ML20237D221
| ML20237D221 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 08/14/1998 |
| From: | Beasley J SOUTHERN NUCLEAR OPERATING CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| LCV-1016-M, NUDOCS 9808250168 | |
| Download: ML20237D221 (23) | |
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> .c Southern Nuclear Opercting Csmpany. Inc. ,
Post Office Box 1295 Birmingham, Alabama 352011295 Tel 205.992.5000 SOUTHERN COMPANY August 14, 1998 Energ ro serve br mu-LCV-1016-M Docket Nos.: 50-424, 50-425 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555 Ladies and Gentlemen:
VOGTLE ELECTRIC GENERATING PLANT SECOND TEN-YEAR INSERVICE INSPECTION PROGRAM REOUESTS FOR RELIEF Southern Nuclear Operating Company (SNC), the licensee and operator of the Georgia Power Company-owned Vogtle Electric Generating Plant (VEGP), Units 1 and 2, submits herein two requests for relief, RR-26 and RR-31, for the Inservice Inspection (ISI)
Program for the Second Ten-Year Interval. The ISI Program for the two VEGP units was ,
written to the requirements of the 1989 Edition of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section XI, except where relief has been granted by the NRC. The enclosed requests for relief supersede or otherwise supplement portions of the ISI Program previously submitted to the NRC.
l Request for Relief RR-26 was originally submitted to the NRC on May 29,1997 for j review and approval. The request for relief proposed the use of ASME Section XI Code Case N-533, " Alternative Requireraents for VT-2 Visual Examination of Class 1 Insulated Pressure-Retaining Bolted Connections", in lieu of the requirements of the 1989 Edition of ASME Section XI. It was the desire of SNC to extend the Code Case to Class 2 and 3 components for reasons including, but not limited to, personnel safety and i i radiation exposure. Additionalinformation was requested by the NRC staff and the i original request for relief was revised and re-submitted to the NRC on September 10, 1997. By NRC letter dated October 24,1997, RR-26 was approved by the NRC for use but was limited to Class 1 only. The use of the proposed alternative for Class 2 and 3 systems was denied by the NRC which indicated that the frequency of once per hokf ,
9808250168 980814-PDR 0 ADOCK 05000424 PDR
U. S. Nuclear Regulatory Commission LCV-10l*6-M Page Two inspection period for insulation removal had not been shown to satisfy 10 CFR j
50.55a(a)(3) authorization criteria. Subsequent to that time, the ASME Code has taken action to revise ASME Section XI Code Case N-533 to also include Class 2 and 3 systems. As a result, SNC has revised RR-26 to use the guidance of the proposed Code Case revision. In addition, where the request for reliefis applied to Class 2 and 3 systems, the frequency between insulation removals will not exceed forty (40) months with the following exception: -
If a reactor refueling outage is in progress or is scheduled to begin within six months when the 40 months expires, insulation removal and the VT-2 visual examination ' l would be allowed to be deferred provided that they are performed prior to plant startup following the reactor refueling outage.
The proposed alternative to the Code requirements, plus the 40-month frequency J restriction (with the above exception) proposed by SNC, provides an acceptable level of quality and safety' for bolted connections in borated Class 2 and 3 systems and should be authorized by the NRC pursuant to 10 CFR 50.55a(a)(3)(i). Please refer to the enclosed Request for Relief RR-26 (ISI Program Revision 2) for additional details. The enclosed revision of RR-26 supersedes the version submitted to the NRC by our September 10, 1997 letter.
Request for Relief RR-31 is a new request for reliefand proposes an alternative schedule
. for examining the Reactor Pressure Vessel (RPV) shell-to-flange weld. This request for reliefinvolves rescheduling the first inspection period examinations, i.e., fifty percent (50%) of the RPV shell-to-flange weld length (from the flange face), such that the entire
. weld length will be examined at or near the end of the ten-year inspection interval. The 4 I
50% of the weld length examined from the flange face during the first inspection period of the First Ten-Year Interval was examined during the 1R6 and 2R6 maintenance / refueling outages at VEGP-1 and 2, respectively. This was done voluntarily by SNC in order to "re-zero" the examination so that no more than ten years would elapse before being examined again at or near the end of the current Second Ten-Year Interval contingent upon NRC approval of Request for Relief RR-31. Neither of the RPV !
shell-to-flange welds on VEGP-1 and 2 have any inservice repairs or replacements by ,
welding nor any identified flaws or relevant conditions that currently require successive )
inspections. As a result, the proposed alternative to the Code requirements provides an acceptable level of quality and safety and should be authoiized by the NRC pursuant to 10 CFR 50.55a(a)(3)(i). Please refer to the enclosed Request for Relief RR-31 (ISI Program Revision 2) far additional details, including a technical justification for the '
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U. S. Nuclear Regulatory Commission LCV-10f6-M Page Three proposed alternative to the schedule requirements of ASME Section XI.
Your prompt review and approval of the enclosed Requests for Relief RR-26 and RR-31 is respectfully requested so that they may be used during the VEGP-1 maintenance / refueling outage currently scheduled to begin February 28,1999. Because any denial of RR-26 and RR-31 will result in our having to contract for additional support personnel or for an automated RPV inspection tool for the upcoming VEGP-1 outage, we request that relief be authorized by November 1,1998.
A copy of this submittal is being provided directly to Mr. M. T. Anderson ofINEEL Research Center based on an earlier NRC request of August 8,1997.
l Should there be any questions in this regard, please contact this office at your earliest j convenience.
f Sincerely, g
J. B. Beas ey, r.
Vice President, Vogtle Project 1
i JBB/JAE/jae
Enclosures:
- 1. Request for Relief RR-26 (ISI Program Revision 2)
- 2. Request for Relief RR-31 (ISI Program Revision 2) xc: (see next page for distribution) '
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U. S. Nuclear Regulatory Commission L LdV-10f6-M Page Four xc: INEEL Research' Center Mr. M. T. Anderson (w/ enclosures)
Southern Nuclear Operating Company Mr. W. L. Burmeister (w/o enclosures)
' Mr. J. T. Gasser (w/o enclosures)
Mr. M. Sheibani (w/ enclosures)
S14C Document Management (w/ enclosures)
U. S. Nuclear Regulatory Commission Mr. D. II. Jaffe, Senior Project Manager, NRR (w/ enclosures)
Mr. L. A. Reyes, Regional Administrator (w/ enclosures)
Mr. J. Zeiler, Senior Resident Inspector, Vogtle (w/ enclosures)
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ENCLOSURE 1 TO SOUTIIERN NUCLEAR OPERATING COMPANY LETTER LCV-1016-M VOGTLE ELECTRIC GENERATING PLANT SECOND TEN-YEAR INSERVICE INSPECTION PROGRAM REOUEST FOR REI IEF RR-26 Request for Relief RR-26 as submitted herein supersedes the similar request for relief originally submitted to the NRC in SNC letter LCV-1016 dated May 29,1997, and later in SNC letter LCV-1016-A dated September 17,1997. Request for Relief RR-26 (ISI Program Revision 2) immediately follows this introduction. The page numbers which are found in the enclosed request for relief correspond with the page numbers in the VEGP-1 and 2 Inservice Inspection Program-Second Ten-Year Interval which accompanied the SNC letter of May 29,1997.
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l SOUTIIERN NUCLEAR OPERATING COMPANY l VOGTLE ELECTRIC GENERATING PLANT, UNITS 1 AND 2 SECOND TEN-YEAR INTERVAL i REQUEST FOR RELIEF NO. RR-26 l
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l I. System / Component (s) for Which Reliefis Requested: I ASME Class 1,2, and 3 bolted connections in borated systems. This includes the following systems which are borated for the purposes of reactivity control:
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e Reactor Coolant System, e Chemical and Volume Control System, e Residual Heat Removal System,
. Safety injection System, and
- Nuclear Sampling System - Liquid.
II. Code Requirement:
Subparagraph IWA-5242(a) of the 1989 Edition of ASME Section XI states in part "For systems borated for the purposes of controlling reactivity, insulation shall be removed from pressure i
retaining bolted connections for visual examination VT-2."
III. Code Requirement from Which Reliefis Requested:
l Reliefis requested from removing insulation from pressure-retaining bolted connections and l performing VT-2 visual examinations for the purpose of detecting boric acid residue when systems are at the pressure and temperature requirements ofIWA-5000, IWB-5000, IWC-5000, and IWD-5000.
IV. Basis for Relief:
Subparagraph IWA-5242(a) specifies that insulation must be removed from pressure-retaining bolted connections for VT-2 visual examination during the performance of system pressure testing. This is applicable to the following systems:
= Reactor Coolant System (System consists of Class 1 and 2 components),
e Chemical and Volume Control System (System consists of Class 1,2, and 3 components),
e Residual Heat Removal System (System consists of Class 1 and 2 components),
e Safety Injection System (System consists of Class 1 and 2 components), and
- Nuclear Sampling System - Liquid (System consists of Class 2 components).
i 7-105 Rev.2
SOUTIIERN NUCLEAR OPERATING COMPANY VOGTLE ELECTRIC GENERATING PLANT, UNITS 1 AND 2 SECOND TEN-YEAR INTERVAL
, REQUEST FOR RELIEF NO. RR-26 (continued)
IV. Basis for Relief (continued):
Class 1 components:
Table IWB-2500-1, Examination Category B-P requires a system leckage test (IWB-5221) and corresponding VT-2 visual examination on Class I components each refueling outage prior to plant startup. This system leakage test is performed in Mode 3 when the Reactor Coolant System is at Nominal Operating Pressure (= 2235 psig) and Nominal Operating Temperature
(= 550 F to 650 F). The majority of the Class I components are in the Reactor Coolant System; however, some portions of the Class 1 boundary extend to include portions of Safety Injection, Chemical and Volume Control, and Residual Heat Removal systems. All Class I components are in containment. The removal and installation ofinsulation during the performance of system pressure testing inside containment presents the following hazards:
Increased potential for debris to be in containment which could migrate to the Containment Emergency Sumps and restrict the suction of the Emergency Core Cooling System during accident conditions, such as a Loss of Coolant Accident (LOCA). All debris is required to l be removed from containment prior to entering Mode 4, e Increased potential for personnel heat stress since the containment ambient temperature may be a high as 100'F, e
Increased potential for personnel burn injuries due to installation ofinsulation in proximity of extremely hot components, e Increased personnel safety hazard since ladders would have to be used to inspect many of the bolted connections and replace the insulation. Temporary work platforms / scaffolding inside containment are removed prior to entering Mode 4, e Increased radiation exposure to personnel since temporary shielding is removed prior to entering Mode 4, and l
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e Increased potential for impacting outage duration due to the amount of manpower required to support insulation removal and examinations during Mode 3 following refueling outage activities.
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SOUTHERN NUCLEAR OPERATING COMPANY
, VOGTLE ELECTRIC GENERATING PLANT, UNITS 1 AND 2 SECOND TEN-YEAR INTERVALS I REQUEST FOR RELIEF NO. RR-26 (continued)
IV. Basis for Relief (continued):
Class 2 components: 1
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Table IWC-2500-1, Examination Category C-H requires a system pressure test (IWC-5221) i during system functional and system inservice tests and corresponding VT-2 visual examination on Class 2 components once each inspection period. The following discusses the applicable l systems and the basis for relief for each Class 2 system:
- 1. Basis for Relief for the Reactor Coolant System (RCS):
The Class 2 portions of RCS are located adjacent to the Class 1 boundary and are classified as Class 2 based on line size and isolation valve criteria. The system inservice tests for these Class 2 portions of RCS are VT-2 examined in Modes 1,2, and 3 and therefore the same basis for relief as provided for Class 1 applies. These Class 2 pressure boundaries are located in containment.
- 2. Basis for Relief for the Chemical and Volume Control System (CVCS):
For those portions of CVCS which are located inside containment (e.g., Charging, l Letdown, Excess Letdown, Alternate Pressurizer Spray, Reactor Coolant Pump Seal Leakoff) the same basis for relief as provided for Class I above applies except that the l '
system operating temperatures are less resulting in less potential for burn-related injuries.
The VT-2 examinations are performed in Modes 1,2, and 3.
~ For those portions of CVCS which are located outside containment, radiation levels, high component temperatures (e.g., approx. 290 F for Letdown), and availability of personnel l during non-outage times may preclude removing insulation while the pressure-retaining bolted connections are pressurized. CVCS is inservice during power operation and, as such, many, if not all, components will be uninsulated, VT-2 examined, and reinsulated when the pressure-retaining bolted connections are pressurized. However, as previously addressed, conditions may be present which may not allow insulation removal except
! during refueling outages.
- 3. Basis for Relief for the Residual Heat Removal System (RHR):
The RHR System is placed inservice during a shutdown prior to refueling activities in Mode 4 when RCS is = 350 *F and = 350 psig. RHR remains inservice in Modes 5 and 6 7-107 Rev.2 I.
SOUTHERN NUCLEAR OPERATING COMPANY VOCTLE ELECTRIC GENERATING PLANT, UNITS 1 AND 2 SECOND TEN-YEAR INTERVAL
, REQUEST FOR RELIEF NO. RR-26 (continued)
IV. Basis for Relief (continued):
Class 2 components (continued):
- 3. Basis for Relief for the Residual Heat Removal System (RHR) (continued):
during the refueling outage and remains inservice in Mode 4 during startup following the refueling outage. The VT-2 examinations are performed in either Mode 4 or Mode 5 when I the RCS is = 350 psig.
l For those portions of RHR which are located inside containment, the same basis for relief l
as provided for Class 1 above applies except that the system operating temperatures are less, resulting in less potential for burn-related injuries. It is impractical to attempt to limit VT-2 examinations to Mode 5 in order to avoid the complications of performing RHR pressure tests in Mode 4. The VT-2 examinations are performed in Modes 4 or 5.
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- For those portions of RHR which am located outside containment, radiation levels, high l component temperatures, availability of personnel, and increased thermal loads on chilled water room cooling systems may preclude removing insulation while the pressure-retaining
' bolted connections are pressurized. It is significantly more prudent to uninsulate, VT-2 examine, and reinsulate the pressure-retaining bolted connections in RHR when the system is not pressurized during non-outage times or during refueling outages when the system is not at the required pressure.
- 4. Basis for Relief for the Safety Injection System (SI):
The system pressure tests performed on SI are either system functional tests or system inservice tests as follows:
Some of the system functional tests are performed during Modes 1,2, and 3 when RCS l
pressure is greater than SI pump discharge pressure. VT-2 examinations are performed on L
portions of SI during various activities and tests which require a SI pump to be in operation. The scope of these VT-2 examinations includes components located both inside l l and outside containment. The performance of the VT-2 examinations during these activities is generally performed in less than one hour to minimize run time on the SI pumps.
The remainder of the system functional tests are performed during Mode 6 and defueled conditions with the reactor vessel head removed. VT-2 examinations are performed on portions of SI which are pressurized during check valve flow testing activities which 7-108- Rev.2 L_______-.__._--_--_- - - - - - - _ _ . - - - - - - - _
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SOUTHERN NUCLEAR OPERATING COMPANY I
VOGTLE ELECTRIC GENERATING PLANT, UNITS 1 AND 2 l SECOND TEN-YEAR INTERVAL REQUEST FOR RELIEF NO. RR-26 j l
(continued) )
IV. Basis for Relief (continued):
Class 2 components (continued):
- 4. Basis for Relief for the Safety Injection System (SI) (continued):
involve injection of water into the reactor pressure vessel. The scope of these VT-2 examinations includes components located both inside and outside containment. The l
performance of the VT-2 examinations during these activities is generally performed in less than one hour to minimize run time on the applicable pumps and to minimize the impact on critical path testing during refueling outages. j i
Some of the system inservice tests are performed during Modes 1,2, or 3 on portions of SI which are pressurized by the SI accumulator tanks. The SI accumulator tanks are generally depressurized during refueling outages. The scope of these VT-2 examinations includes ;
components which are located only inside containment. j l
l The remainder of the system inservice tests are performed on portions of SI which are l pressurized by the static head of the refueling water storage tank. The VT-2 examinations on these portions of SI are generally performed during power operation (Mode 1) but may be performed in other Modes if tank levels are adequate. The scope of these VT-2 examinations includes components which are located only outside containment.
l For those portions of SI which are located outside containment, radiation levels and availability of personnel during non-outage times or during system functional testing may preclude removing insulation while the pressure-retaining bolted connections are pressurized. l
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For those portions of SI which are located inside containment and VT-2 examined dunng Modes 1,2, and 3, the same basis for relief as provided for Class 1 above applies except that the system operating temperatures are less, resulting in less potential for burn-related l
injuries.
For those portions of SI which are located inside containment and VT-2 examined during Mode 6 and defueled conditions with the reactor vessel head removed, containment radiation levels and availability of personnel during system functional testing may preclude removing insulation while the pressure-retaining bolted connections are pressurized.
I i 7-109 Rev.2
SOUTHERN NUCLEAR OPERATING COMPANY VOGTLE ELECTRIC GENERATING PLANT, UNITS 1 AND 2 SECOND TEN-YEAR INTERVAL REQUEST FOR RELIEF NO. RR-26 (continued)
IV. pasis for Relief (continued):
Class 2 components. (continued):
- 5. Basis for Relief for the Nuclear Sampling System - Liquid:
The liquid portions of the Nuclear Sampling System are used for providing samples for analysis purposes of the RCS, CVCS, and RHR. This system is located both inside and
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outside containment and is subject to the same system pressure tests as the systems for which a ,s used to provide samples. Therefore, the same basis for relief as discussed above for RCS, CVCS and RHR is applicable to the liquid portions of the Nuclear Sampling System.
Class 3 components:
Table IWD-2 J00-1, Exarnination Category D-A requires a system inservice test (IWD-5221) and corresponding VT-2 visual examination of Class 3 components once each inspection period. l Subpa agraph IWA-5242(a) is applicable to the boric acid storage tank and boric acid transfer portions of CVCS. System inservice tests are performed as follows:
l Some of the system inservice tests are performed on portions of CVCS which are pressurized by the static heaii af the boric acid storage tank. The scope of these VT-2 examinations includes components which are located only outside containment. The VT-2 examinations on these l
portions of CVCS r.re generally performed during power operation (Mode 1) but may be performed in other Modes if tank levels are adequate. These system inservice tests are generally performed during power operation and, as such, many, if not all. components will be uninsulated, VT-2 examined, and reinsulated when the pressure-retaining bolted connections are pressurized.
However, availability of personnel during non-outage times may preclude removing insulation while the pressure-retaining bolted connections are pressurized.
The remainder of the system inservice tests are performed on portions of CVCS which are pressurized when a boric acid transfer pump is operating. The scope of these VT-2 examinations includes components which are located only outside containment. The VT-2 examinations on l
these portions of CVCS are generally performed during power operation (Mode 1) with a boric acid transfer pump running with system valves aligned in a recirculation flowpath which precludes injecting high concentrations of borated water into CVCS and ultimately into the RCS.
The boric acid transfer pumps are operated as necessary to perform system functions and necessary testing and, as such, are ne continuously in operation. Since these pumps are not l
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SOtTTHERN NIICLEAR OPERATING COMPANY VOGTLE ELECTRIC GENERATING PLANT, LINITS I AND 2 SECOND TEN-YEAR INTERVAL REOlIEST FOR RELIEF NO. RR-26 (continued)
IV. Basis for Relief (continued):
Class 3 components (continued):
continuously in operation, availability of personnel during non-outage times may preclude removing insulation while the pressure-retaining bolted connections are pressurized.
The 1983 Edition through Summer 1983 Addenda of ASME Section XI was applicable for the First Ten-Year Interval at VEGP. IWA-5242 of the 1983 Edition through Summer 1983 Addeada of ASME Section XI did not require insulation removal; therefore, this request for relief was not needed at VEGP during the First Ten-Year Interval.
V. Alternate Examination:
Class I components:
Insulated Class I pressure-retaining bolted connections will be uninsulated and VT-2 examined once each refueling outage while the connections are at atmospheric or static pressures. The bolted connections will be examined again with the insulation installed during the regularly scheduled system pressure test at nominal system operating temperature and pressure as required by Table IWB-2500-1, Examination Category B-P.
Class 2 and 3 components:
Insulated Class 2 and 3 pressure-retaining bolted connections will be uninsulated and VT-2 examined once per inspection period, when the connection is not at p> essure. In addition, where this request for reliefis applied, the frequency between the insulation removals (including the VT-2 examination) will not exceed forty (40) months except as follows. If a reactor refueling outage is in progress or is scheduled to start within six months when the 40 mor.ths expires, insulation removal and the VT-2 examination would be allowed to be deferred provided that they are performed prior to plant startup following the reactor refueling outage. These examinations may be performed when the connections are not at the pressures required by IWA-5000, IWC-5000, and IWD-5000.
The bolted connections will be examined again with the insulation installed during the regularly scheduled (once per inspection period) system pressure test as required by Table IWC-2500-1, Examination Category C-H, and Table IWD-2500-1, Examination Category D-A.
7-111 Rev.2
SOtJTIIERN NINEAR OPERATING COMPANY VOGTLE ELECTRIC CENERATING PLANT, llNITS 1 AND 2 SECOND TEN-YEAR INTERVAL REOLIEST FOR RELIEF NO. RR-26 (continued) 1 VL Justification for Grantine Relief:
Class I components:
The following items are justification for granting relief for Class 1 Components:
- 1. Evidence ofleakage through pressure-retaining bolted connections which are in systems which are borated for the purpose of controlling reactivity is readily detectable by visual observation when systems are not at operating temperature and pressure. The boric acid concentrations are sufficiently high such that boric acid residues will be present ifleakage has occurred at the pressure-retaining bolted connection.
- 2. In addition, the ASME Section XI Code Committee has issued Code Case N-533 (copy l provided as Attachment I to this request for relief) which allows as an alternative for Class 1 pressure-retaining bolted connections that insulation may be removed and VT-2 examined when the connection is not pressurized. The Code Case also requires that any evidence ofleakage be evaluated in accordance with IWA-5250. Refer to Request for Relief RR-25 for details concerning relief from IWA-5250(a)(2).
3, Compliance with the Code presents the hardships previously discussed which are:
. Increased potential for debris to be in containment which could migrate to the Containment Emergency Sumps and restrict the suction of the Emergency Core Cooling System during accident (LOCA) conditions. All debris is required to be removed from containment prior to entering Mode 4,
. Increased potential for personnel heat stress since the containment ambient temperature may be as high as 100 F,
. Increased potential for personnel burn injuries due to installation ofinsulation in l proximity of components greater than 500 F, l
. Increased personnel safety hazard since ladders would have to be used to inspect many of the bolted connections and replace the insulation. Temporary work platforms / scaffolding inside containment are removed prior to entering Mode 4,
. Increased radiation exposure to personnel since temporary shielding is removed prior to entering Mode 4, and 7-llla Rev.2 l
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SOLITHERN NIICLEAR OPERATING COMPANY VOGTLE ELECTRIC GENERATING PLANT,(INITS 1 AND 2 SECOND TEN-YEAR INTERVAL l REOtIEST FOR RELIEF NO. RR-26 1 (continued)
VL Justification for Grantine Relief (continuedh Increased potential for impacting outage duration due to the amount of manpower required to support insulation removal and examinations during Mode 3 following refueling outage activities.
- 4. For the reasons discussed above, SNC has determined that implementation of the proposed alternatives to the Code requirements provides an acceptable level of quality and safety and therefore requests that the proposed alternative be authorized pursuant to 10 CFR 50.55a(a)(3)(i).
Class 2 and 3 components:
The proposed alternative provides the following two-phase methodology for ensuring the integrity of the pressure-retaining bolted connections.
- 1. Removing insulation and performi.'g a visual examination (at the modified time period specified in the Alternate Examination), when the pressure-retaining bolted connection is not at pressure, will allow for detection of previously occurring leakage through the presence of boric acid crystals. Boric acid concentrations are sufficiently high such that boric acid residues will be present and can be visually observed ifleakage has occurred at the pressure-retaining bolted connection.
- 2. Performing a system pressure test on the bolted connection with the insulation in place, utilizing the Code specified holding time to allow time for any leakage to penetrate the insulation, will provide a means of detecting any significant leakage.
The proposed alternative provides reasonable assurance that the structural integrity of the pressure-retening bolted connections will be maintained, thereby, continuing to provide an acceptable level oflevel of quality and safety. Therefore, approval of this proposed alternative should be authorized pursuant to 10 CFR 50.55a(a)(3)(i).
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SOUTHERN NUCLEAR OPERATING COMPANY VOGTLE ELECTRIC GENERATING PLANT, UNITS 1 AND 2 SECOND TEN-YEAR INTERVAL REQUEST FOR RELIEF NO. RR-27 (continued)
VII. Implementation Schedule:
This request for reliefis applicable to the Second Ten-Year Interval which commenced May 31, 1997.
VIII. Status:
The Class 1 portion of this request for relief was previously approved by the NRC on October 24,1997 and remains virtually unchanged (except for minor editorial changes) due to this revision to RR-26.
Reliefis requested by this revision of RR-26 for Class 2 and 3 as discussed herein and is awaiting NRC review and approval.
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9 SOUTHERN NUCLEAR OPERATING COMPANY VOGTLE ELECTRIC GENERATING PLANT, UNITS 1 AND 2 SECOND TEN-YEAR INTERVAL REQUEST FOR RELIEF NO. RR-26 I
(Continued) j ATTACHMENT 1 CASE N-533 CASES OF ASME BOILER AND PRESSURE VESSEL CODE Approval Date: March 14,1995 See NumericalIndex for expiration and any restfirmation dates.
Case N 533 Alternative Requirements for VT-2 Visual Examination of Class 1 Insulated Pressure-Retaining Bolted ConnectionsSection XI, Division 1 Inquiry: What alternative requirements may be used in lieu of those of IWA 5242(a) to remove insulation from Class 1 pressure-retaining bolted connections to perform a VT-2 visual examination?
Reply: It is the opinion of the Committee that, as an alternative to the requirements of IWA-5242(a) to remove insulation from Class 1 pressure-retaining bolted connections to perform a VT-2 visual exami-nation, the following requirements shall be met.
(a) A system pressure test and VT-2 visual exam-ination shall be performed each refueling outage without removal of insulation.
(b) Each refueling outage the insulation shall be removed from the bolted connection, and a VT-2 vis-ual examination shall be performed. The connection is not required to be pressurized. Any evidence of leakage shall be evaluated in accordance with IWA-5250.
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n ENCLOSURE 2 TO SOUTHERN NUCLEAR OPERATING COMPANY j LETTER LCV-1016-M VOGTLE ELECTRIC GENERATING PLANT SECOND TEN-YEAR INSERVICE INSPECTION PROGRAM l REOUEST FOR REI IEF RR-31 New Request for Relief RR-31 is being added to the VEGP-1 and 2 Inservice Inspection Program-Second Ten-Year Interval which was submitted originally to the NRC by SNC letter LCV-1016 dated May 29,1997. . The page numbers which are found in the enclosed l request for relief correspond with the page numbers being added by Revision 2 of the
!- Program document.
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SOUTIIERN NUCLEAR OPERATING COMPANY VOGTLE ELECTRIC GENERATING PLANT, UNITS 1 AND 2 SECOND TEN-YEAR INTERVAL REQUEST FOR RELIEF NO. RR-31 I. System / Component (s) for Which Reliefis Requested:
Volumetric examination of Reactoc Pressure Vessel (RPV) Shell-to-Flange (B-A) welds,11201-V6-001-WO3 and 21201-V6-001-WO3.
II. Code Requirement:
Item No. Bl.30, Examination Category B-A, Table IWB-2500-1 of ASME Section XI requires a volumetric examination of RPV Shell-to-Flange welds as described in Footnotes 3 and 4 of the referenced table. Footnote 3 states: "If' partial examinations are conducted from the flangt face, the remaining volumetric examinations required to be conducted from the vessel wall may be performed at or near the end of each inspection interval." Footnote 4 states: "The examination of shell-to-flange welds may be performed during the first and third inspection periods in conjunction with the nozzle examinations of Examination Category B-D (Program B). At least fifty percent (50%) of the shell-to-flange weld length shall be examined by the end of the first inspection period, and the remainder by the end of the third inspection periods."
III. Code Requirement from Which Reliefis Requested:
As an alternative to the existing schedule, reliefis requested to allow the rescheduling of fifty percent (50%) of the examinations of the RPV Shell-to-Flange weld (flange face) from the end of the first inspection period to at or near the end of the ten-year inspection interval. Footnote 4, as referenced above, requires at least 50% of the shell-to-flange weld length to be examined by the end of the first inspection period. This rescheduling would allow examination of the RPV Shell-to-Flange and other RPV welds or appurtenances, i.e., nozzle-to-shell welds, nozzle inside radius, and nozzle-to-safe end welds, as addressed in Request for Relief RR-2, to be performed as part of the " Ten-Year Inservice Inspection" at which time the RPV is usually examined. Refer to the attached Tables 1 and 2 which show the current schedule and the proposed schedule, respectively.
IV. Basis for Relief:
To comply with Table IWB-2500-1 during the First Ten-Year Interval,50% of the RPV Shell-to-Flange weld length was scheduled and namined from the flange face during the first inspection period. The remaining 50% of the weld length was scheduled and examined from the flange face during the third inspection period as allowed by Footnote 4 of Table IWB-2500-1 With the partial examinations being conducted from the flange face, the remaining volumetric examinations from the vessel wall, were scheduled and examined at or near the end of the First Ten-Year Interval as allowed by Footnote 3 of Table IWB-2500-1.
i l
7-127 Rev.2
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\,
SOUTHERN NUCLEAR OPERATING COMPANY VOGTLE ELECTRIC GENERATING PLANT, UNITS I AND 2 SECOND TEN-YEAR INTERVAL REOUEST FOR RFI IEF NO. RR-31 (continued) !
IV. Basis for Relief (continued):
At the end of the First Ten-Year Intervals during the IR6 and 2R6 maintenance / refueling outages at VEGP-1 and 2, respectively, one hundred percent (100%) of the RPV Shell-to-Flange we!d length was examined from the flange face, and 100% of the same weld length was examined from the vessel wall. These examinations were pan of those performed for the First Ten-Year Interval.
Southern Nuclear Operating Company has concluded that rescheduling of these examinations such that they are performed at or near the end of the inspection interval will have little, if any, effect on the quality of examinations while providing a substantial benefit to SNC. This conclusion is based on the following:
- 1. Examinatio.nfuality - This request for reliefinvolves rescheduling the first period examinations (50% of the RPV Shell-to-Flar.ge weld length, from the flange face) such that the entire weld length will be examined at or near the end of the inspection interval. Fifty percent of the weld length examined from the flange face during the first inspection period of the First Ten-Year Interval was re-examined during maintenance / refueling outages IR6 and 2R6. This was done voluntarily by SNC in order to "re-zero" the examination so that no more than ten years would elapse before being examined again at or near the end of the Second Ten-Year Interval contingent upon NRC approval of this request for relief. By "re-zeroing" the RPV Shell-to-Flange weld during the First Ten-Year Interval such that no more than ten years would elapse before being examined again at or near the end of the Second Ten-Year Interval, the proposed rescheduling of this examination should have no effect on the quality of the overall RPV Shell-to-Flange examinations.
- 2. Examination Continuity - The potential evaluation of flange face examinations presents difficulties due to limited transducer manipulation, long metd paths, and extensive beam spread limitations (Refer to Figure I which depicts the RPV Shell-to-Flange weld). As a result, accurate sizing technology is not as practical for examinations conducted from the flange face when compared to those conducted from the vessel wall. By using proven automated sizing technology from the vessel wall such as that used during the " Ten-Year Inservice Inspectinn", more accurate sizing results are achievable. With approval of this request for relief, SNC would only use a RPV flange face examination tool and a mechanized RPV inspection tool once at or near the end of the inspection interval. In addition, approval of this request for relief would allow SNC to re-synchronize the examination schedule, thus providing a readily available means of characterizing any indications observed from the flange face examination, should they occur.
7-128 Rev.2 L_...____._ _ J
4 SOUTIIERN NUCLEAR OPERATING COMPANY q
VOGTLE ELECTRIC GENERATING Pl. ANT, UNITS 1 AND 2 -
SECOND TEN-YEAR IN'i ERVAL REQUEST FOR RELIEF '40. RR-31 (continued; i In addition to the foregoing, SNC is anticipating NRC approval to re-schedule the RPV Examination Category B-D and B-F welds or appurtenances, i.e., Nozzle-to-Vessel, Inside Radius Sections, and Nozzle-to-Safe End welds, to the end of the inspection interval, as submitted to the NRC in Request for Relief RR-2. In that request for relief, SNC requested pemlission from the NRC to use ASME Section XI Code Case N-521," Alternative Rules for Deferral ofInspections of Nozzle-to-Vessel Welds, Inside Radius Sections, and Nozzle-to-Safe End Welds of a Pressurized Water Reactor (PWR) Vessel". Use of that Code Case would allow for the rescheduling of Examination Category B-D and B F welds or appurtenances to be examined at or near the end of the inspection interval provided that certain conditions were met, e.g., no inservice repairs or replacements by welding, no identified flaws or relevant conditions that currently require successive inspections, and the unit is not in the first inspection interval.
Although a different Examination Category is involved, i.e., B-A, than those addressed in Code Case N-521, the VEGP-1 and 2 RPV Shell-to-Flange welds meet those same conditions in that examinations from both the flange face (100%) and the vessel wall (100%) were performed at the end of their respective first inspection intervals. In addition, no inservice repairs or replacements by welding have occurred and no flaws or relevant conditions that require successive inspections were identified for that particular weld on either VEGP unit.
Finally, the proposed rescheduling allows SNC significant opportunities for savings in contractor cost, critical path time, radiation exposure, and internal manpower requirements while still maintaining compliance with the examination requirements of the Code. Approval of this request for relief for the RPV Shell-to-Flange (B-A) welds, in conjunction with approval of Request for Relief RR-2, would allow the B-A, B-D, and B-F welds or appurtenances to be examined at the same time. By performing the examinations at the same time, additional radiation exposure reduction can be realized, as well as reductions in mobilization and coordination efforts, set-up time, and examination time. In addition, performance of the examinations at one time at or near the end of the Second Ten-Year Interval ec.utitutes a Cost-Beneficial Licensing Action (CBLA)in that savings in excess of $100,000 are expected to be {'
realized over the remaining lives of the two VEGP units due in part to savings from not having to mobilize the RPV inspection vendor in the first inspection period of the inspection intervals.
i V. Alternate Examination:
The volumetric examinations of the RPV Shell-to-Flange welds will continue to be performed at VEGP-1 and 2. However, the entire length of these welds will be scheduled and examined from the flange face at or near the end of the Second Ten-Year Interval, instead of 50% of the weld length being examined from the flange face in the first inspection period and the remainder in the third inspection period. The entire length of the RPV Shell-to-Flange welds will be scheduled and examined from the vessel wall at or near the end of the Second Ten-Year Interval.
7-129 Rev.2 I
1 l
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SOUTHERN NUCLEAR OPERATING COMPANY VOGTLE ELECTRIC GENERATING PLANT, UNITS 1 AND 2 SECOND TEN-YEAR INTERVAL REQUES1 FOR RELIEF NO. RR-31 (continued)
VI. Justification for Granting Relief:
The proposed alternative will provide an acceptable level of quality and safety. The welds on VEGP-1 and 2 will still be volumetrically examined during the Second Ten-Year Interval except that the subject welds will have their entire length examined from both the flange face and vessel wall at or near the end of the interval rather than examining just a portion of the welds from the flange face during the first inspection period. Approval of the proposed alternative will not result in more than ten years elapsing between the RPV Shell-to-Flange weld examinations conducted during the First Ten-Year Interval and those scheduled for examination during the Second Ten-Year Interval. Therefore, it is requested that this alternative be authorized pursuant to 10 CFR 50.55a(a)(3)(i). Denial of this request for relief would eliminate potential opportunities for savings in contractor costs, radiation exposure, and internal manpower requirements.
VII. Implementation Schedule:
This request for reliefis applicable to the Second Ten-Year Interval which commenced May 31, 1997.
7-130 Rev.2
SOUTIIERN NUCLEAR OPERATING COMPANY {
VOGTLE ELECTRIC GENERATING PLANT; UNITS 1 AND 2 :
SECOND TEN-YEAR INTERVAL f REQUEST FOR RELIEF NO. RR-31 TABLE I -(Current Schedule)
INTERVAL 1 INTERVAL 2 INTERVAL 3 INTERVAL 4 1 2 3 1 2 ? 1 2 3 2 1 3 A 50% of B-50% of A-50% of B-50% of A-50% of B-50~6 of A-50% of B-50% of weld weld weld weld weld weld weld weld length length length length length length length length from from from from from from from from Flange Flange Flange Flange Flange Flange Flange Flange Face Face Face Face Face Face Face Face C-100% C-100% C-100% C-100%
ofweld ofweld ofweld ofweld length length length length from from from from Vessel Vessel Vessel Vessel l Wall Wall Wall Wall TABLE 2 -(Proposed Schedule) l INTERVAL 1 INTERVAL 2 INTERVAL 3 INTERVAL 4 1 2 3 1 2 3 1 2 3 2 1 3 A-50% of B-50% of B-50% of B-50% of B-50% of weld weld weld weld weld length length length length length from from from from from Flange Flange Flange Flange Flange Face Face Face Face Face A-50% of A 50% of A-50% of A-50% of weld weld weld weld length length length length from from from from Flange Flange Flange Flange Face Face Face Face C.100% C-100% C-100% C-100%
ofweld ofweld ofweld ofweld length length length length from from from from Vessel Vessel Vessel Vessel Wall Wall Wall Wall Legend A - Initial 50% of shell-to-flange weld length examined from the flange face.
l B - Remaining 50% of shell-to-flange weld length examined from the flange face.
C - 100% of shell-to-flange weld length examined with mechanized equipment from the vessel wall.
7-131 Rev.2 l
A SOUTHERN NUCI FAR OPERATING COMPANY l VOGTLE ELECTRIC GENERATING PLANT. UNITS 1 AND 2 {
SECOND TEN-YEAR INTERVAL 1 BEQUEST FOR RFI IEF NO. RR-31 VEGP-1 RPV Shell to Flange Weld (11201-V6-001-WO3) l Typical for VEGP-2 J l
l R 91.37 TO CLAD
--- ~1.50
-+
MANUAL FLANGE EXAM:
n f1.50 6*OUT, 3.5'IN j [10*lN & 13.5*lN i R 95.
(REF _ R 86.28 TO CLAD KEYWAYS AT:
0*,90*,180* & 270* 1
/ SEE SHEETS #01 & 43 !
TO TOP =' - 14.50
@ 8S L - 2.01 gE R 83.53 15.05
._ / Il 11 II 21.01 N21.95 LIMIT 5.12 yNOM L
- START OF 28.16 TER 2/ 7 RPV Shell to Flange Wald 5.16 u /
R 85.44 TO CLAD
- 48.84 44,' SCAN
--11.75 N -
-My g
\ - 55.58 -80 SCAN Floure 1 7-132 Rev. 2
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