ML112870050
| ML112870050 | |
| Person / Time | |
|---|---|
| Site: | Palisades  |
| Issue date: | 12/07/2011 |
| From: | Mahesh Chawla Plant Licensing Branch III |
| To: | Entergy Nuclear Operations |
| Mahesh Chawla, LPL3-1, 415-8371 | |
| References | |
| TAC ME5263 | |
| Download: ML112870050 (9) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 December 7, 2011 Vice President, Operations Entergy Nuclear Operations, Inc.
Palisades Nuclear Plant 27780 Blue Star Memorial Highway Covert, MI 49043-9530
SUBJECT:
UPDATED REACTOR PRESSURE VESSEL PRESSURIZED THERMAL SHOCK EVALUATION FOR PALISADES NUCLEAR PLANT (TAC NO. ME5263)
Dear Sir:
By letter dated December 20,2010 (Agencywide Documents Access and Management System (ADAMS) Accession Number ML110060692), Entergy Nuclear Operations, Inc., {the licensee}
submitted a request for Palisades Nuclear Plant to revise its reactor pressure vessel {RPV}
pressurized thermal shock {PTS} evaluation based on new information on surveillance data for the limiting RPV weld fabricated with weld wire heat No. W5214. This issue is addressed pursuant to Title 10 of the Code of Federal Regulations {10 CFR} 50.61, which provides the Nuclear Regulatory Commission's (NRC's) regulation regarding protection against PTS, and 10 CFR 54.21 (c){1 Hiii), which requires managing the aging effects for the period of extended operation.
The NRC staff has completed its review of the December 20, 2010 submittal and the licensee's response dated May 24, 2011 (ADAMS Accession No. ML11145A180), to the staffs request for additional information. The staff has concluded that the provided information related to the proposed PTS evaluation is in accordance with 10 CFR 50.61 and the established staff position of using surveillance data. Hence the staff concludes that the PTS screening criteria will not be reached until April 2017. Accordingly, the new chemistry factor for the limiting weld can be used in future pressure temperature limit applications. Enclosed is the staff's safety evaluation. If you have any questions regarding this matter, I may be reached at 301-415-8371.
Sincerely, Mahesh Chawla, Project Manager Plant Licensing Branch 111-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-255
Enclosure:
Safety Evaluation cc: Distribution via ListServ
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO FACILITY OPERATING LICENSE NO. DPR-20 UPDATED PRESSURIZED THERMAL SHOCK EVALUATION ENTERGY NUCLEAR OPERATIONS, INC PALISADES NUCLEAR PLANT DOCKET NO. 50-255
1.0 INTRODUCTION
By application dated December 20,2010 (Agencywide Document Access and Management System (ADAMS) Accession No. ML110060692), as supplemented by letter dated May 24, 2011 (ADAMS Accession No. ML11145A180), Entergy Nuclear Operations, Inc (ENO, the licensee), requested changes to the reactor pressure vessel (RPV) pressurized thermal shock (PTS) evaluation for the Palisades Nuclear Plant (PNP). The supplement dated May 24, 2011, in response to the staff's request for additional information (RAI) provided additional information that clarified the application.
The proposed changes would revise the PTS evaluation to incorporate the new information on surveillance data for the limiting RPV welds (axial welds) fabricated with weld wire heat number W5214 (Weld W5214). The PTS evaluation and its underlying methodology are reviewed by the Nuclear Regulatory Commission's (NRC's) Division of Engineering.
2.0 REGULATORY EVALUATION
The NRC has established requirements in Part 50 of Title 10 of the Code of Federal Regulations (10 CFR Part 50) to protect the integrity of the reactor coolant pressure boundary in nuclear power plants. The regulations in 10 CFR 50.61, "Fracture toughness requirements for protection against pressurized thermal shock events," provides the fracture toughness requirements protecting the RPVs of pressurized water reactors against the consequences of PTS. Licensees are required to perform an assessment of the RPV materials' projected reference temperature (RT PTS) values through the end of their operating license. The rule requires each licensee to calculate the end-of-license RT PTS value for each RPV beltline material. The RT PTS value for each beltline material is the sum of the unirradiated reference temperature nil ductility (RT NOT), a shift in the RT NOT value caused by neutron irradiation of the material (.6RT NOT), and a margin value to account for uncertainties (M).
The regulations in 10 CFR 50.61 also provides screening criteria against which the calculated values are to be evaluated. The screening criteria are 270 OF for plates, forging, and axial weld materials and 300 OF for circumferential weld materials. The regulations in 10 CFR 50.61 provide a discussion regarding the calculations of.6RTNOT and the M value. In 10 CFR 50.61,
.6RTNOT is the product of a chemistry factor and a fluence factor, where the fluence factor is
-2 dependent upon the neutron fluence at the clad-to-base metal interface and the chemistry factor is dependent upon information from either the surveillance material or from the tables in 10 CFR 50.61. If the RPV beltline material is not represented by surveillance material, its chemistry factor may be determined using the tables and the methodology documented in 10 CFR 50.61.
The chemistry factor determined from the tables in 10 CFR 50.61 depends upon the amount of copper (Cu) and nickel (Ni) in the material. If the RPV beltline material is represented by surveillance material, its chemistry factor may be determined from the surveillance data using the methodology documented in 10 CFR 50.61. The methods of determining RT PTS values in 10 CFR 50.61 are equivalent to the methods of determining RT NDT values in Regulatory Guide (RG) 1.99, Revision (Rev.) 2, "Radiation Embrittlement of Reactor Vessel Materials."
In accordance with 10 CFR 50.61, all license renewal applications (LRAs) have addressed their plants' PTS evaluations for 60 years of operation. For plants which exceed the PTS screening criteria during the period of extended operation, such as PNP, the PTS evaluation for the period of extended operation is addressed pursuant to 10 CFR 54.21(c)(1)(iii), which requires managing the aging effects for the period of extended operation. For these plants, the estimated date of reaching the PTS criteria was usually provided in the LRA and evaluated in the safety evaluation (SE) report for it.
3.0 TECHNICAL EVALUATION
3.1 Non-Limiting PNP RPV Beltline Materials (Attachment 2 to the December 20, 2010, submittal)
The revised PTS evaluation for the entire RPV beltline materials is contained in Attachment 2, "Revised Pressurized Thermal Shock Evaluation for the Palisades Reactor Pressure Vessel," to the licensee's December 20, 2010, submittal. This PTS evaluation is based on revised neutron fluence values, additional surveillance data from PNP and other plants' RPVs, revised aRTNDT values for all surveillance data using a specific hyperbolic tangent curve fitting program (CVGRAPH, Version 5.0.2), and revised chemistry data for certain RPV beltune materials based on the best-estimate values from the Combustion Engineering Owners Group (CEOG) NPSD 1039, Rev. 2 report (1997), "Best Estimate Copper and Nickel Values in CE [Combustion Engineering] Fabricated Reactor Vessel Welds." The NRC staff reviewed this revised PTS evaluation and found no further discussion is needed for non-limiting RPV beltUne materials because this part of the license's evaluation is completely in accordance with the 10 CFR 50.61 methodology and the use of best-estimate chemistry data is consistent with the established position regarding Cu and Ni values for RPV materials. Further, the credibility evaluation of the relevant surveillance data for non-limiting RPV beltline materials showed that the data are credible or not credible without ambiguity, making use (or not) of surveillance data in determining the chemistry factor for the relevant RPV material straightforward. Hence, the staff's evaluation foclJses on the limiting beltline material Weld W5214. This material was also the limiting RPV material in all previous PTS evaluations for PNP. The PTS evaluation for the limiting material is contained in Attachment 1, "Evaluation of Surveillance Data for Weld Heat No. W5214 for Application to Palisades PTS Analysis," to the licensee's December 20,2010, submittal.
- 3 3.2 The Limiting PNP RPV Beltline Weld W5214 (Attachment 1 to the December 20,2010, submittal) 3.2.1 Surveillance Data Completeness Similar to the case for non-limiting RPV materials, the PTS evaluation for the limiting RPV weld, Weld W5214, is also based on revised neutron f1uence values, additional surveillance data from PNP and other plants' RPVs, revised llRTNOT values for all surveillance data using a specific hyperbolic tangent curve fitting program, and revised chemistry data for certain RPV beltline materials based on the best-estimate values from the CE NPSD-1039, Rev. 2, report. However, the most significant information in the PTS evaluation for Weld W5214 is identification of 11 surveillance data for this weld: two from PNP; three from H. B. Robinson, Unit 2 (Robinson 2);
two from Indian Point, Unit 2 (lP-2); and four from Indian Point, Unit 3 (lP-3). This is a result of reviewing all surveillance capsule reports in the industry. Consideration of surveillance data for the same weld wire heat from other plants is acceptable to the staff if appropriate adjustments for the RPV operating temperature and surveillance specimen chemistry data are made, as illustrated in the examples provided in the 1998 NRCllndustry workshop slides on RPV integrity issues, "Generic Letter 92-01 and RPV Integrity Assessment," (ADAMS Accession No. ML110070570).
Regarding the completeness of the additional surveillance data from all sources, the licensee's May 24,2011, RAI response indicated that the licensee has reviewed other references, including the following: the NRC's Reactor Vessel Integrity Database; Electric Power Research Institute's Reactor Vessel Materials Database; the CEOG NPSD-1039, Rev. 2 report; and the CEOG NPSD-1119, Rev. 1, report (1998), "Updated Analysis for Combustion Engineering Fabricated Reactor Vessel Welds Best Estimate Copper and Nickel Content." Further, the RAI response stated that the licensee discussed with Westinghouse Electric Company and AREVA (formerly Babcock and Wilcox Company) to identify and confirm the surveillance capsules that contained Weld W5214 and weld wire heat number 27204 (Weld W27204) which have been removed and tested to date in the industry. Based on the above, the NRC staff determined that the licensee's evaluation for the limiting weld has considered all relevant surveillance data available in the industry and no data is likely to be missed in ENO's survey.
3.2.2 PNP Supplemental Surveillance Capsule Data The two supplemental surveillance data were obtained from PNP surveillance capsules SA-60-1 and SA-240-1. Both used 18 millimeter (m m) inserts which were reconstituted into full size Charpy specimens after irradiation for testing. Two issues are evaluated here: (1) whether PNP's reconstituted full sized Charpy specimens can produce meaningful Charpy curves and (2) whether the plant-specific surveillance data are credible.
Regarding whether the specimens in the PNP supplemental surveillance capsules were prepared in accordance with American Society of Testing and Materials (ASTM) Standard E 185-82, "Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels," the licensee's May 24, 2011, RAI response states that, "[c]onfirmation that equivalent results are obtained from reconstituted Charpy specimens as compared to full size Charpy specimens is documented in BAW-2184" [Verification of Reconstituted Charpy V-Notch Test Values]. "Extensive studies also have been conducted, including an ASTM round
-4 robin in which many United States and International laboratories participated..., that further support the guidance in ASTM E 1253" ["Standard Guide for Reconstitution of Irradiated Charpy Specimens"]. The ASTM round robin effort is documented in NUREG/CR-6777 (2002), "Results and Analysis of the ASTM Round Robin on Reconstitution." The staff reviewed both reports and agrees with the NUREG conclusion, "there is generally little effect of reconstitution on Charpy properties when the 14 mm insert is used for all reconstitution techniques for all energy levels...
[ASTM] E 1253 requires an insert size larger than 14 mm by the size of the HAZs caused by the reconstitution process. Therefore, it is concluded that the insert size requirement of [ASTM] E 1253 cannot be universally reduced." Since the reconstituted specimens in the PNP supplemental surveillance capsules were 18 mm, the fundamental requirement of NUREG/CR 6777 was met. Hence, the new PNP surveillance data are considered valid, and a credibility test may be performed to determine whether the surveillance data can be used in estimating the
~RTNOT value.
As for the plant-specific surveillance data credibility, 10 CFR 50.61, as well as RG 1.99, Rev. 2, specified five criteria to judge whether the plant-specific surveillance data are credible and whether the results from the plant-specific surveillance program must be integrated into the RTNOT estimate. Credibility Criterion 1 regarding identification of the limiting material is satisfied because the specimens were made of the limiting Weld W5124 material. Criterion 2, regarding small scatter in plotting the Charpy curve, is satisfied when the licensee used the computerized hyperbolic tangent curve fitting in estimating the Charpy curve. Criterion 5, regarding scatter of the correlation monitor (CM) surveillance data, is satisfied when the licensee demonstrated in Table 10 that the CM surveillance data fall within the scatter band of the data base for the material.
Criterion 3 sets limits on scatter of the plotted ~RTNOT values. The regulations in 10 CFR 50.61 require, "[w]here there are two or more sets of surveillance data 'from one reactor, the scatter of
~RTNOT values must be less than 28 OF for welds and 17 OF for base metal." The licensee plotted the one-reactor (Le., PNP) surveillance data in Figure 2 of Attachment 1 to the December 20,2010, submittal, which clearly shows that PNP's supplemental surveillance data fall within the one standard deviation (one sigma) scatter band of 28 OF for credible weld surveillance data. Therefore, the NRC staff determined that the two supplemental PNP surveillance data can be used, along with the 9 surveillance data from other plants, in estimating the ~RTNOT value for its limiting RPV weld.
Criterion 4 sets limits on the temperature difference between the vessel and the surveillance capsule. Table 8 of Attachment 1 to the December 20,2010, submittal showed PNP's effective full power days and RPV temperature as a function of operating cycle number. Using this information, the NRC staff has verified the time-weighted average temperatures for capsules SA-60-1 and SA-240-1 and the time-weighted RPV average temperature that was calculated by the applicant and presented in Table 8 of Attachment 1 to the December 20, 2010, submittal.
This exercise confirmed that the temperature differences between the RPV and the two PNP supplemental surveillance capsules are within the Criterion 4 limit.
It should be noted that a discussion of meeting the five criteria for surveillance capsules containing Weld W5214 for Robinson 2, IP-2, and IP-3 is not needed here. This is because, unlike the PNP case which represents new surveillance data, the cases for other plants represent existing surveillance data which have been reported in each plant's surveillance
- 5 capsule report and used in its license amendment requests such as the pressure temperature (P-T) limit and PTS evaluations.
3.2.3 Credibility of the 11 Surveillance Data for Weld W5214 from Four Plants Section 3.2.2 of this SE discusses how the PNP surveillance data meet Criterion 3 regarding the scatter of aRTNOT values for the surveillance data. This section discusses how the relevant multi-plant surveillance data meet Criterion 3. Although 10 CFR 50.61 mentioned use of surveillance data from other plants, specific guidance was not given there. Hence, the licensee referenced the NRC staffs guidance from the 1998 NRCllndustry workshop on RPV integrity issues for assessing the credibility of the entire 11 surveillance data. After adjustment for the chemistry data difference between the PNP limiting RPV beltline material and the surveillance data and for the temperature difference between the surveillance specimen irradiation and the RPV cladding/base metal interface, the licensee presented the calculated scatter for all surveillance data in Table 7 of Attachment 1 to the December 20,2010, submittal. This table shows that four of the 11 surveillance data exceed the one-sigma scatter band. Practically speaking, however, only two data are of concern because the other two are located very close to the one-sigma limit of 28 OF.
For surveillance data from a single RPV, Criterion 3 requires a one-sigma scatter of 28 OF for considering the surveillance welds of the same heat credible. Applying the same criterion to surveillance data from multiple RPVs is debatable because although adjustment has been made in the aRTNOT to account for differences in operating temperature and surveillance specimen irradiation temperature among RPVs, not all factors due to differences in host reactors have been identified and accounted for. Due to the large neutron f1uence range and the number of RPVs (i.e., 4) hosting the surveillance capsules, it is not a surprise to the NRC staff that the one-sigma curves do not bound all 11 surveillance data. This does not mean that the entire surveillance data should be discarded, but means that caution should be taken in using these surveillance data. This is consistent with what Criterion 3 implies. Criterion 3 states that,
"[e]ven if the range in the capsule fluences is large (two or more orders of magnitude), the scatter may not exceed twice those values [Le., 28 OF for welds and 17 OF for base metal]." This suggests that as long as the scatter of the surveillance data does not exceed two sigmas, the data could still be considered.
When surveillance data are used, RG 1.99, Rev. 2 allows use of half the one sigma for aRTNOT (i.e., Y:z ot.) in the margin calculation:
Margin = 2 (Oi2 + Ot.2)1/2, where 0i is the one sigma for the initial RT NOT Here, the licensee used full Ot. in the margin calculation for the RT PTS value of the limiting RPV weld to compensate for the unsatisfactory surveillance data scatter. This prudent approach has been used in past P-T limit or PTS applications where not all plant-specific surveillance data met the scatter limit. Considering this prudent approach, the NRC staff determined that the measured shifts for these surveillance data can be used in calculating the chemistry factor for Weld W5214 and the calculated PTS results summarized in Table 15 of Attachment 1 to the December 20, 2010, submittal is acceptable. Table 15 shows that the PTS screening criterion of 270 OF for the limiting weld will be reached when the neutron f1uence value reaches 1.685E+19 n/cm2 (E > 1 MeV). Since the NRC staff has accepted the licensee's calculated PTS
-6 results, the calendar date of April 2017 corresponding to this neutron fluence value as shown in Table 17 of Attachment 1 to the December 20,2010, submittal is also acceptable to the NRC staff.
Some secondary technical issues are discussed below in Section 3.2.4 of this SE. They provide additional support for the NRC staffs acceptance of the proposed PTS approach and results.
3.2.4 Other Reviewed Areas Updated Surveillance Data Using New Curve Fitting and Projected Capsule Neutron Fluence Values Regulatory Guide 1.190, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," (ADAMS Accession No. ML010890301) describes acceptable ways to calculate reactor vessel neutron fluence. RG 1.190 states that fluence calculations should adhere to NRC-approved methodology and provides acceptable qualification criteria. (ADAMS Accession No. ML110060695) to the licensee's December 20, 2010, submittal letter describes the fluence calculations performed in support of the present request.
The attachment indicates that the neutron transport calculations were carried out using the DORT (Discrete Ordinates Radial Transport) code, and performing a three dimensional flux synthesis. The DORT code uses nuclear data from the BUGLE-96 wide-group cross section library. The code also treats anisotropic scattering with a Pslegendre expansion, and models the vessel geometry using an S16 order of angular quadrature. RG 1.190 recommends that, at a minimum, fluence calculations be based on BUGLE-96 nuclear data, using P3 and S6 approximations. The methods used in the licensee calculations are consistent with the nuclear data suggested by RG 1.190, and exceed the minimum calculational rigor suggested by RG 1.190. On this basis, the NRC staff finds the fluence calculations acceptable. The licensee reported surveillance capsule dosimetry analysis results agree within the 20% uncertainty specified in RG 1.190, and on this basis the NRC staff finds that the f1uence calculations are acceptably qualified.
In accordance with industry practice, the licensee's PTS evaluation has reflected the updated capsule neutron fluence values and the revised Charpy curves for the 11 surveillance data based on hyperbolic tangent curve fitting in the relevant surveillance capsule reports.
Comparing with previously reported values, the NRC staff found that the majority of the refitted 11 surveillance data indicated a change of less than two degrees Fahrenheit in the measured
~RTNDT values, indicating a small effect due to use of the specific hyperbolic tangent curve fitting program and revised neutron f1uence. Two IP-2 data indicated a difference of six degrees Fahrenheit. However, by examining the fitted plots for the Charpy data sets, the NRC staff verified that the revised plots are reasonable and the revised shifts for these two surveillance data are valid. Based on the above, the NRC staff determined that the licensee's surveillance weld Charpy curves are of acceptable quality and the measured shifts for these surveillance data can be considered for calculation of the chemistry factor for the limiting RPV weld.
- 7 Time-Weighted Average Temperatures for Surveillance Capsules and the RPV The licensee's time-weighted average temperatures for surveillance capsules and the RPV are calculated based on simple averaging:
(Avg. Temp) capsule or RPV =HEFPY x temp)i I ~(EFPY)i' where "i" is the relevant cycle number For a surveillance capsule, the relevant cycles are the cycles between capsule insertion and withdrawal; for the RPV, all cycles are relevant. Simple averaging is adequate for this application, considering the small temperature differences among cycles. For example, except for the first two cycles, the operating temperature difference among cycles is within 3 of for PNP, 6 OF for IP-2, and 3 of for IP-3. The NRC staff verified the calculated time-weighted average temperatures for surveillance capsules and the RPV for PNP (Table 8 of Attachment 1 to the December 20,2010, submittal), IP-2 (Appendix H to Attachment 1 to the December 20, 2010, submittal), and IP-3 (Appendix H to Attachment 1 to the December 20,2010, submittal).
Although the report did not present similar information for each cycle for Robinson 2, the NRC staff still accepts the average capsule temperatures for Robinson 2 because Robinson 2 has confirmed these values to ENO and this type of simple averaging is unlikely to create errors.
4.0 CONCLUSION
Based on the NRC staffs review of the information provided in the licensee's December 20, 2010, and May 24,2011, submittals, the staff concludes that the updated PNP PTS evaluation is in accordance with 10 CFR 50.61 and the established staff position of using surveillance data.
Hence the NRC staff concludes that the PTS screening criteria will not be reached until April 2017. Accordingly, the new chemistry factor for the limiting weld can be used in future P-T limit applications.
Principal Contributors: S. Sheng B. Parks Date:
December 7, 2011
Decenber 7, 2011 Vice President, Operations Entergy Nuclear Operations, Inc.
Palisades Nuclear Plant 27780 Blue Star Memorial Highway Covert, MI 49043-9530
SUBJECT:
UPDATED REACTOR PRESSURE VESSEL PRESSURIZED THERMAL SHOCK EVALUATION FOR PALISADES NUCLEAR PLANT (TAC NO. ME5263)
Dear Sir:
By letter dated December 20, 2010 (Agencywide Documents Access and Management System (ADAMS) Accession Number ML110060692), Entergy Nuclear Operations, Inc., (the licensee) submitted a request for Palisades Nuclear Plant to revise its reactor pressure vessel (RPV) pressurized thermal shock (PTS) evaluation based on new information on surveillance data for the limiting RPV weld fabricated with weld wire heat No. W5214. This issue is addressed pursuant to Title 10 of the Code of Federal Regulations (10 CFR) 50.61, which provides the Nuclear Regulatory Commission's (NRC's) regulation regarding protection against PTS, and 10 CFR 54.21 (c)(1)(iii), which requires managing the aging effects for the period of extended operation.
The NRC staff has completed its review of the December 20,2010 submittal and the licensee's response dated May 24,2011 (ADAMS Accession No. ML11145A180), to the staffs request for additional information. The staff has concluded that the provided information related to the proposed PTS evaluation is in accordance with 10 CFR 50.61 and the established staff position of using surveillance data. Hence the staff concludes that the PTS screening criteria will not be reached until April 2017. Accordingly, the new chemistry factor for the limiting weld can be used in future pressure temperature limit applications. Enclosed is the staff's safety evaluation. If you have any questions regarding this matter, I may be reached at 301-415-8371.
Sincerely, IRA!
Mahesh Chawla, Project Manager Plant licenSing Branch 111-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-255
Enclosure:
Safety Evaluation cc: Distribution via ListServ DISTRIBUTION:
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