ML083530980
| ML083530980 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 12/18/2008 |
| From: | Constellation Energy Group, Nuclear Generation Group |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| TAC MD9773 | |
| Download: ML083530980 (9) | |
Text
ATTACHMENT (1)
RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION: CCNPP, UNIT NO. 2 PROPOSED RELIEF REQUEST ISI-020 Calvert Cliffs Nuclear Power Plant, Inc.
December 18, 2008
ATTACHMENT (1)
RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION: CCNPP, UNIT NO. 2 PROPOSED RELIEF REQUEST ISI-020 The responses provided below are necessary for the Nuclear Regulatory Commission (NRC) staff to complete their review of ISI-020 (Reference 1).
Question 1:
The Nuclear Regulatory Commission (NRC) Staff seeks clarification of information regarding observed indications from recent ISI in Table 2 of Proposed Alternative ISI-020, Attachment (1). Provide a summary of each inspection report, including the initial inspection before being put into service, and a brief description of the procedure used at each inspection. Clearly state when the indications were found and if their discovery can be attributed to improved ultrasonic inspection procedures or if the indications were the result of aging mechanisms.
CCNPP Response:
A summary of each inspection report; 1999-2nd 10-year Inservice Inspection (ISI), 1986-1st 10-year ISI, and 1976-Pre-Service Inspection (PSI), is provided in Table A below. A brief description of the essential variables of the ultrasonic procedures used for each 10-year reactor pressure vessel inspection is provided in Table B below.
As shown in Table A, comparisons of recent results to previous examinations show that of the nine recordable indications found in the Calvert Cliffs Nuclear Power Plant Unit 2 vessel using qualified and demonstrated Performance Demonstration Initiative (PDI) automated ultrasonic testing (UT) methods in 1999; none were recordable during the 1976 PSI and one was listed as a spot indication in the 1987 1st 10-year ISI examination. This spot indication had length but no measurable through-wall extent by the examination technique and procedure used at that time. All nine indications were characterized as sub-surface (not ID-connected) and as such, are not related to aging mechanisms. The discovery of the eight indications not previously reported is attributable to two major factors:
- 1. Improvements in examination equipment - Improvements to automated ultrasonic testing equipment, especially in the area of data acquisition and search unit design, have increased the sensitivity and resolution capabilities of ultrasonic examination systems used for reactor pressure vessels.
- 2. Improvements in examination procedures - Unlike the PSI and 1st 10-year ISI examinations, the American Society of Mechanical Engineers (ASME)Section XI, Appendix VIII PDI UT techniques used for the 1999 examinations have detection (recording) criteria which are based on signal-to-noise ratio rather than just exceeding a pre-defined signal amplitude threshold from an artificial (drill hole) reflector. As shown in Table B, the amplitude-based method used during the PSI used a 3/8 diameter side-drilled hole reflector to establish the reporting threshold over the full vessel wall thickness. Given that such a large reflector was used to set the recording threshold it is not surprising that no indications were reported. The vessel UT procedure was revised for the 1st 10-year ISI to include the use of refracted longitudinal waves and much smaller 1/16 diameter side-drilled holes to set the sensitivity for recording under clad and near surface indications from the vessel ID to the first quarter of the wall thickness; while leaving the remaining three-quarters of the vessel wall recording criteria the same as used for the PSI. This technique change led to the reporting of the largest percent through-wall indication (3.36%) reported during the 1999 10-year ISI examination (Indication #3 on Table A - Weld 2-203A at 28.8°/217.6). This was characterized in 1987 as a spot or no measurable through-wall extent indication. This finding is consistent with the through-wall sizing techniques available for use at that time, especially for sizing non-planar reflectors such as slag.
1
ATTACHMENT (1)
RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION: CCNPP, UNIT NO. 2 PROPOSED RELIEF REQUEST ISI-020 Table A- Calvert Cliffs Unit 2 Automated Ultrasonic Reactor Pressure Vessel Exams Indication Summary and Comparison Flaw Location 1999 - 2nd ISI 1987 - 1st ISI (1) 1976 Pre-Service (PSI)
Weld Number Ind. # Azimuthal /
a/t % length (in.) a/t % length (in.) a/t % length (in.)
Elevation 405A 1 77.8° / 86.2" 2.19 1.44 Not Recordable Not Recordable Inlet Nozzle-to-Shell @ 60° 1-203B 2 210.5° / 111.5" 0.56 1.12 Not Recordable Not Recordable Upper Shell Long @ 210° 2-203A 3 28.8° / 217.6" 3.36 0.96 Spot (2) 1.00 Not Recordable Middle Shell Long @ 30° 4 29.5° / 224.7" 1.39 0.8 Not Recordable Not Recordable 2-203C 5 271.8° / 173.9" 1.1 0.8 Not Recordable Not Recordable Middle Shell Long @ 270° 3-203A 6 88.1° / 261.3" 1.04 0.79 Not Recordable Not Recordable Lower Shell Long @ 90° 8-203 7 19.6° / 132.5" 1.49 1.23 Not Recordable Not Recordable Upper-to-Middle Shell Circ 10-203 8 9.8° / 328.7" 1.94 0.94 Not Recordable Not Recordable Lower Shell-to-Lower Head 9 25.1° / 328.8" 2.63 0.79 Not Recordable Not Recordable Circ NOTES:
(1)
The 1987 Examination was conducted in accordance with earlier Editions of the ASME Section XI Code which required examination of only 5% of the length of some circumferential welds and only 10% of the length of some longitudinal welds. Therefore, some weld areas that contained indications in 1999 may not have been examined in 1987.
(2)
Spot reflectors have no measurable through-wall dimension.
2
ATTACHMENT (1)
RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION: CCNPP, UNIT NO. 2 PROPOSED RELIEF REQUEST ISI-020 Table B - Calvert Cliffs Unit 2 Automated Ultrasonic Reactor Pressure Vessel Exams Ultrasonic Technique Summary and Comparison Exam Examination ASME Sec. Ultrasonic Ultrasonic Reference Level Reason / Equipment Recording Criteria XI Code Year Procedure Transducer Sensitivity Setup Year Type All raw data recorded, Under Clad and Near- Under Clad and Near- Detected Flaw surface: Refracted surface: Set to 80% Screen Indication must be Longitudinal Wave Amplitude from 1/8" Side- seen on two data Tandem (SLIC 40), drilled Hole that Gives channels and have EDAS-II 1992 w/
2.0 MHz Maximum Response signal-to-noise ratio of 2nd 10- Enhanced Addenda SwRi-PDI-at least 2:1 Year ISI / Data through 1993 - AUT-1 Rev. 3 All raw data recorded, 1999 Acquisition Appendix VIII (Detection) 1.5" to OD: Peak Response Detected Flaw System (PDI) 1.5" to OD: 45/55 Set to 80% Screen Indication must be Degree Shear Wave Amplitude from 1/4" Side- seen on two data Dual Element, drilled Hole that Gives channels and have 1.5 MHz Maximum Response signal-to-noise ratio of at least 2:1 Under Clad and Near- Under Clad and Near-surface: Refracted surface: Peak Response Set Longitudinal Wave to 40-80% Screen 20% DAC Dual Element (50/70 Amplitude from 1/16" Side-Standard 1974 w/ Wide Body), Drilled Hole that Gives 1st 10-Year Automated Addenda SwRI-NDT- 2.25 MHz Maximum Response ISI / 1987 UT with PaR through 700-11 Rev. 1 50% of 2%T Square tool Summer 1975 1/4T to OD: 45 and 1/4T to OD: Peak Response Notch Response for 60 Degree Shear Wave Set to 40-80% Screen planar surface Single Elements, Amplitude from 3/8" Side- reflectors and 20%
2.25 MHz drilled Hole at 1/4T DAC for all other Indications Standard Full Thickness: Peak 1971 w/ 45 and 60 Degree Automated SwRI-NDT- Response Set to 75% Screen PSI / 1976 Winter 1971 Shear Wave Single 50% DAC UT with PaR 700-1 Rev. 10 Amplitude from 3/8" Side-Addenda Elements, 2.25 MHz tool drilled Hole at 1/4T 3
ATTACHMENT (1)
RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION: CCNPP, UNIT NO. 2 PROPOSED RELIEF REQUEST ISI-020 Question 2:
Identify source documents for determining the manganese content for the welds and plates in the reactor pressure vessel and describe how you averaged them when more than one data sources are available.
CCNPP Response:
The manganese content for the welds and plates used in the calculation of through-wall cracking frequency was assumed based on the "conservative estimates for chemical element weight percentages" in Table 4 of the proposed Pressurized Thermal Shock (PTS) Rule, 10 CFR 50.61a. Table 4 of the proposed PTS Rule identifies values of 1.45 for plates and 1.63 for welds.
Note: The Calvert Cliffs reactor pressure vessel (RPV) heads were recently replaced. The Unit 2 RPV head was replaced in 2007. Therefore, to clarify Table 2 of proposed Alternative ISI-020 (Reference 1),
we are adding a note (Note 1) as shown below. This Table replaces Table 2 in Reference 1.
Table 2. Additional Information Pertaining to RPV Inspection Inspection methodology: ASME Section XI and Regulatory Guide 1.150 (Reference 7) inspections were Performance Demonstration Initiative (PDI) qualified per Reference 8 and were therefore performed in accordance with the requirements of ASME Section XI Appendix VIII.
Number of past inspections: (1) All welds have been inspected at least twice with the exception of 5 lower head Meridional welds which have been inspected once.
Number of indications found: A total of 9 indications were detected in the most recent ISI. All 9 indications are acceptable in accordance with IWB-3500 of Section XI of the ASME Code. Four of these indications are located in the reactor vessel beltline region. Three of these indications meet the Allowable Number of Flaws requirements for the proposed voluntary PTS Rule (10 CFR 50.61a) in SECY-07-0104 (Reference 9). One indication, in plate material with a through-wall extent of 0.60, does not meet the requirements in SECY-07-0104.
Additional information regarding this flaw is provided in Table 3 below.
Proposed inspection schedule The third ISI is currently scheduled for 2009. The third ISI is for balance of plant life: proposed to be performed in 2019. The fourth ISI is proposed to be performed in 2039.
NOTE:
(1)
Note that the inspections were performed on the original RPV head. The Unit 2 RPV head was replaced in 2007. Therefore, the replacement head-to-flange weld has not received a prior ISI examination.
4
ATTACHMENT (1)
RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION: CCNPP, UNIT NO. 2 PROPOSED RELIEF REQUEST ISI-020 Question 3:
The application states that the applicable American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) is the 1989 Edition, No Addenda. Considering the unique nature of the proposed alternative, which was intended to apply to the end of the current license, the NRC staff requests you revise the proposed alternative according to the following:
a) The Edition and Addenda of the ASME Code should be that which you would use for the fourth interval for other ASME Code applications.
b) The duration of the proposed alternative should end in 2036 when the current license expires; therefore, there will only be one additional scheduled ISI in 2019 covered by this request.
CCNPP Response:
a) The Edition referenced in ISI-020 is the 1998 Edition of ASME XI which is the Edition used for the Third Inservice Inspection 10-Year Interval. However, the 2004 Edition, no Addenda will be used for the Fourth Inservice Inspection Interval and Calvert Cliffs concurs with referencing the 2004 Edition, no Addenda.
b) Calvert Cliffs concurs with the proposed duration of this alternative (end of the current license which occurs in 2036) and acknowledges that only one additional ISI (scheduled for 2019) would be covered by this request.
Question 4:
The staff notes the following:
a) The licensees request, based on Topical Report (TR) WCAP-16168,Risk-Informed Extension of the Reactor Vessel In-Service Inspection Interval,, demonstrated that the TWCF95-TOTAL are orders of magnitude less than that for the bounding pilot plant vessel.
b) The calculation of TWCF95-TOTAL used the shift in the Charpy transition temperature produced by irradiation defined at the 30 ft-lb energy level, T30, calculated according to NUREG-1874, Recommended Screening Limits for Pressurized Thermal Shock (PTS).
c) Before the T30 calculation in NUREG-1874 can be used for risk-informed decisions, the licensee should verify that the NUREG-1874 values of T30 for the limiting baseline materials are comparable with the results from a plant-specific or integrated surveillance program if the surveillance has been deemed consistent. The criteria set forth in proposed 10 CFR 50.61a, published on August 11, 2008 (73 FR 46557), paragraphs (f)(6)(i) through (f)(6)(iv) are to be implemented when three or more surveillance data points at different neutron fluences exist for the limiting beltline material.
Demonstrate, if applicable, using the criteria set forth in proposed 10 CFR 50.61a (f)(6)(i) through (f)(6)(iv) that the embrittlement model used in your application to calculate T30 is applicable to the Calvert Cliffs Nuclear Power Plant, Unit No. 2.
CCNPP Response:
The proposed optional PTS Rule, 10 CFR 50.61a, requires the evaluation of plant specific information that could affect the calculation of the irradiation induced shift in reference temperature, T30, for the beltline materials. In order to make this determination, the proposed PTS Rule provides requirements for 5
ATTACHMENT (1)
RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION: CCNPP, UNIT NO. 2 PROPOSED RELIEF REQUEST ISI-020 evaluation of surveillance capsule data. These requirements are specified in paragraphs (f)(6)(i) through (f)(6)(iv) of the proposed PTS Rule. In summary the requirements consist of a Mean Deviation Test, a Slope Deviation Test, and an Outlier Deviation Test. The equations for performing these tests are included as Equations 8 through 12 and the acceptance criteria for these tests are provided in Tables 5, 6, and 7 of the proposed PTS Rule, 10 CFR 50.61a. These tests must be performed when three or more measurements of surveillance data are available for any of the reactor vessel beltline materials. The through-wall cracking frequencies that were calculated for the Calvert Cliffs Unit 2 reactor vessel inservice inspection interval relief request (ISI-020) were developed using the equations for irradiation induced shift from the technical basis (NUREG-1874) for the proposed PTS Rule, 10 CFR 50.61a. To demonstrate applicability of the NUREG-1874 irradiation induced shift equations, results from surveillance tests of Calvert Cliffs Unit 2 material were checked against the acceptance criteria of the proposed PTS Rule, 10 CFR 50.61a.
The Calvert Cliffs Unit 2 surveillance program includes the weld metal for the circumferential weld 9-203 (weld wire heat 10137) and plate D-8907-2 (heat number C-5386-1). Only two surveillance capsules have been withdrawn and analyzed for the Calvert Cliffs Unit 2 reactor vessel. However, the Calvert Cliffs Unit 2 lower shell weld seams, 3-203-A, B, and C are fabricated from weld wire heat 33A277 which is the weld wire heat for the Calvert Cliffs Unit 1 circumferential weld, 9-203. This weld wire heat is included in the Calvert Cliffs Unit 1 surveillance program and also the Farley Unit 1 reactor vessel surveillance program. There are two points of data from the Calvert Cliffs Unit 1 program and six points of data from the Farley Unit 1 program. Therefore, the evaluation of surveillance data required by the proposed PTS Rule has been performed for this weld material.
The inputs and results of the surveillance data evaluations are provided in Table C. The data for Calvert Cliffs Unit 1 was obtained from Reference 2 while the data for Farley Unit 1 was obtained from Reference 3. As can be seen in Table C, the surveillance results satisfy the criteria in the proposed PTS Rule for all three tests. Therefore, the use of the equations contained in NUREG-1874 for calculation of T30 is acceptable for Calvert Cliffs Unit 2.
6
ATTACHMENT (1)
RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION: CCNPP, UNIT NO. 2 PROPOSED RELIEF REQUEST ISI-020 Table C - 10 CFR 50.61a Surveillance Data Evaluation for Weld Metal 33A277 (Welds 3-203A, B, C) Calvert Cliffs Unit No. 2 Capsule 263° 97° Y U X W V Z Calvert Calvert Plant Cliffs Cliffs Farley Farley Farley Farley Farley Farley Unit 1 Unit 1 Unit 1 Unit 1 Unit 1 Unit 1 Unit 1 Unit 1 Input Data Copper (Weight %) 0.240 0.240 0.140 0.140 0.140 0.140 0.140 0.140 Phosphorous (Weight %) 0.014 0.014 0.016 0.016 0.016 0.016 0.016 0.016 Nickel (Weight %) 0.180 0.180 0.190 0.190 0.190 0.190 0.190 0.190 Manganese (Weight %) 1.050 1.050 1.060 1.060 1.060 1.060 1.060 1.060 Fluence (x1019 n/cm2, E > 1.0MeV) 0.620 2.640 0.612 1.730 3.060 4.750 7.140 8.470 EFPY 2.98 11.07 1.15 3.08 6.11 12.43 20.16 24.26 Time Averaged Coolant 548.00 548.00 544.00 540.25 540.86 541.75 541.72 541.43 Temperature (°F)
Measured T30 Transition 59.00 93.00 66.90 75.10 87.40 98.30 117.50 113.50 Temperature (°F)
Calculated Values Predicted T30 Transition 91.35 119.20 65.09 89.18 104.39 118.19 135.38 144.36 Temperature (°F)
Residual (r) -32.35 -26.20 1.81 -14.08 -16.99 -19.89 -17.88 -30.86 Mean Deviation Test Slope Deviation Test Outlier Deviation Test Mean Deviation -19.56 T-Statistic -0.89 Allowable Actual Maximum Mean Residual 21.75 Critical T-Statistic 3.14 Largest r* 3.02 -1.23 Pass/Fail? Pass Pass/Fail? Pass Second largest r* 2.05 -1.17 Pass/Fail? Pass 7
ATTACHMENT (1)
RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION: CCNPP, UNIT NO. 2 PROPOSED RELIEF REQUEST ISI-020
REFERENCES:
- 1. Letter from Mr. M. D. Flaherty (CCNPP) to Document Control Desk (NRC), dated October 1, 2008, Revised Request to Extend the Inservice Inspection Interval for Reactor Vessel Weld Examinations -
Relief Requests (ISI-020 and ISI-021)
- 2. BAW-2160, Analysis of Capsule 97° Baltimore Gas & Electric Company Calvert Cliffs Nuclear Plant No. 1, June 1993
- 3. WCAP-16964-NP, Analysis of Capsule Z from the Southern Nuclear Operating Company Joseph M. Farley Unit 1 Reactor Vessel Radiation Surveillance Program, October 2008 8