ML043270051
| ML043270051 | |
| Person / Time | |
|---|---|
| Site: | Duane Arnold  |
| Issue date: | 01/06/2005 |
| From: | Raghavan L NRC/NRR/DLPM/LPD3 |
| To: | Peifer M Nuclear Management Co |
| Raghavan L | |
| References | |
| TAC MC2181 | |
| Download: ML043270051 (13) | |
Text
January 6, 2005 Mr. Mark A. Peifer Site Vice President Duane Arnold Energy Center Nuclear Management Company, LLC 3277 DAEC Road Palo, IA 52324-0351
SUBJECT:
DUANE ARNOLD ENERGY CENTER - RE: ALTERNATIVES FOR EXAMINATION OF REACTOR PRESSURE VESSEL CIRCUMFERENTIAL SHELL WELDS RELIEF REQUEST NDE-R047 (TAC NO. MC2181)
Dear Mr. Peifer:
By letter dated February 12, 2004, Nuclear Management Company, LLC, submitted relief request NDE-R047, requesting relief from the requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code,Section XI, "Rules for Inservice Inspection of Nuclear Power Plant Components," regarding reactor shell weld inspections at the Duane Arnold Energy Center (DAEC).
The U. S. Nuclear Regulatory Commission staff has evaluated the above request and, based on the information provided, concludes that the proposed alternative will provide an acceptable level of quality and safety. Therefore, the proposed alternative under relief request NDE-R047 is authorized pursuant to Section 50.55a(a)(3)(i) of Title 10 of the Code of Federal Regulations, for the remaining term of the operating license for DAEC.
Our safety evaluation is enclosed.
Sincerely,
/RA by M.Kotzalas for L.Raghavan/
L. Raghavan, Chief, Section 1 Project Directorate III Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket No. 50-331
Enclosure:
Safety Evaluation cc w/encl: See next page
Mr. Mark A. Peifer January 6, 2005 Site Vice President Duane Arnold Energy Center Nuclear Management Company, LLC 3277 DAEC Road Palo, IA 52324-0351
SUBJECT:
DUANE ARNOLD ENERGY CENTER - RE: ALTERNATIVES FOR EXAMINATION OF REACTOR PRESSURE VESSEL CIRCUMFERENTIAL SHELL WELDS RELIEF REQUEST NDE-R047 (TAC NO. MC2181)
Dear Mr. Peifer:
By letter dated February 12, 2004, Nuclear Management Company, LLC, submitted relief request NDE-R047, requesting relief from the requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code,Section XI, "Rules for Inservice Inspection of Nuclear Power Plant Components," regarding reactor shell weld inspections at the Duane Arnold Energy Center (DAEC).
The U. S. Nuclear Regulatory Commission staff has evaluated the above request and, based on the information provided, concludes that the proposed alternative will provide an acceptable level of quality and safety. Therefore, the proposed alternative under relief request NDE-R047 is authorized pursuant to Section 50.55a(a)(3)(i) of Title 10 of the Code of Federal Regulations, for the remaining term of the operating license for DAEC.
Our safety evaluation is enclosed.
Sincerely,
/RA by M.Kotzalas for L.Raghavan/
L. Raghavan, Chief, Section 1 Project Directorate III Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket No. 50-331
Enclosure:
Safety Evaluation cc w/encl: See next page DISTRIBUTION:
PUBLIC DBeaulieu Dweaver DLPMDPR PDIII-1 Reading THarris SCoffin WRuland OGC BBurgess, RIII LRaghavan ACRS GHill(2)
ADAMS Accession Number: ML043270051
- per memo dated 7/9/04 OFFICE PDIII-1/PM PDIII-1/LA OGC EMCB/SC PDIII-1/SC NAME DBeaulieu THarris RHoefling MMitchell*
LRaghavan DATE 11/23/04 11/23/04 12/1/04 7/9/04 OFFICIAL RECORD COPY
Duane Arnold Energy Center cc:
Mr. John Paul Cowan Executive Vice President &
Chief Nuclear Officer Nuclear Management Company, LLC 700 First Street Hudson, MI 54016 John Bjorseth Plant Manager Duane Arnold Energy Center 3277 DAEC Road Palo, IA 52324 Steven R. Catron Manager, Regulatory Affairs Duane Arnold Energy Center 3277 DAEC Road Palo, IA 52324 U. S. Nuclear Regulatory Commission Resident Inspectors Office Rural Route #1 Palo, IA 52324 Regional Administrator, Region III U. S. Nuclear Regulatory Commission 2443 Warrenville Road, Suite 210 Lisle, IL 60532-4352 Jonathan Rogoff Vice President, Counsel & Secretary Nuclear Management Company, LLC 700 First Street Hudson, WI 54016 Bruce Lacy Nuclear Asset Manager Alliant Energy/Interstate Power and Light Company 3277 DAEC Road Palo, IA 52324 Daniel McGhee Utilities Division Iowa Department of Commerce Lucas Office Buildings, 5th floor Des Moines, IA 50319 Chairman, Linn County Board of Supervisors 930 1st Street SW Cedar Rapids, IA 52404 Craig G. Anderson Senior Vice President, Group Operations 700 First Street Hudson, WI 54016
ENCLOSURE SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION ALTERNATIVES FOR EXAMINATION OF REACTOR PRESSURE VESSEL CIRCUMFERENTIAL SHELL WELDS NUCLEAR MANAGEMENT COMPANY, LLC DUANE ARNOLD ENERGY CENTER DOCKET NO. 50-331
1.0 INTRODUCTION
By letter dated February 12, 2004, Nuclear Management Company, LLC (the licensee),
submitted Relief Request NDE-R047, requesting relief from the requirements of American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components, regarding examination of reactor pressure vessel (RPV) circumferential shell welds at the Duane Arnold Energy Center (DAEC).
The relief request proposed an alternative to the RPV circumferential shell welds examination requirements of ASME Code,Section XI, for the remaining portion of the license period.
2.0 REGULATORY EVALUATION
Inservice inspection (ISI) of the ASME Code Class 1, 2, and 3 components is performed in accordance with Section XI of the ASME Code and applicable addenda as required by Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(g), except where specific relief has been granted by the U. S. Nuclear Regulatory Commission (NRC) pursuant to 10 CFR 50.55a(g)(6)(i). 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 difficulty without a compensating increase in the level of quality and safety.
Section 50.55a(g)(4) states further that ASME Code Class 1, 2, and 3 components (including supports) must meet the requirements, except design and access provisions and preservice examination requirements, set forth in the ASME Code,Section XI, to the extent practical within the limitations of design, geometry, and materials of construction of the components.
The regulations require that ISI of components and system pressure tests conducted during the first 10-year interval and subsequent intervals comply with the requirements in the latest edition and addenda of Section XI of the ASME Code incorporated by reference in 10 CFR 50.55a(b) 12 months prior to the start of the 120-month interval, subject to the limitations and modifications listed therein.
The applicable ISI Code of Record for the third 10-year ISI interval of the DAEC unit is the 1989 Edition of ASME Section XI.
3.0 TECHNICAL EVALUATION
3.1 Background
Boiling-Water Reactor Vessel and Internals Project (BWRVIP)-05 Report By letter dated September 28, 1995, as supplemented by letters dated June 24 and October 29, 1996, and May 16, June 4, June 13, and December 18, 1997, and January 13, 1998, the BWRVIP submitted the proprietary report BWRVIP-05, BWR Vessel and Internals Project, BWR Reactor Pressure Vessel Shell Weld Examination Recommendations (BWRVIP-05). As modified, the BWRVIP report proposed to reduce the scope of inspection of BWR RPV shell welds from essentially 100 percent of all RPV shell welds to examination of essentially 100 percent of the axial shell welds and essentially 0 percent of the circumferential shell welds, except at the intersections of the axial and circumferential shell welds, thereby including 2 to 3 percent of the circumferential shell welds. In addition, the report included proposals to provide alternatives to ASME Code requirements for successive and additional examinations of circumferential shell welds if flaws are identified.
On July 28, 1998, the NRC staff issued a safety evaluation (SE) on the BWRVIP-05 report.
This evaluation concluded that the failure frequency of RPV circumferential shell welds in BWRs was sufficiently low to justify elimination of ISI of these welds. In addition, the evaluation concluded that the BWRVIP proposals on successive and additional examinations of circumferential shell welds were acceptable. The evaluation indicated that examination of the circumferential shell welds shall be performed if axial shell weld examinations reveal an active, mechanistic mode of degradation.
In the BWRVIP-05 report, the BWRVIP concluded that the conditional probabilities of failure for BWR RPV circumferential shell welds are orders of magnitude lower than that of the axial shell welds. As a part of its review of the report, the NRC conducted an independent, probabilistic fracture mechanics (PFM) assessment of the results presented in the BWRVIP-05 report. The staff conservatively calculated the conditional probability of failure from RPV axial and circumferential shell welds at neutron fluence values corresponding to 32 effective full power years (EFPYs) and 64 EFPY for a BWR nuclear plant, as indicated in Tables 2.6-4 and 2.6-5 of the staffs July 28, 1998, SE. The failure frequency for a RPV is calculated as the product of the event frequency for the critical (limiting) transient and the conditional probability of failure for the weld.
1 The staff has identified that the NRC SE on BWRVIP-05 is mistakenly referenced as dated on July 30, 1998, in GL 98-05. For clarification purposes the staff wants to add that this SE is a letter addressed to Carl Terry, BWRVIP Chairman, dated July 28, 1998, and titled "Final Safety Evaluation of the BWR Vessel and Internals Project BWRVIP-05 Report (TAC No. M93925)."
The staff determined the conditional probability of failure for axial and circumferential shell welds in BWR vessels fabricated by Chicago Bridge and Iron (CB&I), Combustion Engineering, and Babcock and Wilcox. The staffs analysis identified a cold overpressure event that occurred in a foreign reactor as the limiting event for BWR RPVs and used it in the PFM calculations. The staff estimated that the probability for the occurrence of the limiting overpressurization transient was 1 x 10-3 per reactor year. For each vessel of a specific fabricator, Table 2.6-4 of the staffs July 28, 1998, SE identifies the conditional failure probabilities at 32 EFPY for the limiting plant-specific conditions (with the highest projected reference temperature).
Generic Letter 98-05 On November 10, 1998, the NRC issued Generic Letter (GL) 98-05, Boiling Water Reactor Licensees Use of the BWRVIP-05 Report to Request Relief from Augmented Examination Requirements on Reactor Pressure Vessel Circumferential Shell Welds, which stated that BWR licensees may request permanent (i.e., for the remaining term of operation under the existing, initial license) relief from the ISI requirements of 10 CFR 50.55a(g) for the volumetric examination of RPV circumferential shell welds upon demonstrating that:
(1) At the expiration of their license, the circumferential shell welds will continue to satisfy the limiting conditional failure probability for circumferential shell welds in the staffs July 28, 1998, SE 1, and (2) Licensees have implemented operator training and established procedures that limit the frequency of cold overpressure events to the amount specified in the staff's July 28, 1998, SE.
Licensees would still need to perform the required inspections of essentially 100 percent of all axial shell welds.
3.2 Licensee Evaluation ASME Code Requirement for which Relief is Requested The licensee requested relief from the following requirements of ASME Code,Section XI, 1989 Edition:
Subarticle IWB-2500, Examination and Pressure Test Requirements, Table IWB 2500-1, Examination Category B-A, Item No. B1.11 Subarticle IWB-2420, Successive Inspections Subarticle IWB-2430, Additional Examinations Component(s) for which Relief is Requested The requested relief from the Table IWB 2500-1 requirements applies to:
ISI Class 1, Code Category B-A, Pressure Retaining Welds in Reactor Vessel, Item B1.11, "Circumferential Shell Welds Licensees Proposed Alternative to the ASME Code In accordance with 10 CFR 50.55a(a)(3)(i), and consistent with information contained in NRC GL 98-05, the licensee considers the following alternate provisions for the subject shell weld examinations:
Inspection Scope The failure frequency for RPV circumferential shell welds is sufficiently low to justify their elimination from the ISI requirement of ASME Code Section Xl, Table IWB-2500-1, Examination Category B-A, Item No. B1.11. The ISI examination requirements of ASME Code Section Xl, Table IWB-2500-1, Examination Category B-A, Item No. B1.12, RPV Longitudinal Shell Welds, shall be performed, and shall include inspection of the circumferential shell welds only at the intersections of these welds with the axial shell welds, or approximately 2 to 3 percent of the RPV circumferential shell welds. Instead of using the ASME Code terminology, longitudinal shell welds, this SE has used axial shell welds throughout the evaluation. The procedures for these examinations shall be qualified such that flaws relevant to the RPV integrity can be reliably detected and sized, and the personnel implementing these procedures shall be qualified in the use of these procedures.
Successive Examination of Flaws For ASME Code Section Xl, Table IWB-2500-1, Examination Category B-A, Item No.
B1.11, RPV circumferential shell welds, referred to henceforth as only RPV circumferential shell welds, at intersections with axial shell welds, successive examinations per Subarticle IWB-2420 are not required for non-threatening flaws (original vessel material or fabrication flaws such as inclusions which exhibit negligible or no growth during the life of the vessel), provided that the following conditions are met:
1.
The flaw is characterized as subsurface in accordance with BWRVIP-05; 2.
The non-destructive examination technique and evaluation that detected and characterized the flaw as originating from material manufacture or vessel fabrication is documented in a flaw evaluation report; and 3.
The vessel containing the flaw is acceptable for continued service in accordance with Subarticle IWB-3600, Analytical Evaluation of Flaws, and the flaw is demonstrated acceptable for the intended service life of the vessel.
For RPV axial shell welds, all flaws shall be reinspected at successive intervals consistent with ASME Code and regulatory requirements.
Additional Examinations of Flaws For RPV circumferential shell welds at intersections with axial shell welds, additional requirements per Subarticle IWB-2430 are not required for flaws provided the following conditions are met:
1.
If the flaw is characterized as subsurface in accordance with BWRVIP-05, then no additional examinations are required; 2.
If the flaw is not characterized as subsurface in accordance with BWRVIP-05, then an engineering evaluation shall be performed, addressing the following as a minimum:
- a determination of the root cause of the flaw,
- an evaluation of any potential failure mechanisms,
- an evaluation of service conditions which could cause subsequent failure, and
- an evaluation per Subarticle IWB-3600 demonstrating that the vessel is acceptable for continued service; and 3.
If the flaw meets the criteria of Subarticle IWB-3600 for the intended service life of the vessel, then additional examinations may be limited to those welds subject to the same root cause conditions and failure mechanisms, up to the number of examinations required by IWB-2430(a). If the engineering evaluation determines that there are no additional welds subject to the same root cause conditions or no failure mechanism exists, then no additional examinations are required.
For RPV axial shell welds, additional examination for flaws shall be in accordance with Subarticle IWB-2430. All flaws in RPV axial shell welds shall require additional weld examinations consistent with ASME Code and regulatory requirements. Examinations of the RPV circumferential shell welds shall be performed if RPV axial shell welds reveal an active, mechanistic mode of degradation.
Licensees Bases for Alternative BWRVIP-05 provides the technical basis to justify relief from the examination requirements for RPV circumferential shell welds. The results of the NRC's evaluation of BWRVIP-05 are documented in the staff's July 28, 1998, SE. NRC GL 98-05 permits BWR licensees to request permanent (i.e., for the remaining term of operation under the existing, initial license) relief from the ISI requirements of 10 CFR 50.55a(g) for the volumetric examination of RPV circumferential shell welds. This relief can be granted by demonstrating that the following GL criteria are satisfied:
1.
At the expiration of their license, the circumferential shell welds will continue to satisfy the limiting conditional failure probability for circumferential shell welds in the staff's July 28, 1998, SE, and 2.
Licensees have implemented operator training and established procedures that limit the frequency of cold overpressure events to the amount specified in the staff's July 28, 1998, SE.
GL 98-05 also states that licensees will still need to perform the required inspections of essentially 100 percent of all axial shell welds.
Licensees Response to GL 98-05 Criterion 1 The submittal states that, The NRC evaluation of BWRVIP-05 utilized a probabilistic fracture mechanics (PFM) analysis to estimate the RPV weld failure probabilities. Three key assumptions of the PFM analysis are: (1) the neutron fluence used was the estimated end-of-life mean [sic] fluence; (2) the chemistry values are mean values based on vessel types; and (3) the potential for beyond-design-basis events is considered.
The licensees information regarding effects of irradiation on bounding DAEC RPV circumferential shell weld properties are presented in a table on page 7 of this SE. This table provides a comparison of the limiting RPV circumferential shell weld parameters for the DAEC unit to those found in Table 2.6-4 of the staff's July 28, 1998, SE for a CB&I vessel. The considerably lower copper content and slightly higher nickel content for the DAEC bounding weld result in a much lower chemistry factor for the DAEC unit than that from the NRC analysis. The 32 EFPY inside diameter fluence for the unit is also lower than the NRC estimated 32 EFPY fluence. As a result, the shift in reference temperature for the DAEC limiting circumferential shell weld is lower than the 32 EFPY shift from the NRC analysis. Although the unirradiated reference temperature for the unit is higher than the NRC limit, the overall result for the DAEC limiting circumferential shell weld is a lower calculated mean reference temperature than the NRC mean analysis value.
Therefore, the RPV weld embrittlement due to fluence is calculated to be less than the NRC's limiting case, and the unit's RPV circumferential shell weld failure probability is bounded by the conditional failure probability in the NRC's limiting plant-specific analysis through the projected end of license.
Licensees Response to GL 98-05 Criterion 2 The licensee has procedures in place for the DAEC unit that guide operators in controlling and monitoring reactor pressure during all phases of operation, including cold shutdown. Use of these procedures minimizes the potential for low temperature overpressurization (LTOP) events, and is reinforced through operator training. A primary system leakage test is performed prior to each restart after a refueling outage.
The DAEC test procedure contains additional requirements to aid in the prevention of an LTOP event, and requires a briefing prior to test commencement with all involved personnel. RPV temperature and pressure are required to be monitored and controlled to within the Technical Specifications (TS) pressure and temperature (P-T) limits during the entire test.
The staff's July 28, 1998, SE discussed the risk of cold overpressurization due to control rod drive (CRD) injection if a loss of station power occurs during the pressure test. A subsequent restart of the CRD pumps without restoring the reactor water cleanup (RWCU) pump would cold pressurize the RPV. The licensee implemented two special precautions in DAECs surveillance test procedure to preclude this from occurring: (1) allow the system to depressurize by opening the cleanup system drain header control valve in the event of a loss of offsite power, and (2) instruct operators to immediately trip the CRD pump if RWCU isolates. During cold shutdown, the CRD and RWCU systems are used to control RPV water level and pressure. The low flow rate of these pumps allows sufficient time for operators to react; therefore, they are unlikely to cause overpressurization during cold shutdown.
Effects of Irradiation on Bounding RPV Circumferential Shell Weld Properties Duane Arnold Energy Center Parameter Description DAEC Comparative Parameters At 32 EFPY for the Bounding Circumferential Shell Weld Wire Heat/Lot 07L669 Lot K004A27A USNRC Limiting Plant-Specific Analysis Parameters at 32 EFPY for CB&I RPV Circumferential Shell Welds Copper (Cu), wt%
0.03 0.10 Nickel (Ni), wt%
1.02 0.99 Chemistry Factor 41 134.9 End of Life Inside Diameter Fluence, x1019 n/cm2 0.355 0.51 Initial (unirradiated)
Reference Temperature RTNDT(U), EF
-50
-65 Increase in Reference Temperature RTNDT, EF 26.4 109.5 Mean (irradiated)
Reference Temperature RTNDT(u) + RTNDT, EF
-23.6 44.5 Other than the CRD system, the other high pressure coolant sources that could inadvertently initiate and result in an LTOP event are high pressure coolant injection (HPCI), reactor core isolation cooling (RCIC), and feedwater/condensate systems.
During cold shutdown there is no steam available to drive the turbine driven HPCI or RCIC pumps. Therefore, the HPCI and RCIC systems will not cause a cold overpressure event while the DAEC unit is in the cold shutdown operating mode. During a normal RPV fill sequence prior to pressure testing, the condensate system is used to fill the reactor. To prevent injection by an inadvertent start of a feedwater pump, injection of feedwater with vessel water level greater than 211 inches is controlled by a high water interlock. This interlock prevents starts of the feedwater pumps when RPV water level is equal to or greater than 211 inches. Defeating this interlock is procedurally and administratively controlled. Further, the likelihood of overpressurization at low temperature during startup is reduced by an administrative action requiring the RPV head vents not be closed until the RPV coolant temperature reaches 212 EF. Therefore, the feedwater/condensate system does not present a significant potential for overpressurization.
The standby liquid control (SLC) system is also a high pressure water source to the RPV. There are no automatic starts associated with this system. Operation of the SLC system requires an operator to manually start the system by a keylock switch.
Procedures have been developed for operation of the SLC system, and operators are trained on the system operation. Therefore, this system does not present a significant potential for overpressurization.
The low pressure coolant sources include the residual heat removal (RHR), low pressure coolant injection (LPCI), and the core spray (CS) systems. Based on observation and alarm of RPV water level, operators would detect and terminate an inadvertent injection of LPCI or CS. Further, a cold overpressure event is prevented by plant procedures which require the operator to place the RPV head vent valves in an open position when RPV coolant temperature is below 212 EF during cold shutdown with vessel head in tension. During refueling outages, a CS pump may be used for RPV and cavity fill; however, overpressurization is prevented by having RPV head removed under these conditions. When the condensate, CS, or RHR system is activated in the event of a loss of shutdown cooling, the safety relief valve is opened in accordance with Abnormal Operating Procedure to prevent an overpressure event.
In addition to the procedural barriers, licensed operators are provided specific training regarding the methods of controlling RPV water level and the TS requirements on P-T limits. Simulator sessions are conducted which include plant heat-up and cool-down.
3.3 Staff Evaluation As described previously, GL 98-05 provides two criteria that BWR licensees requesting relief from ISI requirements of 10 CFR 50.55a(g) for the volumetric examination of RPV circumferential shell welds must satisfy. These criteria are intended to demonstrate that the conditions at the applicants plant are bounded by those in the staff's July 28, 1998, SE. The licensee will still need to perform the required inspections of essentially 100 percent of all axial shell welds.
Circumferential Shell Weld Conditional Failure Probability The staffs July 28, 1998, SE evaluated the conditional failure probability of circumferential shell welds for the limiting plant-specific case of BWR RPVs manufactured by different vendors, including CB&I, using the highest mean irradiated RTNDT to determine the limiting case. Since the DAEC RPV was fabricated by CB&I, the licensee compared the mean irradiated RTNDT for DAEC to that for the limiting CB&I case described in Table 2.6-4 of the staffs July 28, 1998, SE. As indicated in the licensees evaluation, the mean RTNDT for the limiting DAEC circumferential shell weld is lower than that for the limiting CB&I case; therefore, the staff agrees with the licensees conclusion that the conditional failure probability for the DAEC RPV circumferential shell welds is bounded by the conditional failure probabilities in the staffs July 28, 1998, SE through the end of the current license period.
The licensee has used an NRC-approved methodology to estimate the 32 EFPY peak fluence value at the inside surface of the RPV. The resulting value of 0.355 x 1019 n/cm2 (E $ 1.0 MeV) is smaller than the value of 0.51 x 1019 n/cm2 (E $ 1.0 MeV) for the limiting CB&I case reported in the staffs July 28, 1998, SE. The licensees methodology of calculating the fluence estimates was provided by General Electric Nuclear Energy (GENE) topical report NEDC-32983 PA, Revision 1, Licensing Topical Report, General Electric Methodology for Reactor Pressure Vessel Fast Neutron Fluence Evaluations, which was approved by the NRC by letter dated September 14, 2001, and adheres to the guidance of Regulatory Guide (RG) 1.190, Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence.
Therefore, the fluence value of 0.355 x 1019 n/cm2 (E $ 1.0 MeV) at the inside surface of the DAEC RPV is acceptable.
The staffs July 28, 1998, SE provides a limiting conditional failure probability of 2.0 x 10-7 per reactor year for a limiting plant-specific mean RTNDT of 44.5 EF for CB&I-fabricated RPVs.
Comparing the information in the NRC reactor vessel integrity database with that in the submittal, the staff confirmed that the mean RTNDT of the limiting RPV circumferential shell weld at DAEC is projected to be -23.6 EF at the end of the current license. In this evaluation, the chemistry factor, RTNDT, and mean RTNDT were calculated consistent with the guidelines of RG 1.99, Revision 2, Radiation Embrittlement of Reactor Vessel Materials. The calculated value of mean RTNDT for the RPV circumferential shell welds at DAEC is significantly lower than that for the limiting plant-specific case for CB&I-fabricated RPVs, indicating that the conditional failure probability of the DAEC RPV circumferential shell welds is much less than 2.0 x 10-7 per reactor year.
Minimizing the Possibility of Low Temperature Overpressurization The licensee established operator procedures to control vessel pressure at low coolant temperatures. These procedures are reinforced through periodic operator training. Reactor conditions which may result in cold overpressurization are primary system leakage test, normal vessel fill sequence, and cold shutdown with tensioned head.
Primary System Leakage Test: Prior to the commencement of the test, the personnel involved are briefed on the procedure. During the test, potential overpressurization paths are avoided by administrative and/or hardware controls. RPV pressure and temperature are monitored and controlled to be within the TS P-T limits. The test procedures contain additional requirements to aid in the prevention of a low temperature overpressure event.
Normal Vessel Fill Sequence: The condensate system is used to fill the reactor. To minimize the potential of overpressurization, the feedwater pumps are prevented from starting during reactor testing when the reactor water level is high. Defeating this interlock is procedurally and administratively controlled. The RCIC and the HPCI systems are steam driven and cannot be operated during a cold shut down. Finally, the SLC system is also a high pressure delivery system, but it does not have an automatic initiation signal.
Cold Shutdown with Tensioned Head: With the head tensioned and the coolant temperature at or lower than 212 EF, plant operating procedures require that the RPV head vent valves be in the open position to prevent pressurization.
In summary, the probability of cold overpressurization is minimized by a combination of operator training, plant procedures, and hardware modifications. The staff finds the measures instituted by the licensee at DAEC to prevent cold overpressurization reasonable and appropriate.
4.0 CONCLUSION
The NRC staff has reviewed the licensees submittal and finds that the licensee has demonstrated that the appropriate criteria in GL 98-05 and the staffs July 28, 1998, SE have been satisfied regarding permanent relief (i.e., for the remaining portion of the current license period of 32 EFPY) from ISI requirements of ASME Code,Section XI, Table IWB-2500-1, Examination Category B-A, Item No. B1.11, for the volumetric examination of RPV circumferential shell welds. Hence, the staff concludes that the licensees relief request, pursuant to 10CFR 50.55a(a)(3)(i), is acceptable and is consistent with the information contained in NRC GL 98-05. The staff has also determined that the proposed alternative provides an acceptable level of quality and safety.
Principal Contributor: S. Sheng Date: January 6, 2005