance with 10 CFR 50.55a(a)(3)(ii), the licensee proposed to perform additional examinations on two other main steam nozzle-to-vessel welds that are scheduled for examination during the second period. This alternative is in lieu of performing additional examinations on the remaining nine RPV nozzle-to-vessel welds during the current (1998 refueling outage). The licensee stated:

. . . the District requests that the additional examinations of nozzle to vessel welds during RFO-18 be limited to the two (2) Main Steam nozzle to vessel welds. In addition, as noted above, the nine (9) nozzle to vessel weld examinations scheduled for the second period will be completed in RFO-19 along with the reexamination of Main Stecm Nozzle to Vessel weld N3A. This includes the reexamination of the two (2) additional Main Steam nozzle to vessel welds  ;

examined during RFO-18. Since the hardship is significant with no compensating l increase in safety, relief is requested in accordance with 10 CFR 50.55a(a)(3)(ii)."  !

Licensee's Basis for Procosed Attemative (as stated):

• During the Fall 1998 Refueling outage, the last outage of the first period, an indication was identified in Main Steam nozzle to vessel weld N3A that exceeds the acceptance criteria of IWB-3512. This condition did not appear on the radiographs but it does have a sufficient interface to reflect ultrasonic signals. The indication was previously identified in 1976 and 1986. However, it was incorrectly dispositioned as being located in the nozzle forging. The indication was identified again in 1998 (RFO-18) and dispositioned to be located in the weld. The indication has been evaluated as original construction and is acceptable for continued operation in accordance with IWB-3600. The evaluation results (fracture mechanics evaluation) were previously submitted to the NRC in Reference 2. This indication is similar to weld indications previously identified in three (3) of the Feedwater nozzle to vessel welds (Reference Relief Request RI-19). Specific relief is requested on the basis that the required examinations present a hardship and that the proposed attemative would provide an acceptable level of quality and safety.

• Basis for Hardship I

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3 "The nine (9) nozzle to vessel welds examinations, scheduled for the first period of the Third Ten-year interval, have been performed during the Fall 1998 Refueling outage with the discovery of one reportable indication. The Code requires the current sample size to be expanded to include examinations required for the subsequent period, which in nine (9) additional nozzles. In order to plan, schedule, and perform the ultrasonic examinations of the additional welds, the nozzles are first hydrolyzed to reduce the dose, and then scaffolding is erected (as needed),

insulation and shield blocks are removed, and the surface is cleaned. These activities, plus the time for the examinations, followed by the reinstallation of shield blocks and insulation, and scaffold removal will add several days to the outage.

The dose for performing these additional weld examinations during RFO-18 is estimated to be 12 person-rem after hydrolyzing the nozzles. The district plans to reduce future outage radiation dose by implementing a modification to inject zine before RFO-19. Based on zine injection,less than 12 person-rem is estimated to be received when the examinations are performed again during the second period.

Furthermore, the performance of these additional examinations during RFO-18 would generate additional radioactive waste for disposal due to the housekeeping required after the completion of examinations. The increased exposure coupled with the increased industrial safety risk with erecting and working off scaffolding outweighs the benefits of examining the nine (9) additional nozzles. The above activities represent an unnecessary hardship without a compensating increase in safety as discussed below.

• Quality and Safety Basis for Relief in regards to Quality and Safety, the following four (4) points provide the basis for relief from performing the second period inspections during RFO-18:

"1) The indication is a welding discontinuity that has been present since original construction based on the following:

"a. The indication is subsurface near the mid-depth of the nozzle and is considered to be an original fabrication discontinuity and not service induced. The only potential service degradation mechanisms that could affect the steam nozzles are fatigue, stress corrosion cracking and erosion corrosion. For all these mechanisms, cracking indications would be expected to occur at the inside surface.

• b. The indication is the same indication as was identified in 1976 and 1986. The two (2) previous examinations identified the indication as being in the nozzle forging instead of the weld as determined in 1998. (See Dimensional Comparison discussion) Since the location of the indication was incorrectly identified by previous examinations, the discrepancy or " condition" was entered into the District's corrective action program. A review of past examination records of the other nozzle to vessel welds was conducted to l

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4 determine the " extent of conditions" with no similar conditions being identifed.

"c. The indication has not changed in size based on the recording data and evaluation of that data.- (See Sensitivity Comparison discussion)

• 2) Reportable indications have not been observed during previous examinations of the additional nine (9) nozzle to vessel welds.

• 3) The District expanded the sample by completing examinations of two (2) additional Main Steam nozzle to vessel welds pursuant to later editions of the Code. No reportable indications were identifed. The basis for limiting the expansion to two (2) additional Main Steam to vessel welds is discussed below:

"In the 1991 Addenda, IWB-2430(a) was revised to clarify that the additional examinations shall be selected from welds, areas, or parts of similar material and service. This clarification was provided to ensure that the additional examinations focused on welds, areas, or parts that would be subjected to a possible common service degradation mechanism. A common service degradation mechanism does not exist for original welding related discontinuities. The Main Steam nozzles are unique when considering all aspects of service conditions applicable to reactor vessel nozzles. The most notable factor is that the main steam nozzles are in a steam environment during normal plant operation. The thermal cycles that the vessel can experience are not greatly different when comparing the upper vessel where the steam nozzles are located and the mid-vessel where most other nozzles are located. However the steam environment has significantly different heat transfer properties which make the thermal response of the nozzles in the steam region different from those below the water line.

"In addition to the main steam nozzles,'the top head nozzles are also in i the steam region. However, the top head nozzles are small diameter l nozzles welded to the hemispherical top head, where it is just three  ;

inches thick. The steam nozzles are 24 inch diameter nozzles welded to the six inch thick cylindrical shell. Therefore, the stress conditions ,

associated with the nozzle-specific geometries are significantly different. Therefore, it is appropriate to limit the expansion sample to I two (2) Main Steam to vessel welds.

"4) During RFO-19, at the beginning of the second period, all nine (9) l nozzles will be examined in accordance with the requirements for  !

ASME Section XI. In addition, the nozzle N3A will be reexamined.

" Description of Indication - Examination History i

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"Preservice ultrasonic examinations were performed in accordance with ASME I. Section XI,1971 Edition. No actual data reports were available for review; I

however the summary report for the preservice examinations shows main steam nozzle N3A as having no recordable indications.

• During the first refueling outage in 1976, manual ultrasonic examinations,were performed on nozzle N3A in accordance with the 1974 Edition of ASME Section XI.

During the examination with the 60' shear wave transducer one (1) indication with the following parameters was recorded: Distance to weld centerline = 3.25"; Metal path to reflector = 5.4"; Length from W, to maximum amplitude = 14.5"; and Maximum amplitude = 100% distance amplitude correction (DAC). The evaluation of this indication determined that it was located in the nozzle forging and not in the weld.

" Main Steam nozzle (N3A) was examined again in 1986 with similar results. During the examination with the 60' shear wave transducer one (1) indication with the following parameters was recorded: 'W distance = 1.75"; Metal path to reflector =

5.3"; and Maximum amplitude = 90% DAC. The evaluation of this indication  ;

determined that it was located in the nozzle forging and not the weld.

"In 1998 (RFO-18) the examination with the 60' shear wave transducer identified one (1) recordable indication with the following parameters: W distance =1.20";

Metal path to reflector = 5.5"; Length to maximum amplitude = 13"; and Maximum amplitude = 80% DAC. This indication is located at mid-wall along the nozzle side fusion line between 10.5" and 22.75". No flaw indication could be found in the nozzle forging as was found in 1976 and 1986. The 1976 and 1986 indication mislocation were entered into the District's corrective action program. The following comparisons clarifies the size and locational differences.

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" Comparison - Dimensional Comparison "The construction drawings show the weld prep to be a 1.5" gap between the nozzle and the vessel wall with a backing ring attached to the outside. After welding, the backing ring was removed and a cosmetic weld cap was installed.

This cap was machined to a 5.5" radius to provide a smooth transition between the nozzle outside diameter and the vessel wall. This resulted in a weld cap which is approximately 3.5" wide on top of a weld which is 1.5" wide. Due to this 2.0" difference between the weld cap and weld width, the plotted location of indications could also vary by 2.0" dependent on where the initial reference point W is located. l "The metal path to reflector is very consistent in all three examinations which would indicate the same reflector. The 14.5" length location in 1976 is consistent with the length of the 1998 location. The differences in 'W locations (distance from the initial reference point 'W,') may be explained by differences in the reference point.

If the 1976 data used the center of the weld crown cap as 'W,' then the 3.25"

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l recorded is essentially the same measurement as the 1.2" location recorded in 1998 which used the nozzle blend radius weld toe as W,' reference.

" Sensitivity Comparison "The 1976 and the 1986 exeminations were performed to editions of the ASME code which required only a 50% DAC recording level with 0.9" increment measurements. Using this criterion, this indication was recorded as a series of spot indications which were observed all around the weld at this metal path. The l 1998 data was recorded at 20% DAC levels which resulted in the 12.25" length.

l Other indications were observed below the 20% recording levels at the same metal l path around the weld.

" Acceptance Comparison "The flaw signal characteristics are typical of those of a slag type flaw with multiple l reflectors and the ability to maintain signals over a wide range of skew angles. The flaw signal characteristics are not similar to those exoected from lack of fusion or fatigue cracking. These indications have been recorded since 1976 considering the l observed signal characteristics.

"Because of the change in recording criteria, no direct comparisons can be made to the length of these indications. If the indications have been characterized as planar flaws in the weld metal in 1976 and 1986, additional information would have been recorded as required by procedure, and the flaw indications would have been j evaluated for acceptance. Due to the evaluation of the indications as being non- l relevant reflectors in the nozzle forging in 1976 or as being laminar reflectors in I 1986, sufficient planar sizing data is not available to determine the acceptability of the indications to the standards of the Code Editions in effect at those times.

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" Conclusions  !

"1) The subject indication is a weld discontinuity indicative of slag inclusion that  :

has been present since original construction. l l

l "2) The subject indication is the same indication found in the 1976 and 1986 i examinations.

"3) The change in dimensional characteristics of the subject indication is related to the change in sensitivity of the new technology and inspection techniques."

Evaluation Paragraph IWB-2430(a) requires that additional examinations be performed when initial examinations reveal indications that exceed the acceptance standards of l

j Table IWB-3410-1. The additional examinations shall be performed during the same l outage and include the remaining welds, areas, or parts in the inspection item listing and l

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scheduled for this and the subsequent period. The licensee has identified an indication in j main steam nozzle-to-vessel Weld N3A that exceeds the acceptance standards, and l proposed to perform the ad,ditional examinations based on selection from welds, areas, or i -.:

pais of similar material and service. as allowed by the 1991 Addenda of Section XI. For the main steam nozzle-to-vessel Weld N3A, the only welds with similar material and service are the other main steam nozzles.

The current Code (1989 Edition) requires that the nine item B3.10 RPV nozzle-to-vessel welds scheduled for examination during the second period be examined during the current outage, and again in the second period. Later editions of the Code (1991 Addenda) were revised to clarify that additional examinations need only be performed on welds, areas, or parts that would be subject to a possible common service degradation mechanism. However, the licensee has determined that the subject mid-wall indication found in Weld N3A is a fabrication flaw and would not exhibit a common service degradation mechanism. Regardless, the licensee is proposing to examine two additional main steam nozzle-to-vessel welds during the current outage. These nozzles are also schedu!ed for future pxamination during the second period.

Performing the additional examinations this outage on the nine item B3.10 that are scheduled for examination in the second period, will require hydrolyzing the nozzles to reduce dose, erecting and removing scaffolding (as needed), removing and reinstalling insulation and shield blocks, and cleaning the surface. These activities, in addition to the time required for examinations, would add several days to the outage and result in an estimated 12 person-rem of personnel radiation exposure. Imposition of these additional examinations would cause a considerable hardship on the licensee.

The licensee has examined nine RPV nozzle-to-vessel welds during the first period. ,

Further examinations were performed during the current outage on two additional main steam nozzle-to-vessel welds originally scheduled for examination in period two. These o

two main steam nozzles and main steam nozzle-to-vessel Weld N3A will be reexamined  !

during the second period along with the remaining item B3.10 nozzle-to-vessel welds

originally scheduled for examination during the second period. The Code also requires l

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that main steam nonle-to-vessel Weld N3A be reexamined during the next three (3)

inspection periods before reverting to the original schedule of successive examinations

[lWB-2420(c)] to validate the fracture mechanics analysis that was performed. Because the main steam nonles are subjected to the same service conditions, the measures proposed by the licensee should detect existing patterns of degradation, and provide reasonable assurance of the structuralintegrity of these nonles. Requiring the licensee to examine, during the current outage, additional nozzles that do not experience similar service conditions, and are already scheduled for examination during the next period, would not provide a compensating increase in the level of quality and safety. Therefore, l l

it is recommended that the proposed attemative be authorized pursuant to 10 CFR l

50.55a(a)(3)(ii).

3.0 CONCLUSION

The INEEL staff evaluated the licensee's submittal and concluded that the licensee's proposed alternative to examine two additional main steam nozzle-to-vessel welds during l the current outage, in addition to the nine RPV nozzle-to-vessel welds that were examined during the first period, provides reasonable assurance of continued inservice l structural integrity at CNS. Therefore, it is recommended that the proposed attemative be authorized pursuant to 10 CFR 50.55a(a)(3)(ii).

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