ML12353A241
| ML12353A241 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek  |
| Issue date: | 01/04/2013 |
| From: | Markley M Plant Licensing Branch IV |
| To: | Matthew Sunseri Wolf Creek |
| Lyon C | |
| References | |
| TAC ME9078 | |
| Download: ML12353A241 (14) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 January 4, 2013 Mr. Matthew W. Sunseri President and Chief Executive Officer Wolf Creek Nuclear Operating Corporation Post Office Box 411 Burlington, KS 66839
SUBJECT:
WOLF CREEK GENERATING STATION - REQUEST FOR RELIEF NO. 13R-07 FOR THE THIRD 10-YEAR INSERVICE INSPECTION PROGRAM INTERVAL (TAC NO. ME9078)
Dear Mr. Sunseri:
By letter dated July 2,2012, as supplemented by letter dated October 15, 2012, the Wolf Creek Nuclear Operating Corporation (the licensee) requested relief from certain requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),
Code Case N-729-1 (N-729-1), "Alternative Examination Requirements for PWR [Pressurized Water Reactor] Reactor Vessel Upper Heads with Nozzles Having Pressure-Retaining Partial Penetration WeldsSection XI, Division 1," as conditioned in Title 10 of the Code of Federal Regulations (10 CFR), Section 50.55a, for examination of reactor vessel head penetration nozzle welds at the Wolf Creek Generating Station (WCGS). The licensee's request for relief No. 13R-07 contains a proposed alternative to the examination coverage of two reactor vessel head nozzle penetrations of the control rod drive mechanism (CRDM). The licensee requested to use the proposed alternative for the remaining duration of the third 10-year inservice inspection (lSI) interval, which ends on September 2, 2015.
Based on the information provided by the licensee in request for relief No. 13R-07, the U.S. Nuclear Regulatory Commission (NRC) staff has determined that the proposed alternative provides reasonable assurance of structural integrity of CRDM penetration nozzles numbers 77 and 78, and that complying with the ASME Code requirement would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.
Accordingly, the NRC staff concludes that the licensee has adequately addressed all of the regulatory requirements set forth in 10 CFR 50.55a(a)(3){ii) and 10 CFR 50.55a{g)(6)(ii)(D), and authorizes the use of relief request No. 13R-07 at WCGS for the remainder of the third 10-year lSI interval, which ends on September 2, 2015.
All other ASME Code,Section XI, requirements for which relief was not specifically requested and approved in this relief request remain applicable, including third-party review by the Authorized Nuclear Inservice Inspector.
M. Sunseri -2 The detailed results of the NRC staff review are provided in the enclosed safety evaluation. If you have any questions concerning this matter, please call Mr. F. Lyon of my staff at (301) 415-2296 or by electronic mail at fred.lyon@nrc.gov.
Sincerely, Michael T. Markley, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-482 Enclosure As stated cc w/encl: Distribution via Listserv
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION THIRD 10-YEAR INSERVICE INSPECTION PROGRAM INTERVAL REQUEST FOR RELIEF NO. 13R-07 WOLF CREEK NUCLEAR OPERATING CORPORATION WOLF CREEK GENERATING STATION DOCKET NO. 50-482
1.0 INTRODUCTION
By letter dated July 2, 2012 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML12193A559), as supplemented by letter dated October 15, 2012 (ADAMS Accession No. ML12341A228), the Wolf Creek Nuclear Operating Corporation (the licensee) requested relief from certain requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code), Code Case N-729-1 (N-729-1),
"Alternative Examination Requirements for PWR [Pressurized-Water Reactor] Reactor Vessel Upper Heads with Nozzles Having Pressure-Retaining Partial-Penetration WeldsSection XI, Division 1," as conditioned in Title 10 of the Code of Federal Regulations (10 CFR),
Section 50.55a, for examination of reactor vessel head penetration nozzle welds at the Wolf Creek Generating Station (WCGS). The-licensee's request for relief 13R-07 contains a proposed alternative to the examination coverage of two reactor vessel head nozzle penetrations of the control rod drive mechanism (CRDM). The licensee requested to use the proposed alternative for the remaining duration of the third 10-year inservice inspection (lSI) interval, which ends on September 2, 2015.
2.0 REGULATORY EVALUATION
Pursuant to 10 CFR 50.55a(g)(6)(ii), the U.S. Nuclear Regulatory Commission (NRC, the Commission) may require the licensee to follow an augmented lSI program for systems and components for which the Commission deems that added assurance of structural reliability is necessary. The regulations in 10 CFR 50.55a(g)(6)(ii)(D) require augmented lSI of reactor pressure vessel head penetration nozzles of pressurized-water reactors (PWRs) in accordance with N-729-1, subject to the conditions specified in paragraphs (2) through (6) of 10 CFR 50.55a(g)(6)(ii)(D).
The regulations in 10 CFR 50.55a(a)(3), state, in part, that alternatives to the requirements of 10 CFR 50.55a(g) may be used when authorized by the NRC if the licensee demonstrates:
Enclosure
-2 (i) the proposed alternatives 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.
Based on the above regulations, the NRC staff concludes that regulatory authority exists to authorize an alternative to N-729-1 as conditioned in 10 CFR 50.55a(g)(6)(ii)(D), as requested by the licensee.
3.0 TECHNCIAL EVALUATION
3.1 ASME Code Components Affected
Code Class: 1
Reference:
ASME Code Case N-729-1 as conditioned in 10 CFR 50.55a(g)(6)(ii)(D)
Item No.: B4.20
==
Description:==
UNS N06600 Nozzles and UNS N06082 or UNS W86182 Partial-Penetration Welds in Head.
Reactor vessel head CRDM penetration nozzle base material and J-groove weld that attaches the nozzle base material to the underside of the reactor vessel head for penetration nozzle numbers 77 and 78.
3.2 Applicable Code Edition and Addenda ASME Code Section XI, 1998 Edition through 2000 Addenda, as augmented by N-729-1, as conditioned in 10 CFR 50.55a(g)(6)(ii)(D).
3.3 Applicable Code Requirement (as stated by the licensee) 10 CFR 50.55a(g)(6)(ii)(D)(1) requires that examinations of the reactor vessel head penetration nozzles be performed in accordance with [N-729-1] subject to the conditions specified in paragraphs 10 CFR 50.55a(g)(6)(ii)(D)(2) through (6).
Paragraph -2500 of [N-729-1] states, in part:
If obstructions or limitations prevent examination of the volume or surface required by [Figure 2 in N-729-1] for one or more nozzles, the analysis procedure of Appendix I shall be used to demonstrate the adequacy of the examination volume or surface for each such nozzle. If Appendix I [of N-729-1] is used, the evaluation shall be submitted to the regulatory authority having jurisdiction at the plant site.
-3 Figure 2 in [N-729-1] requires that the volumetric or surface examination coverage distance below the toe of the J-groove weld (i.e. dimension "a") be 1.5 inches for incidence angle, 6, less than or equal to 30 degrees; 1 inch for incidence angle, 6, greater than 30 degrees; or to the end of the tube, whichever is less. These coverage requirements are applicable to [WCGS] reactor vessel head penetration nozzles as shown in Table 1.
Ta ble 1. WCGS Reactor VesseIHeadPenetraf Ion C overage Requlrement s Penetration Numbers Incidence Angle, e Required Coverage, "a" (degrees) (inches) 1 to 29 s 30 1.5 30 to 78 > 30 1.0 3.4 Reason for Request In its letter dated July 2, 2012, the licensee stated, in part, that Due to the physical configuration of certain reactor vessel head penetration nozzles, full examination volume coverage required by [N-729-1] Table 1 cannot be achieved for reactor vessel head penetration nozzle numbers 77 and 78, therefore, use of Mandatory Appendix I is requested In accordance with 10 CFR 50.55a(g)(6)(ii)(D){6).
Reactor vessel head CRDM penetration nozzles at WCGS have two styles of ends, referred to as Type "X" and Type "Y" (Figure 1). Penetrations 1 through 73 are Type "Y" that are essentially a smooth wall cylinder with rounded edges at the outer diameter and inner diameter at the bottom of the penetration.
Penetrations 74 through 78 have a threaded outside diameter and an internal taper.
The design of reactor vessel head penetration nozzles 74 through 78, referred to as Type "X," (Figure 1) includes a threaded section, approximately 1.19 inch in length at the bottom of the nozzles. These penetrations are located at the 48.7 degree location. The dimensional configuration at this location is such that the distance from the lowest point at the toe of the J-groove weld to the top of the threaded region could be less than the required coverage dimension "a" shown in Figure 2 of [N-729-1]. Therefore, deviation from the required inspection coverage is sought for reactor vessel head penetrations 77 and 78, as the required coverage for these two penetrations cannot be obtained.
For the initial examinations of reactor vessel head penetration welds performed in accordance with [NRC letter EA-03-009, "Issuance of First Revised NRC Order (EA-03-009) Establishing Interim Inspection ReqUirements for Reactor Pressure Vessel Heads at Pressurized Water Reactors," dated February 20, 2004 (ADAMS Accession No. ML040220181 )], a similar relief request was previously submitted for inability to examine the required examination volume
[licensee's dated October 5, 2006, as supplemented on November 1, 2006
-4 (ADAMS Accession Nos. ML062890419 and ML063110631, respectively)]. This previous request was approved by the NRC on December 7,2006 in [ADAMS Accession No. ML063330294].
3.5 Proposed Alternative and Basis for Use In its letter dated July 2, 2012, the licensee stated, in part, that As an alternative to the volumetric and surface examination coverage requirements shown as dimension "a" in Figure 2 of [N-729-1], WCGS proposes the use of attainable ultrasonic examination distances shown in Table 2. The required examination coven:~ge dimension for the other penetrations will be met or exceeded.
Table 2: WCGS Inspection Coverage Obtained for CRDM Penetrations Having Limited Coverage Penetration No. ! E> (degrees) I N*729*1 Required Inspection Coverage
. Exam Coverage Obtained (inches)
. . (inches) [Proposed coverage]
77 48.7 1.0 0.6 78 48.7 1.0 0.88 Appendix I of [N-729-1] provides the analysis procedure for evaluation of an alternative examination area or volume to that specified in Figure 2 of [N-729-1] if impediments prevent examination of the required zone. Section 1-1000 of
[N-729-1] requires, for alternative examination zones below the J-groove weld, that analyses shall be performed using at least the stress analysis method (Section 1-2000 [of N-729-1]) or the deterministic fracture mechanics analysis method (Section 1-3000 [of N-729-1]) to demonstrate that the applicable criteria are satisfied. The techniques described in Section 1-2000 were validated in
[WCAP-16589-P, Revision 0, "Structural Integrity Evaluation of Reactor Vessel Upper Head Penetrations to Support Continued Operation: Wolf Creek," August 2006 (ADAMS Accession No. ML062890422, proprietary)].
The licensee noted that although not required, it also validated WCAP-16589-P, Revision 0, in accordance with the deterministic fracture mechanics analysis described in Section 1-3000. The licensee stated that the fracture mechanics analysis does not fully meet the requirements of N-729-1, Section 1-3200(a), Method 1 in that WCAP-16589-P, Revision 0, used the crack growth formula in the Electric Power Research Institute (EPRI) report, "Materials Reliability Program (MRP) Crack Growth Rates for Evaluating Primary Stress Corrosion Cracking (PWSCC) of Thick Wall Alloy 600 Material (MRP-55), Revision 1."
-5 3.5.1 Stress Analysis in Accordance with N-729-1 Section 1-2000 In its letter dated July 2, 2012, the licensee stated, in part, that Section 1-2000 of N-729-1 requires that the plant-specific analysis demonstrate that the hoop and axial stresses (in the penetration nozzle below the J-groove weld) remain below 20 kips per square inch (ksi) (tensile) over the entire region outside the alternative examination zone but within the examination zone defined in Figure 2 of [N-729-1].
The distance below the J-groove weld that requires examination, as determined by the pOint at which the CRDM penetration hoop stress distribution for the operating tensile stress levels is less than 20 (ksi) tension, was obtained from Appendix A of [WCAP-16589-P, Revision 0] ....
The hoop stress distribution plots for penetration numbers 77 and 78 are provided in Figure 2 of the relief request. The hoop stress distribution plots in Figure 2 of the relief request indicate that the stresses are below 20 ksi outside of the licensee's proposed alternate examination zone below the bottom of the J-groove weld. The hoop stress distribution plots display the downhill side as this is more limiting. Also, stress distribution plots shown are for the inside and outside surface. Table 3 below summarizes the distance from below the toe of the downhill side J-groove weld to where both the inside and outside surface hoop stress drops below 20 ksi for penetration numbers 77 and 78.
Table 3: Distance Below Toe of Downhill Side J-Groove Weld Where Hoop Stress is Less Than 20 KSI Penetration Nozzle Source Distance Below Toe of No. Downhill Side J-Groove Weld Where Hoop Stress
< 20 ksi (inch) 77 and 78 Figure 2 in the relief 0.30 request 3.5.2 Deterministic Fracture Mechanics Analysis in Accordance with N-729-1 Section 1-3200, Method 1 The licensee stated that the fracture mechanics analysis in WCAP-16589-P, Revision 0, demonstrates that a potential axial crack in the unexamined zone will not grow to the toe of the J-groove weld prior to the next examination performed at the frequency speCified in Table 1 of N-729-1. The results of the analysis are shown as flaw tolerance charts, which can be used to determine minimum required inspection coverage. This ensures that any flaws initiated below the J-groove weld, in the region of the penetration nozzle not being inspected, would not reach the bottom of the J-groove weld before the next inspection. The flaw tolerance chart for penetration numbers 77 and 78 is shown in Figure 3 of the relief request.
The licensee explained that the flaw tolerance chart demonstrates that a postulated through-wall flaw at the bottom edge of the proposed alternative examination zone will not grow to the toe of
-6 the J-groove weld within an inspection interval of four refueling cycles. The crack growth prediction shows greater than six effective full power years (EFPY) of operation required to grow the postulated flaw to the toe of the J-groove weld.
3.5.3 Surface Examination The regulations in 10 CFR 50.55a(g)(6)(ii)(D)(3) state, in part, that "... if a surface examination is being substituted for a volumetric examination on a portion of a penetration nozzle that is below the toe of the J-groove weld, the surface examination shall be of the inside and outside wetted surface of the penetration nozzle not examined volumetrically .... The licensee stated that to 11 reduce personnel radiation exposure, the nozzles are typically inspected using remotely operated volumetric examination equipment The licensee explained that although dye penetrant testing of threaded surfaces is possible, it is not practical. The threaded outside diameter makes a dye penetrant examination on the lower section of the penetration impractical because of excessive bleed out from the threads. The licensee noted that eddy current examination would similarly not be effective due to the threaded configuration. Additionally, the radiation levels under the reactor vessel head are estimated to be 10,000 millirem (mRem)/hour at the bottom of the CRDM nozzles resulting in an exposure of approximately 2500 mRem per nozzle.
According to the licensee, these dose rates are consistent with dose rates measured at the Seabrook Station during 2006, as reported in a Seabrook Station request for relief dated October 27,2011 (ADAMS Accession No. ML11307A370). The reactor vessel head configuration at Seabrook Station is similar to that at WCGS. At Seabrook Station, radiation levels under the reactor vessel head were measured during its previous inspection in 2006 and ranged from 7000 mRem/hour to 10,000 mRem/hour at the bottom of the CRDM nozzles, resulting in an exposure of approximately 1750 to 2500 mRem per nozzle to perform surface examination. Therefore, the licensee did not propose alternative surface examinations for nozzle penetration numbers 77 and 78.
3.6 Duration of Proposed Alternative The alternative requirements of this request will be applied for the remaining duration of the current third 10-year lSI interval, which began on September 3,2005, and ends on September 2, 2015, as discussed in the licensee's letter dated October 15, 2012, submittal.
3.7 NRC Staff Evaluation Pressurized-water reactors have experienced primary water stress-corrosion cracking (PWSCC) in CRDM penetration nozzles as the result of the combination of susceptible material, such as nickel-based Alloy 600 or Alloy 82/182 weld metal, corrosive enVironment, and tensile stresses. The examination and analysis requirements of N-729-1, as conditioned by 10 CFR 50.55a(g)(6)(ii)(D), are intended to monitor and ensure the leak tightness and structural integrity of CRDM penetration nozzles and their associated J-groove welds.
-7 3.7.1 Flaw Evaluation As discussed above, in 2006, the licensee submitted a similar relief request for CRDM nozzle numbers 77 and 78 based on the analyses presented in WCAP-16589-P, Revision O. At the time, the NRC staff concluded that WCAP-16589-P, Revision 0, was acceptable and approved the previous relief request by letter dated December 7, 2006.
To support the alternative examination coverage for nozzle numbers 77 and 78 in the current relief request, the licensee used the stress analysis and deterministic fracture mechanics analysis in WCAP-16589-P, Revision 0, to demonstrate that the proposed ultrasonic testing (UT) examination distance provides a reasonable assurance of structural integrity and leak tightness until the next scheduled examination. The result of the licensee's stress analysis as shown in Figure 2 of the current relief request demonstrated that the stresses in the minimum achievable inspection coverage distance below the bottom of the J-groove weld are in excess of 20 ksi (tenSile) and in the uninspected region are less than 20 ksi. Experimental tests have shown that PWSCC would most likely initiate if the tensile stresses reach 20 ksi. The structural integrity of the nozzle region that has stresses exceeding 20 ksi will be monitored by periodic examination. The structural integrity of the nozzle region with less than 20 ksi stresses is ensured because a crack is not likely to initiate with stresses below 20 ksi. If a crack does initiate in the uninspected nozzle region and grow toward the J-groove weld, periodic examinations required by N-729-1 will be performed and provide the opportunity to identify the flaw prior to its growing into the weld, which will allow the licensee to take corrective actions.
The licensee's fracture mechanics calculation demonstrates that a hypothetical crack in the uninspected region will not propagate to the toe of the J-groove weld before the next examination. The licensee performed three dimensional finite element stress calculations of the penetration nozzle. The analysis included operating stresses and the J-groove weld residual stresses, and used the as-designed weld geometry and sizes.
1 For the current relief request, by electronic mail dated September 4, 2012 , the NRC staff asked the licensee to describe how it validated WCAP-16589-P to be applicable to the upcoming examinations and to clarify why the WCAP-16589-P does not fully meet Section 1-3200(a) of N-729-1. To address the validation of the WCAP-16589-P, in the October 15, 2012, submittal, the licensee explained that the stress analysis in WCAP-16589-P was performed using the design weld dimensions specific to WCGS. The stress analYSis demonstrated that the hoop and axial stresses on the nozzle inside and outside surfaces remain below 20 ksi (tensile) over the entire region outside the alternative examination zone. When the WCAP-16589-P analysiS was compared to the requirements of Section 1-2000 of N-729-1, the licensee determined that the requirements of Section 1-2000 of N-729-1 were met. The licensee also reviewed the fracture mechanics analYSis in WCAP-16589-P and compared to the requirements of Section 1-3000 of N-729-1. Since the alternative examination zone is below the J-groove weld, the applicable requirements are those of Section 1-3200. The operating temperature of the head has not changed since the analysis in WCAP-16589-P was performed.
The NRC staff's September 4,2012, request for additional information (RAI) was included verbatim in the licensee's RAI response dated October 15, 2012.
-8 To address how the WCAP-16589-P does not fully meet Section 1-3200(a) of N-729-1, the licensee stated that WCAP-16589-P was prepared prior to the NRC requirement to use N-729-1. WCAP-16589-P referenced EPRI MRP-55 as the source for the crack growth formula used in the analysis, not Appendix 0 of the ASME Code,Section XI as required by N-729-1.
However, since the same formula for crack growth rate is used in both EPRI MRP-55 and Appendix 0, there is no technical difference, and WCAP-16589-P does meet the technical requirements for Section 1-3200(a) of N-729-1 as discussed in the October 15,2012, submittal.
The NRC staff asked the licensee to provide the as-designed (i.e., as-analyzed) and as-built dimensions of the J-groove weld and to discuss whether the flaw evaluation in WCAP-16589-P is based on the as-designed or as-built J-groove weld dimensions. If the flaw evaluation was based on the as-designed dimensions, demonstrate that it bounds the flaw evaluation of the as built dimension.
By letter dated October 15, 2012, the licensee provided the length of the as-designed and as-built weld configurations of CRDM nozzle numbers 77 and 78. The actual J-groove weld height was measured using the ultrasonic test data and is listed below for the as-built dimension.
Penetration Nozzle ! As-designed (inches) I As-Built (inches)
Number i 77 1.46 1.98 78 1.46 2.04 The licensee explained that the flaw evaluation in WCAP-16589-P is based on the as-designed J-groove weld dimensions which assumed a smaller weld throat than the as-built condition.
Often, the as-built fillet weld dimension on the downhill side of the CRDM nozzle is larger than the as-designed dimension because of access issues during fabrication. When the weld extends further down the outside surface of the head penetration nozzle due to a larger than as designed fillet, it does not negatively affect the distance below the J-groove weld required for examination coverage. The licensee assessed similar CRDM design/configuration and showed that larger as-built J-groove welds have a reduced stress profile relative to smaller as-designed welds and required lesser distance below the weld bottom for a transition to below 20 ksi. The licensee analyzed weld heights of 1.46, 2.35 and 2.97 inches to determine their resulting stress profiles below the J-groove weld. The 20 ksi criterion is reached in shorter distance for the larger fillet welds. The licensee used the as-designed weld height of 1.46 inches to generate stress distributions. The as-built dimensions of the CRDM nozzles are larger than the as-designed dimension of 1.46 inches. The stress distributions for the as-built configuration should be less than the as-designed configuration. Therefore, the stress distributions in the licensee's stress analysis in WCAP-16589-P bounds the as-built configuration.
The NRC staff concludes that WCAP-16589-P is applicable to the proposed alternative examination distance for CRDM nozzle numbers 77 and 78. WCAP-16589-P does meet the requirements for Section 1-3200(a) of N-729-1 and uses appropriate J-groove weld dimension.
The NRC staff concludes that WCAP-16589-P is acceptable to support the proposed alternative in the upcoming CRDM penetration nozzle examination.
- 9 3.7.2 Examination Coverage Section 4 of the relief request states that penetration nozzle numbers 74 through 78 have a threaded outside diameter and an internal taper, and their configuration poses challenges to achieve the required examination coverage. However, the licensee only requested relief for penetration nozzle numbers 77 and 78. The NRC staff asked the licensee to provide the examination coverage for penetration nozzle numbers 74,75, and 76 and to explain why these three nozzles can meet the required coverage but not nozzle numbers 77 and 78 if they all have the same configuration.
By letter dated October 15,2012, the licensee presented the examination coverage obtained on nozzle numbers 74, 75 and 76 during the 2006 examinations as shown in table below.
Penetration No. e (degrees) N-729-1 Required Inspection Exam Coverage Coverage (inches) Obtained (inches) 74 48.7 1.0 1.04 75 48.7 1.0 1.08 i 76 48.7 1.0 1.12 The licensee explained that the physical process of welding results in slightly different weld sizes/contours being applied to each component. The licensee noted that the CROM nozzles were attached to the reactor vessel head by a manual welding process. As such, it is not physically possible for a human to apply the exact same amount of weld metal to each component. The licensee stated that this (the slight different weld size/contour) is acceptable as long as the minimum design weld size or contour is met. The licensee further stated that when access is limited (as the case on the downhill side of the peripheral penetrations), this condition is magnified. This is the case in nozzle numbers 77 and 78, as more weld metal was applied to the downhill portion of the weld, resulting in less of the nozzle penetration being available for examination (below the toe of the J-groove weld and above the threads).
The NRC staff concludes that the licensee has demonstrated that nozzle numbers 74,75, and 76 have achieved required examination coverage in accordance with N-729-1. The licensee has satisfactorily explained why CROM nozzle numbers 77 and 78 require relieffrom N-729-1 while nozzle numbers 74,75, and 76 do not.
3.7.3 Hardship for Surface Examination Paragraph 10 CFR 50.55a(g)(6)Oi)(0)(3) permits substitution of a surface examination of the inside and outside wetted surface of the penetration nozzle for ultrasonic examination of the portion of a penetration nozzle that is below the toe of the J-groove weld (Point E on Figure 2 of N-729 -1). The NRC staff notes that the threaded portion on the outside diameter of the CROM nozzle is not amenable to eddy current (ET) examination due to the presence of the threads. In addition, the licensee states that performance of a surface examination on the subject nozzles to obtain the required coverage below the lowest pOint in the J-groove weld toe would result in high-radiation dose.
- 10 By letter dated October 1S, 2012, the licensee clarified that the radiation levels are estimated to be 10 roentgen equivalent man (Rem)/hour at the bottom of the CRDM nozzles was based on the survey taken during the CRDM nozzle examinations conducted at WCGS in 2006. The 2006 survey noted the dose rate near the inside of the reactor vessel head dome to be 9.6 Rem per hour. The licensee noted that it would be impractical to try to install shielding at this location to reduce the dose, as it would require more radiation dose to install the shielding than would be saved during the examinations. The licensee stated that it will measure the radiation dose at the bottom of the CRDM nozzles again during the examinations in February 2013.
The NRC staff concludes that the licensee has acceptable plant-specific radiation dose measurements and that it will measure again the radiation does at the bottom of the CRDM nozzles during the examinations in February 2013. Based on the relatively high radiological dose and as low as reasonably achievable (ALARA) considerations, the staff concludes that performing the surface examination to attain the required coverage would present a hardship.
The NRC staff concludes that the proposed examination coverage of 0.6 inches and 0.88 inches are acceptable for CRDM nozzle numbers 77 and 78, respectively. The NRC staff's conclusion is based on the validity of the stress analysis and fracture mechanics calculation in WCAP-16S89-P, demonstrating that within four refueling cycles, a potential flaw that initiates in the unexamined zone (below the J-groove weld) of the CRDM nozzle numbers 77 and 78 will not propagate into the J-groove weld. At the end of every fourth refueling cycle, the licensee will perform an examination to confirm the structural integrity of CRDM nozzles 77 and 78. The structural integrity of nozzle numbers 77 and 78 and associated J-groove welds will be maintained as supported by the periodic examinations and flaw evaluation.
In summary, the NRC staff concludes that the licensee's proposed alternative inspection coverage for CRDM penetration nozzle numbers 77 and 78 provides reasonable assurance of structural integrity and leak tightness until the next scheduled examination, and that compliance with the surface examination requirements of 10 CFR SO.SSa(g)(6)(ii)(D)(3) would result in hardship without a compensating increase in the level of quality and safety.
4.0 CONCLUSION
Based on the above, the NRC staff concludes that the proposed alternative provides reasonable assurance of structural integrity of the CRDM penetration nozzles numbers 77 and 78 and that complying with the ASME Code requirement would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety. Accordingly, the NRC staff concludes that the licensee has adequately addressed all of the regulatory requirements set forth in 10 CFR SO.SSa(a)(3)(ii) and 10 CFR SO.SSa(g)(6)(ii)(D), and authorizes the use of relief request No. 13R-07 at WCGS for the remainder of the third 10-year lSI interval, which ends on September 2, 201S.
- 11 All other requirements in ASME Code,Section XI, and 10 CFR 50.55a for which relief was not specifically requested and approved in this relief request remain applicable, including third-party review by the Authorized Nuclear Inservice Inspector.
Principal Contributor: J. Tsao, DE/EPNB Date: January 4,2013
M. Sunseri - 2 The detailed results of the NRC staff review are provided in the enclosed safety evaluation. If you have any questions concerning this matter, please call Mr. F. Lyon of my staff at (301) 415-2296 or by electronic mail at fred.lyon@nrc.gov.
Sincerely, IRA!
Michael T. Markley, Chief Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-482 Enclosure As stated cc w/encl: Distribution via Listserv DISTRIBUTION:
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