ML20138J015
| ML20138J015 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 05/05/1997 |
| From: | NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20138J009 | List: |
| References | |
| NUDOCS 9705070365 | |
| Download: ML20138J015 (9) | |
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UNITED STATES l
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NUCLEAR REGULATORY COMMISSION 4"
WASHINGTON, D.C. 2065H001 o
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION OF THE SECOND TEN-YEAR INTERVAL INSERVICE INSPECTION PROGRAM PLAN RE0 VEST FOR RELIEF 95-050 FOR l
FLORIDA POWER CORPORATION CRYSTAL RIVER. UNIT 3 DOCKET NO. 50-302
1.0 INTRODUCTION
The Technical Specifications (TS) for Crystal River, Unit 3, states that the l
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 Code (ASME Code) and applicable addenda as required by Title 10 of the Code of Federal Reaulations (10 CFR) 50.55a(g), except where specific written relief has been granted by the Commission pursuant to 10 CFR 50.55a(g)(6)(1).
Section 50.55a(a)(3) states that alternatives to the requirements of paragraph (g) may be used, when authorized by the NRC, if (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 difficulties without a compensating increase in the level of quality and safety.
Pursuant to 10 CFo ;0.55a(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, t
geometry, and materials of construction of the components. The regulations require that inservice examination of components and system pressure tests conducted during the first 10-year interval and subsequent intervah 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) 12 months prior to the l
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 Crystal River, Unit 3, second 10-year inservice inspection (ISI) interval is the 1983 Edition through Summer 1983 Addenda.
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9705070365 970505 PDR ADOCK 05000302 p
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2 Pursuant to 10 CFR 50.55a(g)(5), if the licensee determines that conformance with an examination requirement of Section XI of the ASME Code is not practical for its facility, information shall be submitted to the Commission in support of that determination and a request made for relief from the ASME Code requirement. After evaluation of the determination, pursuant to 10 CFR 50.55a(g)(6)(i), the Commission may grant relief and may impose alternative requirements that are determined to be authorized by law, will not endanger life, property, or the common defense and security, and are otherwise in the public interest, giving due consideration to the burden upon the licensee that could result if the requirements were imposed.
In a letter dated February 6,1996, Florida Power Corporation (licensee),
submitted to the NRC its Second Ten-Year Inservice Inspection Interval Program Plan Request for Relief 95-050, Revision 1, for Crystal River, Unit 3.
The licensee also provided additional information in its letter dated January 15, 1997.
2.0 EVALUATION The staff, with technical assistance from its contractor, the Idaho National Engineering Laboratory (INEL), has evaluated the information provided by the licensee in support.of its Second Ten-Year Inservice Inspection Interval l
Program Plan ~ Request for Relief 95-050, Revision 1, for Crystal River, Unit 3.
I Based on the information submitted, the staff adopts the contractor's conclusions and recommendations presented in the attached Technical Letter Report (TLR).
ASME Section XI, Examination Category B-D, Items B3.90 and B3.100 require 100%
volumetric examination of the reactor pressure vessel (RPV) nozzle-to-vessel welds and nozzle inside radius sections as defined by Figure IWB-2500-7 and require that all nozzles are to be examined by the end of a 10-year interval.
The licensee requested relief from the ASME Section XI requirement of 100%
volumetric examination for the core flood nozzle-to-shell Welds Bl.4.1A and Bl.4.2A, and the core flood nozzle inner radius Sections Bl.4.1B and Bl.4.28.
Component geometry and flow restrictors in the bore of each nozzle obstruct t
I access and preclude performance of the 100% volumetric examination.
These physical ~ restrictions make the Code requirements impractical to inspect the subject core flood nozzles at Crystal River, Unit 3.
To meet the Code requirements, the license would be required to make design modifications to provide access for the examinations. The staff determined that the Code requirements are impractical and that imposition of this requirement would be a burden on the licensee.
The licensee has performed the Code-required volumetric examinations to the extent practical, and obtained a coverage of 85% of the nozzle-to-shell welds and 51% of the nozzle inside radius sections.. Based on:
(1) the percentage of the nozzle-to-vessel welds and inside radius sections that were examined; j
(2) all other RPV nozzle-to-vessel welds and nozzle inside radius sections are i
required to be examined each 10-year inspection interval; and (3) examination i
during preservice exams and results of preservice inspection (PSI) showing no I
i; significant flaws, the staff determined that any existing patterns of degradation would have been detected. The staff further determined that the i
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i coverage obtained by the licensee provides reasonable assurance of the structural integrity of the core flood nozzle-to-shell Welds Bl.4.1A and Bl.4.2A, and the core flood nozzle inner radius Sections 81.4.1B and Bl.4.28.
3.0 CONCLUSION
S The staff has reviewed the licensee's submittal and concludes that the Code-coverage requirements are impractical for the core flood nozzle at Crystal
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River, Unit 3, and that reasonable assurance of the structural Integrity of j
the core flood nozzles has been provided by the best effort volumetric examination in combination with the PSI results and the examination of other i
RPV nozzles.
Based on its review, the staff has determined that the requested l
relief is authorized by law and will not endanger life or property, or the common defense and security, and is otherwise in the public interest given due consideration to the burden that could result if the requirements were imposed on your facility.
Therefore, relief is granted pursuant to 10 CFR 50.55a(g)(6)(1) for Request for Relief 95-050 Revision 1.
Principal Contributor:
T. McLellan Date: May 1, 1997
TECHNICAL LETTEP REPORT ON THE SECOND 10-YEAR INTERVAL INSERVICE INSPECTION U
REQUEST FOR RELIEF 95-050. REVISION 1 FLORIDA POWER CORPORATION CRYSTAL RIVER. UNIT 3 DOCKET NUNBER 50-302
1.0 INTRODUCTION
By letter dated February 6,1996, the licensee, Florida Power Corporation (FPC), requested relief from the Code coverage requirements for the core flood nozzle-to-vessel welds and inside radius sections at Crystal River, Unit 3.
On February 27, 1996, a conference call was held with the licensee to obtain clarification regarding the limitations encountered.
Consequently, the licensee submitted a revised version of the request in a letter dated January 15, 1997. The Idaho National Engineering and Environmental Laboratory (INEEL) staff has evaluated the information provided by the licensee in support of this request for relief in' the following section.
I 2.0 EVALUATION The Code of record for the Crystal River, second 10-year inservice l
inspection (ISI) interval is the 1983 Edition, through Summer 1983 Addenda of Section XI of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code.
Reauest for Relief 95-050. Revision 1. Examination Cateoory B-D. Items B3.90 and B3.100. Core Flood Nozzle-to-Vessel Welds and Inside Radius j
Sections Code Reauirement: ASME Section XI, Examination Category B-D, Items 83.90 and B3.100 require 100% volumetric examination of all reactor pressure vessel (RPV) nozzle-to-vessel welds and nozzle inside radius sections as defined by Figure IWB-2500-7.
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Licensee's Code Relief Reauest:
In accordance with 10 CFR
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- o 50.55a(g)(5)(iii), the licensee has determined that the Code coverage l
requirements are impractical for core flood nozzle-to-shell Welds
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Bl.4.1A and Bl.4.2A, and the core flood nozzle inner radius Sections Bl.4.1B and Bl.4.28 at Crystal River Unit 3.
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Licensee's Basis for Reauestina Relief (as stated):
" Volumetric examinations of the Core Flood Nozzle-to-Shell welds and Nozzle Inner Radius Sections were performed during pre-service examination and first interval inservice examination using.the immersion i
method.
Examination of these areas was limited due to several factors.
Access to the Nozzle Inner Radius from inside the Core Flood Nozzles was prevented by the flow restrictor being located in the bore of each Core Flood Nozzle. The size of the transducer head was not the limiting i
factor in these cases.
The design of the flow restrictor sleeve did not provide acoustic coupling with the Core Flood Nozzle inner wall thus rendering the examination from this direction ineffective. ~ Additional limitations were encountered due to the design of the reactor vessel wall taper associated with the core support ledge, the proximity to the j
adjacent inlet nozzles and the radius blend between the shell and the bore of the nozzle.
Improvements in volumetric examination methods have since shown the contact examination method to be much more reliable.
Modern reactor vessel inspection equipment has been designed to utilize i
this technique.
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"Because examination angles and limitations are the same for both the immersion method and the contact method, ultrasonic examination coverage for the Core Flood Nozzle-to-Shell welds and Nozzle Inner Radius Sections for the previous inservice examination and current interval examination are basically the same. However, the pre-service examination coverage is not directly comparable to the first interval inservice examination or current interval examination because of the differences in examination volumes and the angles used to achieve that coverage.
"The pre-service examination was performed to very early ASME Code j
requirements which only required straight beam (0 degree) and.45 degree angle beam to cover the area of interest.
The area of interest for the 4
pre-service examination included the weld and adjacent base metal for a i
distance of one thickness (T) beyond the weld fusion line. The first j
interval examination required angles of straight beam, 45 deg., 60 deg.,
and 70 degrees. The examination volume also changed to include the weld plus 1/2T of adjacent base metal beyond the' weld fusion line.
For this
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reason, the pre-service examination coverage is not directly comparable to the first interval inservice examination or the current interval inservice examination. Regardless of the ultrasonic method used, the objective of the examination was to obtain the maximum amount for t
i ATTACHMENT 2
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t I-1 coverage of the area of interest permitted by the geometric
. configuration of the scan area. During the pre-service and first intervalinserviceexaminationsthepyrcentcoveragewasnotcalculated to the detail provided in Reference 1. However, based on the coverage calculated for the 1996 examinations and knowledge of the previous examination requirements, estimates of examination coverage have been developed.
"The estimates of composite coverage for pre-service and
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first interval examinations are as follows:
i Pre-service Examinatim i
90%, best estimate for nozzle-to-shell 40%, best estimate for nozzle inner radius (not delineated as separate exam by the Code)
First Interval Examination 87%, best estimate for nozzle-to-shell 51%, best estimate for nozzle inner radius (not delineated as j
separate exam by the Code)
" Crystal River Unit 3 is currently in its second 10-year intervel of i
j operation.
During the last outage of this interval (Refuel 10, 2/96),
1 volumetric examination of the reactor vessel welds including the Nozzle-modern automated reactor vessel inspection equipment.to-Shell we 4
Reference 2 provides a description of the remotely operated manipulator, URSULA, and i
its1996ReactorVesselinspectionscopeutilizjngthecontactmethod i
i with a four array transducer head.
Reference 3 of the contact method versus the immersion method. addresses the advantages This document is a i
Framatome proprietary document and has not been included in this submittal. Advantages discussed in the report include:
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- Very good detection capability on both automatic clad and manual clad components.
- Excellent signal-to-noise ratio.
- Provides good signal response from the crack tips. This enhances crack through-wall sizing techniques.
- Focuses in the maximum sensitivity in the area of interest.
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- It is not sensitive to clad grain orientation.
"The examinations performed during Refuel 10 of the Nozzle-to-Shell welds and Nozzle Inner Radius Sections were limited by the following
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constraints:
- The flow restrictor located in the bore of the Core Flood Nozzle.
ATTACHMENT 2 8
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- The taper above the core flood nozzle associated with the core.
i support ledge.
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- The adjacent reactor vessel cold leg nozzles.
- The blend between the shell. and the bore of the nozzle.
"As previously mentioned, Framatome has developed for Crystal River Unit 3, an ' Explanation of Limited Coverage for Core Flood Nozzle-to-Shell 1
l weld' (Reference 1). As shown in that reference, the 10R coverage for l
the Core Flood Nozzle-to-Shell consisted of 83% of the weld, 80% for the l
adjacent base metal,100% of the near surface field,'and 85% aggregate Aggregate coverage obtained for the Nozzle Inner Radius was coverage.
51%.
" Review of the pre-service and previous inservice inspection results for i
CR-3 have shown no reportable indications in the Core Flood Nozzle-to-Shell welds or Nozzle Inner Radius Sections. The results of the examinations performed during Refuel 10 indicates that there are not ASME code reportable indications in the scanned areas of the Core Flood Nozzle-to-Shell welds or the Core Flood Nozzle Inner Radius Sections.
Additionally, there have been no ASME Section XI Code unacceptable 4
indications for these welds in any of the Framatome inspected operating plants.
" Based upon the amount of coverage obtained, as shown in Reference 1, FPC considers the percentages examined a good representative sample of the Nozzle-to-Vessel welds and Nozzle Inner Radius Sections. The amount of coverage combined with the improved detection capabilities of the contact
-examination method and the welds service history, demonstrates an
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acceptable level of quality and safety."
Licensee's Proposed Alternative (as stated):
" Exam ~ination of the Core Flood Nozzle-to-Shell welds (Bl.4.1A, Bl.4.2A) and the Core Flood Nozzle Inner Radius Sections (Bl.4.lB, 81.4.28) were performed during Refuel 10 (Spring 1996) from the inside surface of the reactor vessel to the maximum extent practical within the limitations of design and geometry as shown in Reference 1.
Examination of those i
portions of the Core Flood Nozzle-to-Shell welds and the Core Flood Nozzle Inner Radius Sections which were limited, are documented in the Explanation of Limited Coverage report (Reference 1). The total i
aggregate coverage obtained for the Core Flood Nozzle-to-shell welds was 85%. The total aggregate coverage for the Nozzle Inner Radius was 51%."
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Evaluation: ASME Section XI requires 100% volumetric examination of all RPV nozzle-to-vessel welds and inside radius sections. However, component geometry and flow restrictors in the bore of each nozzle obstruct access and preclude performance of the 100% volumetric j
L examination for the core flood nozzle-to-vessel welds and nozzle inside radius sections. These physical restrictions make the Code coverage requirements impractical for the core flood nozzles at Crystal River Unit I
3.
To meet the Code requirements, design modifications would be needed to provide access for examination.
Imposition of this requirement would result in a burden on the licensee.
l The licensee has performed the Code-required volumetric examinations to t
the extent practical, obtaining 85% coverage for the core flood nozzle-to-shell welds and 51% coverage for the nozzle inside radius sections.
In addition, all other RPV nozzle-to-vessel welds and nozzle inside j
radius sections are required to be examined each 10-year inspection interval. Therefore, based on the significant percentage of coverage obtained on the core flood nozzle-to-vessel welds and inside radius sections, and the examination of the other RPV nozzles, it is concluded that any existing patterns of degradation would have been detected and that reasonable assurance of the structural integrity has been provided.
3.0 CONCLUSION
1 The INEEL staff has reviewed the licensee's submittal and concludes that the Code coverage requirements are impractical for the core flood nozzles at Crystal River Unit 3.
Furthermore, reasonable assurance of the structural integrity of-the core flood nozzles has been provided by the best effort volumetric examination in combination with the examination of other RPV nozzles.
Therefore, it is recommended that relief be granted pursuant to 10 i
CFR 50.55a(g)(6)(i) for Request for Relief 95-050, Revision 1.
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4.0 REFERENCES
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Attachment to Licensee's Submittal, " Explanation of Limited Coverage for d.
Core Flood Nozzle-to-Shell Welds and Nozzle Inner Radius Sections."
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Attachment to Licensee's Submittal, pictures depicting "URSULA" l
Manipulator.
Babcock L W11cox, Report to the N0G-NDE Comufttee on the Effectiveness i
of the Near Surface Examination Technique Use for Reactor Vessel Examinations, Report No.
1151476.
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