ML20151Z488
| ML20151Z488 | |
| Person / Time | |
|---|---|
| Site: | Limerick  |
| Issue date: | 06/30/1998 |
| From: | Caine T, Carey R, Tilly L GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20151Z456 | List: |
| References | |
| GE-NE-B1100786, GE-NE-B1100786-02, GE-NE-B1100786-2, NUDOCS 9809210353 | |
| Download: ML20151Z488 (44) | |
Text
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ECR LG 98-01857 Page 30 GENuclear Energy Technical Services Business General Electric Company 95_yg_g3399yg5_9y 175 Curtner Avenue, San Jose, CA 95125 June 1998 Surveillance Specimen Program Evaluation for Limerick Generating Station Unit 2 Prepared by:
@4'I'A MTillMenior Engineer Stmetural Mechanics and Materials i
i Verified by:
Mr R.G. Carey, Engineer Stmetural Mechanics and Materials b
Reviewed by:
T. A. Caine, Manager Structural Mechanics and Matexials Approved by: O L
B. J. Branlund, Project Manager Structural Mechanics and Materials l
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9809210353 980914
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i PDR ADOCK 05000353 P
ECR LG 98-01857 GE NuclearEnergy GE-NE-B1900766-02 IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT Please Read Carefully The only undertakings of General Electric Company respecting information in this document are contamed in the contract between PECo Energy Company (PECo) and General Electric Company, and nothing in this document shall be construed as changing the purchase order. The use of this information by anyone other than PECo, or for any purpose other than that for which it is intended, is not authorized; and with respect to any unauthorized use, General Electric Company makes no representation or warranty, and assumes no liability as to the completeness, accuracy, or usefulness of the information contained in this document, or that its use may not infringe privately owned rights.
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ECR LG 98-01857 GE Nucl:ar Energy GE-NE-B1100786-02 Table of Contents f.att
- 1. ABSTRACT 1
- 2. INTRODUCTION 3
- 3. COMPARISON WITH OTHER SURVEILLANCE DATA 9
- 4. PRESSURE-TEMPERATURE (P-T) CURVES 15
- 5. SUPPLEMENTAL SURVEILLANCE PROGRAM 18
- 6. REVISED SURVEILLANCE SCHEDULE 21
- 7. CONCLUSIONS 25
- 8. REFERENCES 28 APPENDIX A 29 APPENDIX B 35 i
i
ECR LG 98-01GS7 GE Nuclear Energy GE-NE-B1100786-02 Table of Figures g
Figure 2-1: Measured Shift vs. Predicted Shift For Base Metal 7
Figure 2-2: Measured Shift vs. Predicted Shift For Weld Metal 8
Figure 3-1: Measured Shift vs. Predicted Shift For Base Metal 13 l
Figure 3-2: Measured Shift vs. Predicted Shift For Weld Metal 14 Figure 4-1: Comparison of Kr. And Kie 17 Figure 6-1: Km s. EFPY For Limerick 2 Plate Material 23 v
4 Figure 6-2: Predicted Shift vs. EFPY, Limerick 2 Surveillance Capsule Plate 24 Figure A-1: ART vs. EFPY 34 Table of Tables
.P_ age Table 3-1: BWR Surveillance Program Results For Base Metal 11 Table 3-2: BWR Surveillance Program Results For Weld Metal 12 Table A-1: Limerick 2 RPV Material Data 33 iv
ECR LG 98-01857 GE Nuclear Energy M-M-m m n i
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ACKNOWLEDGMENT l
The author would like to thank Sam Ranganath and Brian Frew for valuable technical input to this document.
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- 1. ABSTRACT j
Limerick Generating Station, Unit 2 (Limerick 2, LGS 2) has maintained vessel surveillance programs to meet the intent of 10CFR50, Appendix H [1]. The current surveillance program schedule requires that the first surveillance capsule be removed at eight (8) Effective Full Power Years (EFPY) for LGS 2.
1 The original schedule was developed in accordance with the intent of 10CFR50, Appendix H. This schedule did not account for LGS 2 specific conditions:
Good plate and weld chemistry (low copper content from 0.01% - 0.lfM);
e Low RPV 1/4T 32 EFPY beltline fluence ( 5 x 10" n/cm fluence)[2];
2 e
Resulting low predicted shift in the reference nil-ductility temperature (RTer),
e
<60*F at 32 EFPY.
If the current schedule is used, the measured data for the plate material may not be useful, as the expected shift in RTer (ARTer) is low. In addition, the data normally provided by er.rly testing can be replaced by other sources. Therefore, the surveillance program's withdrawal schedule should be extended.
The extended schedule can bejustified because:
Actual BWR data shows predicted ARTer + margin values based on Regulatory Guide 1.99 Revision 2 (Rev 2) [3] to bound the measured ARTer values; There is inherent conservatism present in the pressure-temperature (P-T) curves for BWRs; The derived fracture toughness values are lower bound values and are based on crack arrest (Ku) rather than the higher crack initiation (Ku) toughness; l
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ECR LG 98-01857 GE Nuclear Energy GE-NE-B1100786-02 Data from other plants can be used to predict the behavior of the material early in plant life.
Based on the esaluation presented in this report, the recommended withdrawal schedule for the first surveillance capsule for Limerick 2 is 15 EFPY. The schedule for the second capsule (currently 20 EFPY) is changed to 30 EFPY at this time, but may be re-evaluated upon removal and evaluation of the first capsule. This new schedule meets the intent of ASTM E185-82 [2], as the first capsule would be removed with the fluence being less 1
than 5 x 10" n/cm and the value of ARTm would be less than 50 F, and the second 2
capsule would be removed before the accumulated neutron fluence of the capsule corresponds to the approximate end of life (EOL) fluence at the reactor inner wall location.
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r ECR LG 98-01857 GE Nuclear Energy GE-NE-B1100786-02
- 2. INTRODUCTION Vessel fracture toughness is a major consideration for nuclear vessels; irradiation is known to decrease the fracture toughness of vessel materials. Therefore, measurement of the long term effects of vessel irradiation is a key component of surveillance programs. PECO Energy Company (PECo) maintains a vessel surveillance program at LGS 2 meeting the intent of 10CFR50, Appendix H to monitor for changes in fracture toughness of vessel beltline materials as required by the NRC.
The Limerick 2 surveillance program meets the intent of 10CFR50, Appendix H and ASTM E185-73 (for design) for the following reasons:
The selected base and weld metals are representative of the vessel beltline materials; The materials have a similar fabrication history to the vessel; The number, type, and design of specimens are equivalent to ASTM E185-73.
The surveillance program implemented at Limerick 2 consists of three specimen holders installed in the reactor during vessel constmetion. The number of holders was determined per ASTM E185-73. Limerick 2 was defined as a case 'A' plant, since the Limerick 2 vessel had an RTm shift less than 100 F and the peak fluence was less than 5 x 10 'n/cm (at 1/4T) over the design lifetime of the plant.
2 The three specimen holde's were designed, built, and analyzed to ASME Section III, r
1968 Edition, with Addenda through Summer 1969. The selection of holder location was established to duplicate as closely as possible the temperature history, neutron flux 3
i ECR LG 98-01057 GE Nuclear Energy GE-NE-B1100786-02 spectrum, and maximum accumulated RPV beltline fluence, considering:
interference / accessibility with other reactor hardware (e.g., jet pumps);
peak fluence as a function of height; e
peak fluence as a function of radial position.
e 1
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~ Using these criteria, the capsules were located at the vessel inner diameter at core mid-height at the 30*,120 and 300 vessel azimuths (available areas considering jet pumps).
To provide baseline information, archive material was made available for additional testing.
At the time the withdrawal schedule of 8 EFPY was determined, ASTM E185-73
~
1 recommended the first and 'second capsules to be removed when the capsule fluence reaches 100% of the wall fluence at 32 EFPY. Current withdrawal schedule requirements per 10CFR50, Appendix H and ASTM E185-82, require that the first specimen holder be removed at 6 EFPY and the second be removed at 15 EFPY. ' All testing and reporting (regardless of withdrawal schedule) is to be performed in accordance with ASTM E185-82.
This capsule withdrawal schedule was recommended for two reasons:
- 1. Data would be provided for future - pressure-temperature (P-T) curve calculations. The data would be used to remove conservatism present in the (P-T) calculations.
The P-T curves would be recalculated after the first capsule had been removed, using the capsule flux wire measurements instead of the conservative calculated fluence.
- 2. The data obtained from the first capsule would be used to identify any anomalous conditions, i.e. a greater than expected shift in RTer.
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ECR LG 98-01657 GE Nucl:ar Energy GE-NE-B1100786-02 However, withdrawal at 8 EFPY of the LGS 2 capsule is not essential for continued safe operation for the following four reasons:
i
- 1. The LGS 2 fluence [4] used for shift predictions in accordance with Rev 2 is i
based upon a conservative calculation, and is exoected to bound the actual
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fluence.
- 2. Predicted shifts bound the measured results based on review of predicted RTer shifts and measured RTer shifts from other BWR surveillance capsules.
Figure 2-1 is a plot of actual shift measurements versus predicted shifts (calculated per Rev 2) for base material. This figure shows that the predicted shift plus margin conservatively bounds the actual shifts measured from BWR surveillance specimen data. The same plot for weld material (Figure 2-2) again shows the predicted shift plus margin term bounds the measured shift.
- 3. Based on actual ART calculations performed in accordance with Rev 2 (see Appendix A), the shift (oRTer + margin) for the Limerick 2 surveillance plate is calculated to be 68 F at 32 EFPY. If the first capsule is removed at 8 EFPY, the actual shift (predicted to be 19 F) may not be large enough to be differentiated from the data scatter, since the predicted fluence on the capsule at 8 EFPY (3.9 x 10" n/cm ) is low, and the chemistry of the Limerick 2 vessel 2
plate materia', good (0.11% - 0.15% copper). Thus, the data obtained may not be useful for predicting the material behavior, as it may be indistinguishable from the unirradiated data.
4.
Supplemental Surveillance Program (SSP) specimens will provide early test data for a plate similar to the Limerick 2 surveillance plate; the plate is the material of concern, as the vessel plate material is limiting throughout plant life.
This program supplements the LGS 2 surveillance program by providing timely detection of anomalous RTer shifts, should any occur. The fluences on the 5
ECR LG 98-01857 GE Nucisir Enstgy Paos40 GE-NE-B1100786-02 SSP capsules are comparable to the fluence for the LGS 2 vessel wall in the time frame ofinterest.
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This report shows that the surveillance capsule testing schedule for LGS 2 should be extended for the following reasons:
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The fluence experienced by the LGS 2 vessel wallis low; l
The LGS 2 vessel plate and weld materialin the beltline region has good alloy
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chemistry (i.e., low copper in the range of 0.01% - 0.15%);
The actual shift in the LGS 2 plate material may not be distinguishable from the data scatter with early testing.
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L The justification for extending the schedule is based on the following reasons-i Predicted shifts bound the actual BWR industry surveillance results; l
l The P-T curve calculations are inherently conservative; e
The supplemental surveillance program will supplement the LGS 2 surveillance i
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1 program by providing for the timely detection of anomalous RTum shifts.
Extension of the surveillance program schedule will ensure that useful data is obtained and continued safe operation of LGS 2 is ensured by using the SSP data and maintaining the LGS 2 P-T curves in accordance with Rev 2.
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ECR LG 98-01857 GE Nuclear Energy Pa9343 GE-NE-B1100786-02
- 3. COMPARISON WITH OTHER SURVEILLANCE DATA The evaluation of the shift in the RT;m for Limerick 2 (see Appendix A) was performed using the techniques of Rev 2 for vessel material and the predicted fluence (i.e., no additional sur;eillance data). These predicted values of RTim shift indicate that the Limerick 2 vessel will not experience a large shift over vessel life. To confirm the conservative predicted shift plus margin values (used to modify the surveillance program i
schedule), a comparison has been made between calculated shift and fluence values, and actual measured surveillance data from other BWRs.
A significant number of surveillance capsules from BWRs have been tested. Table 3-1 is a tabulation of the base metal results from these surveillance programs. The most significant feature, for a range of material chemistries and 11uences, is that the expected shift is bounded by the calculated Rev 2 shift plus margin. For example, the measured.BWR/4, 251" vessel (similar to Limerick 2) shifts are less than the predicted Rev 2 shift plus margin values by an average of 28'F (based upon the 5 complete data sets). For BWR/4-251 capsules, the average first capsule shift obsened was 17 F, while the average predicted shift plus margin was 45*F. This data indicates that the Limerick 2 shift (measured at 8 EFPY) will be small and may not be distinguishable from data scatter.
Similarly, Table 3-2 lists surveillance capsule data for weld material. The measured shifts are bounded by the predicted shift plus margin values. BWR/4-251 weld data (for the 6 complete data sets) shows the predicted shift plus margin to exceed the measured values by an average of 47 F. The average shift observed was 20*F, while the predicted shift plus margin was 67 F.
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l The predicted shift values are plotted against the measured shifts in Figures 3-1 and 3-2 for all BWR data available; the data is from Tables 3-1 and 3-2, respectively. These graphs show that the measured shifts are bounded by the predicted shift the margin l
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ECR LG 98-OiB57 GE Nuclear Energy Paae 44 GE-NE-B1900786-02 term [2]. Based on these data, the measured shift for Limerick 2 would be conservatively bounded by the Rev 2 prediction.
Since fluence has a significant effect on the Rev 2 calculation, use of an appropriate fluence value is essential for accurate shift prediction. The shift + margin predictions in Tables 3-1 and 3-2 utilize fluence values determined from flux wires removed early in plant life. In the case of Limerick 2, no capsule was removed to determine the fluence.
However, Limerick 2 uses a conservative estimate of the fluence [4] which is expected to bound the actual fluence.
Comparison of actual flux wire data from similar I.lants (251" Vessel ID, 764 Fuel Bundles) with the LGS 2 flux estimate has shown the actual results to be conservatively bounded by the prediction. Additional details regarding the conservative fluence estimate and the lack of flux wire capsule data are contained in Appendix B. Based on this information, the fluence used for the ART calculations (as described in Appendix A) for LGS 2 is considered conservative.
Other than fluence, the most uignificant effect on the ART is the chemistry factor (CF).
The CF is determined from the copper and nickel levels, copper having the more significant effect.
A study has been performed [5] on the copper levels present in BWR beltline materials, in response to NRC letter 92-01, Supplement 1. The intent was to identify the plants with significant variation in the reported copper levels. For Limerick 2, the copper level was determined to be consistent with the reported values with no significant variation.
Based on the evaluation of previous surveillance data of actual shifts and fluences, the expecteu 'easured fluence for LGS 2, and the chemistry of the LGS 2 vessel material, the actual shift ir LGS 2 is expected to be conservatively bounded by the calculated value of shift
- margin.
4 10
GE Nuclear Energy GE-NE-B 1100786-02
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Table 3-1: BWR Surveillance Program Results for Base Metal 11
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14
ECR LG 98-01057 GE Nucisar Energy on, so GE-NE-B1100786-02 l
- 4. PRESSURE-TEMPERATURE (P-T) CURVES i
The shift in RTer obtained from surveillance testing is used to evaluate the long tenn effects of irradiation on the fracture toughness of the vessel. The reference fracture l
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l toughness (Km) is determined using the shift in RTmr; Km is part of the calculations of the l
P-T curves performed in accordance with ASME Section XI, Appendix G. The current LGS 2 P-T curves were calculated with the shift in RTer corresponding to 10 EFPY.
1 The Km correlation was developed from several sets of material data on pressure vessel steel (6]. The Km curve was drawn to bound the available data. Thus the correlation has inherent conservatism.
In addition, operation of LOS2 follows the steam saturation curve, therefore, the operating temperatures are expected to be well in excess of the minimum required temperature. During normal and accident conditions, the LGS 2 vessel maintains more than adequate margins. The operational issues of Pressurized Thermal Shock (PTS) and Low Temperature Over Pressurization (LTOP) are not applicable to LGS 2. The limiting case for LGS 2 is the pressure test.
The P-T curve associated with the pressure test is calculated using the crack arrest fracture toughness, Km (Ku).
The static crack initiation fracture toughness, Ku is significantly higher than Km in the temperature range ofinterest [7]. Therefore, use of Km conservatively bounds the fracture toughness of the vessel.
Figure 4-1 is a plot of Ku and Ku as a function of T-RTer [8]. The Ku curve is shown to be lower than the Ku curve, conservatively bounding the fracture toughness. For example, at a pressure test temperature of 192*F and a vessel ART of 93'F (corresponding to 10 EFPY for Limerick 2), the fracture toughness for initiation and arrest are estimated to be:
15
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l ECR LG 98-01857
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Ku = 183.4 ksiVin l
. Km = 79.1 ksiVin.
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Thus the Ku value is approximately 2.3 times the Ku value, clearly showing Ku to
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conservatively bound the calculations.
The combination oflower bound fracture toughness, the LGS 2 operating characteristics and the conservative fracture toughness values indicate that the LGS 2 vessel fracture toughness is not a signi6 cant concern over the life of the plant.
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GE-NE-B1100786-02 GE Nuclear Energy 240 l
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ECR LG 98-01857 GENucla rEnergy Paae s2 GE-NE-B1100786-02
- 5. SUPPLEMENTAL SURVEILLANCE PROGRAM The BWR Owner's Group (BWROG) is in the midst of a supplemental test program designed to significantly increase the amount of BWR surveillance data in a systematic manner which should permit the development of a BWR-specific equivalent to Rev 2.
Description.
The BWROG Supplemental Surveillance Program (SSP) was begun in the late 1980s when the BWROG concluded from their review of BWR surveillance data the following:
1 Due to the smaller number of capsules per plant and the relatively fewer number of BWRs than PWRs, there is limited BWR surveillance data at higher fluences available to analyze; 1
The ARTS associated with Rev 2 imposed some hardships on pressure testing for BWRs, some of which might be relieved if better predictive models of the BWR embrittlement phenomenon were obtained.
In light of these issues, the BWROG prepared supplemental capsules which were installed in Cooper and Oyster Creek. One capsule from Oyster Creek was withdrawn i
l in 1996, with additional withdrawals planned for 2000 and 2002.
1 The results of the SSP will be the equivalent of 84 additional surveillance capsules, compared to about 35 which have been tested to date. These capsules were designed to systematically evaluate embrittlement trends in BWRs. For example:
18
ECR LG 98-01857 GENucl arEnsrgy_
pana m GE-NE-B1100786-02
' The capsules are positioned so that flux differs by a factor of 2. Also, irradiation times differ by a factor of 2. In this way, some capsules have matching flux but with different fluence, while some have matching fluence and a differing flux level; The materials used were selected to bound the range of chenustries in BWR beltline materials, and in most cases gr.e BWR beltline materials; Irradiations are being done in BWRs to correctly simulate conditions like temperature, neutron spectrum and transient operation.
Relationship to Limerick 2 The SSP does not contain Limerick 2 material among the materials in the capsules.
However, the SSP contains material which bounds the Limerick 2 limiting plate (the plate is the limiting material throughout plant life, so the weld is not a signi5 cant concern). The SSP Hatch I plate (Heat #C3985-2) contained in the program has a composition similar to the Limerick 2 surveillance program material, and was made by the same plate manufacturer (Lukens) in the same time period (Hatch 1,1968; j
Limerick 2,1970). The copper content of the Hatch I and LGS 2 surveillance plate are the same (0.11%) and the Hatch I nickel content is higher (0.66% vs. 0.51%).
i The resultant chemistry factors (CF) (per Rev 2) are 75 and 73 for Hatch 1 and LGS 2, respectively. The SSP results will be applicable to Limerick 2 for two reasons:
Generically, the SSP results will be from representative environmental conditions on materials representative of all BWRs, including LGS 2; Specifically, results will be developed which will provide information on a material which is expected to respond to irradiation similar to the plate in the Limerick 2 surveillance program.
The SSP capsules, when tested, will have collected between 5 x 10" n/cm (12.3 EFPY 2
for LGS 2 at 1/4T) and 2 x 10" n/cm (49.2 EFPY for LGS 2 at 1/4T) fluence. Thus, the 2
19 1
)
. ~ ~.
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~...
ECR LG 98-01857 GE Nuciser Energy Paae 54 GE-NE-B1100786-02 results of the SSP are complementary to the LGS 2 surveillance program such that postponement of the capsule withdrawals will have minimal impact on the understanding ofirradiation effects on the LGS 2 vessel.
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ECR LG 98-01857 GE Nuclaar Energy Page 55 GE-NE-B1100786-02 1
- 6. REVISED SURVEILLANCE SCHEDULE l
The surveillance program is intended to characterize the vessel propenies as a function of irradiation over the life of LGS 2. The Charpy impact energy obtained from the prescribed testing is used.to evaluate the reference fracture toughness of the LGS 2 vessel (Km) in accordance with ASME Section III, Appendix G.
The schedule for the surveillance program testing should be designed to obtain the best data, while maintaining safe operation.
l The expected change in fracture toughness of the LGS 2 plate material (the limiting beltline material) as a function of EFPY is plotted in Figure 6-1. Since the pressure test is the 1
limiting case, the calculated Ka is for a 1045 psig pressure test. The pressure test temperature was modified at selected intervals for illustration purposes. This figure shows that the Km used to calculate the P-T curves is expected to conservatively bound the required vessel fracture toughness.
Since the Km is considered a conservative prediction, and the SSP will identify a greater than expected shift relative to LGS 2, the first surveillance capsule testing should be at the time at which a majority of the shift in the vessel RTer has been achieved. Early testing 1
of the surveillance plate specimens may result in the measured shift being less than the data scatter (sometimes resulting in negative shifts in RTer). Correct selection of the removal i
time will ensure useful data from all specimens. If the shift is greater than expected, then l-the margin present in the P-T calculations together with the limiting fracture toughness represent an added margin of safety.
1 Since the SSP can be used to identify anomalous shifts, the first surveillance capsule testing schedule should be developed to measure a significant ponion of the fracture toughness change, as measured by ARTer. Since the limiting plate material for the vessel has a low expected ARTmr f 48'F (at 1/4T) over the life of the plant, the recommended o
schedule should be designed to measure a majority of ARTer of the plate material. Given 21 1
ECR LG 98-01857 GE Nuclear Energy Page 56 GE-NE-B1100786-02 the low expected shift, a criteria of 75% of the expected shin in RTmyr of the plate material was selected to determine the revised schedule. For Limerick 2, 75% of the expected shift is 0.75 (48) = 36 F.
Figure 6-2 is a plot of the capsule shift in RTmyr as a function of EFPY. The surveillance l
capsule material will experience a shift of 37'F in ARTmn over the life of the plant. Using a criteria of 75% of the expected shift of the limiting vessel material (36 F), the capsule will experience this shift for the plate material at approximately 29 EFPY. The removal schedule will be set to 15 EFPY to provide consistency with Limerick Unit 1. In support of this schedule, at 15 EFPY an expected 26*F shift will have occurred in the surveillance capsule limiting material, which will provide sufficient data to detennine the required 1
i vessel material properties. In addition, upon evaluation of the SSP materials, the Limerick 2 schedule can be further evaluated.
The fluence data, as determined from the surveillance capsule flux wires at 15 EFPY, will provide an accurate indication of neutron fluence. As noted in Section 3, the current predicted fluence is conservative. The flux wires in the capsule withdrawn at 15 EFPY will be used to modify the predicted fluence, meeting the requirements of the Limerick 2 Technical Specifications. The use of the flux wires at 15 EFPY will meet the requirements of 10CFR50 Appendix H and ASTM E185.
It is also appropriate at this time to extend the schedule for withdrawal of the LGS 2 second capsule. The current schedule specifies withdrawal of the second capsule at 20 EFPY. Based upon the information provided in this report suppor:ing withdrawal of the first capsule at 15 EFPY, there will be an insignificant shift in material properties at 20 EFPY, after only an additional 5 EFPY. It is appropriate to extend this schedule to 30 EFPY to meet the intent of ASTM E185-82 [2] such that withdrawal of the second capsule occurs before the accumulated neutron fluence of the capsule corresponds to the approximate EOL fluence at the reactor inner wall location. It may also be noted that the current schedule for withdrawal of the second capsule from Limerier Unit 1 is 30 EFPY.
22
GE-NE-B1100786-02 GE Nuclear Energy limerick 2 I/4T RPV Plate K1 vs EFPY for Perssure of 1945 psig 2(Y) 180 -
gg3_
- - - - Requinxi Ki Plate KIR 14 0 -
m 120 -
o s
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'u r-8O y 1(X) -
ee N
m?
"9
- x... a... N........~..... ~.......
1 80 - m.....
10 EFPYP-T Cuives 16 EFPYP-T Curves 24 tFPYP-T Cuives _ 32 EFPY P-T Curves __
~
Pressuit Test 192*F Pressure Test 207'F Picssure Test 215*F Pressure Test 221'F 40 -
20 -
0 0
4 8
12 16 20 24 28 32 DPY Figure 61: Km vs. EFPY for Limerick 2 Plate Material 23
GE-NE-B 1100786-02 GE Nuclear Energy 40 -
75% ofvessel ARTndt(36 F) 35 I
Capsule ARTndt(26 F) 1 y
30
\\
1
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Correspondingsurveillance 25 6
capsule willidrawal l
e scledule l
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Recommended surveillance I
capsule withdrawal I
I scledule: 15 EFPY I
o 8:
0 2
4 6
8 10 12 14 16 18 20 22 24 26 28 30
-32 I
EFPY Figure 6-2: Predicted shill vs. EFPY, Limerick 2 Surveillance Capsule Plate t-24 k
-..m.. m.
m.
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ECR LG 98-01857 GE Nucl:ar En:rgy pm, m GE-NE-B1100786-02
- 7. CONCLUSIONS The purpose of the vessel surveillance program is to characterize the vessel properties as a function ofirradiation. The original schedule for Limerick 2 was a withdrawal schedule of eight (8) EFPY for the first surveillance capsule.
Schedules developed according to 10CFR50, Appendix H, however, are general guidelines for all reactor pressure vessels. The schedules do not take into account some specific characteristics of Limerick 2 such as low fluence and good alloy chemistry (0.01% - 0.15% copper), which results in a low shift in RTer (especially for the LGS 2 plate material).
If the first capsule is removed and tested according to the current schedule (8 EFPY), the data obtained for the plate specimens may be heavily affected by the scatter in Charpy results.
Early information on a material similar to the limiting LGS 2 plate material can be obtained from the SSF to identify anomalous shifts, so the LGS 2 surveillance schedule should be extended. The schedule can be extended for the following reasons:
- 1. Evaluation of similar data obtained from actual surveillance programs has shown the measured fluence, shift and chemistry are bounded by expected values. In particular, the BWR/4 data has shown small RTer shifts for capsules removed from vessels similar to Limerick 2. Therefore, the surveillance capsule withdrawal schedule should be extended based on the conservatism in the calculated shift ofRTer.
- 2. In addition, the P-T curves contain inherent conservatism, as noted in Section 4. The fracture toughness values used for these calculations are considered to be lower bound values and are signiScantly less than the crack initiation fracture toughness in the temperature range ofinterest. At operating temperatures, LGS 2 maintains more than adequate margins; the limiting condition is the pressure test. This conservatism i
25
ECR LG 98-01857 GE Nucl:arEn:rgy Paae so GE-NE-B1100786-02 provides an added margin of safety; therefore, the capsule withdrawal schedule can be modi 6ed.
- 3. In addition, the SSP data will complement the available data on surveillance specimens and also identify any anomalous information inl the predicted values.
This charactenzition will enhance the understanding of vessel embrittlement issues and provide data for LGS 2 using a plate similar to the limiting plate material. Hence the change in schedule for the LGS 2 surveillance specimens will not have a significant effect on the understanding of vessel irradiation issues.
These reasons justify extending the withdrawal schedule while maintaining reactor safety margins, and provide for more accurate measured data near EOL.
Therefore, the surveillance schedule should be modified.
The material property of most concern is the fracture toughness of the vessel; the surveillance schedule should be based on evaluation of this property. Since the fracture toughness (Km) is dependent on the shift in RTer, the optimum EFPY for removal of the capsule ensures useful data (measuring significant shift), while identifying any anomalous conditions. If such an anomalous shift were to occur (which is unlikely), the margin between Ka (primary membrane fracture toughness) and Km, as well as the inherent conservatism of the calculations, can provide a sufficient safety margin for extending the surveillance schedule. In addition, operation of LGS 2 follows the steam saturation curve; the operating temperatures are expected to be well in excess of the minimum required temperature.
As shown in Section 6, the appropriate ARTwr value selected was 75% of the predicted beltline material 32 EFPY change in ARTer.
Using this value to determine the appropriate shift in the capsule (hence the appropriate EFPY), the recommended
. withdrawal schedule for the first Limerick 2 surveillance capsule is 15 EFPY. The schedule for the second capsule (currently 20 EFPY) i.c changed to 30 EFPY at this time, 26
ECR LG 98-01857 GE Nuclear Engrgy Pa9e 61 GE-NE-51100786-02 l
4 but may be re-evaluated upon removal and evaluation of the first capsule. This new schedule meets the intent of ASTM E185-82, as the nrst capsule would be removed with the fluence being less than 5 x 10" n/cm and the value of ARTart would be less than 2
50 F, and the second capsule would be removed before the accumulated neutron fluence of the capsule corresponds to the approximate EOL fluence.at the reactor inner wall
. location. Removal of each capsule at the appropriate EFPY will obtain the most useful j
data for fracture toughness predictions.
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i ECR LG 98-01857 GE Nuclect Energy Page 62 GE-NE-B1100786-02 i
- 8. ' REFERENCES
- 1. " Reactor Vessel Material Surveillance Program Requirements," Appendix H to l
Part 50 of Title 10 of the Code of Federal Reculations, December 1995.
- 2. ASTM E185-82, f
- 3. " Radiation Embrittlement of Reactor Vessel Materials," U.S. NRC Regulatory Guide 1.99, Revision 2, May 1988.
l l
4.
Limerick Generating Station (Unit 2) UFSAR.
- 5. " Bounding Assessment of BWR/2-6 Reactor Pressure Vessel Integrity Issues,"
BWR VIP-0SNP, November 1995.
- 6. S.T. Rolfe and J.M Parsom, Fracture and Fatigue Control in Structures, Prentice-
)
Hall, Inc., New Jerse:/,1977, p. 447.
i
- 7. Ibid., p. 455.
- 8. ASME Section XI, Appendix A,1992 Edition through Summer 1993 Addenda.
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28
l ECR LG 98-01857
. GE Nucl3ar Energy Page 63 GE-NE-B1100786-02 APPENDIX A ADJUSTED REFERENCE TEMPERATURE (ART) CALCULATION l
l 29 l..
ECR LG 98-01657 GE Nucisar Encrgy Page 64 GE-NE-B1100786-02 i
The ART is, according to Rev 2, a function of the initial RTmr, the shift, and a margin term. The shift in RTer is dependent on the chemistry (specifically copper and nickel) and fluence. The methods of Rev 2 are used to determine the ART; the procedure used i
depends on whether or not surveillance specimen data is available.
1 In order to re $ valuate the surveillance specimen program schedule, the ART for both the vesselitself and the specimens must be calculated. For Limerick 2, surveillance specimens have not been tested, which requires the procedure of evaluating ART without surveillance specimens, as described below.
The ART for each beltline materialis given by the following equation:
ART = InitialRTmr - ARTar - Margin (1)
Initial RTer is the reference temperature determined according to AShE Section III, Paragraph NB-2331 for the unitradiated material.
The shift in the reference temperature, ARTer, is determined by a combination of the chemistry and fluence as shown by equation (2):
i
' "'Il (2)
ARTe r = C F
- f 2'-
The CF is the chemistry factor (dependent on the copper and nickel content) and is determined from the tables for weld and base material in Rev 2. The fluence, f, at any depth in the vessel wall, is determined by equation (3),
l L
f = f,, * (e- )
(3) l l
l l
30 l'.
~.
B-p ECR LG 98-01857 GE Nuctsar Energy Page 65 GE-NE-B1100766-02 where fs is the calculated neutron fluence at the vessel ID and x is the depth into the vessel measured from the inner (wetted) surface. For these calculations, the value of f(at 2
1/4T) used was 1.3 x 10" n/cm, obtained from Paragraph 5.3.1.6.3 of the LGS UFSAR.
The Margin term is included to obtain the upper bound values of the ART. Since the Margin term provides upper bound values of the ART (which is a function of CF and fluence), it is unnecessary to add extra conservatism by using the upper bound fluence.
Any uncertainty in the fluence is captured by the Margin term. The Margin term is given by equation (4):
Margm = 2dcrj 4.cri (4) i-(
where ci= standard deviation of the initial RTer c3 = standard deviation for ARTer l
The standard deviation for ARTer, c3, is assumed to be 28 F for welds and 17*F for base i
metal, except that c3 need not exceed 0.50 times the mean ARTer [2]. The conservative nature of the initial RTmr determination results in ci being equal to zero.
i' Using equations (1) through (4), the ART can be calculated for plants with no surveillance data, including Limerick 2.
i EXAMPLE CALCULATION Tc better illustrate the ART methodology, the following calculation was performed for the limiting Limerick 2 base material (Heat #B3416-1); this material's chemistry and initial RTer bound that for the base material in the surveillance capsule. (The %Cu is 0.14 compared to the capsule material 0.11%Cu; the %Ni is 0.65 compared to the capsule l
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ECR LG 98-01857 GE Nuclasr Enstgy Page 66 GE-NE-B1100786-02 L
material 0.51%Ni, and the initial RTer is 40 F compared to the capsule material 10 F.)
The data was obtained from Table 5.3-5 of the LGS UFSAR:
i Initial RTmr:
40*F Nickel:
0.65 %
Copper:
0.14 %
8 2
Peak Fluence:
1.3 x 10 n/cm (32 EFPY at 1/4T)
Wall Thickness:
6.19 inches From Table 2 of Rev 2, the chemistry factor for this heat of matenal is 101. The fluence 58 2
at the 1/4T depth,1,3 x 10 n/cm, was used. The change in reference temperature, ARTer, is calculated according to equation (2);
ARTmr = 101 *0.13( 2* ' 8"")
4RTmr = 101*0.471 = 47.6"F = 48*F For the margin term, the standard deviation of the initial RTer, er, is assumed to be zero.
The standard deviation for ARTer, c, is 17 F, as it is base metal.
a i
Therefore, using equation (1), the ART at 32 EFPY for plate B3416-1 is:
j ART = 40 + 48 + 34 = 122 F This calculation was repeated for all of the vessel beltline materials. The results of the j
calculations for all the beltline materials are shown in Table A-1. Figure A-1 is a plot of the ART against EFPY for the expected plant lifetime for the limiting (at EOL) plate and wcld materials.
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32
GE Nuclear Energy GE-NE-B 1100/U6-02 Pie.
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Seese A9 6498924424927 AE 9 89 9 90 522 44 17 4 00 28 9 54 9 L13 4
$1 Seese AB 491P674 5/S4 89927 AG 9 93 6 92 41
-Se 19 3 es 97 19 3 3s 4
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-99 19 3 09 97 19 3 35 6
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-I t LITI MOEZLE
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- Sesseillence Meerial Table A-1: Limerick 2 RPV Beltline Material Data 33
15 GE Nuclear Energy '
GE-NE-B1100786-02 140 120 100
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ECR LG 98-01857 GE Nublaar Enargy Pan 269 GE-NE-B1100786-02 l'
APPENDIX B i
COMPARISON OF MEASURED DATA WITH FLUENCE PREDICTION L
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I ECR LG 98-01857 l
Page 70 GENuclearEnergy Engineerng and Ucansing Consulting Services 17sCutmet Annue M/C747 San.tose. CA 95125 140W925-1472 BJB-9710R1 November 20,1997 TO:
Bob McCall Betty J. Branlu%nd l
l FROM:
SUBJECT:
Limerick Generating Station Unit 1 and 2 First Cycle Beltline Dosimeter Missing Data
REFERENCE:
- 1) Letter from D. W. Diefenderfer to G. M. Zaimis, " Limerick Generating Station Unit 1 Belt Line Dosimeter," GE-NEBO, San Jose, CA, July 15,1987.
In 1987 the Reference 1 report addressed the concern of the missing first cycle flux wire l
dosimeters for Limerick Unit 1. The purpose of this letter is to update the 1987 letter, j
include the concem that first cycle flux wires are unavailable for Limerick Units 1 and 2, j
, and provide further information regarding the baseline fluence evaluation. This letter addresses the influence of the unavailable flux wire data on the existing P -T curves particularly in light of the power rerate condition. Finally, compliance with 10CFR50 l
L Appendix H and ASTM E185 including the concern of saturation of the flux wires as I
discussed in ASTM E185-82 is considered.
L Summary l
The wire test results would be used to confirm that the predicted peak ID surface fluence (1.7 x 10 n/cm ) used to develop the pre-rerate P-T curves was conservative (note that a 2
2 2
fluence of 1.7 x 10 ' n/cm corresponds to a flux of1.7 x 10' n/cm -sec). GE has reviewed the flux wire data collected to-date for numerous BWRs with a 251" vessel ID 2
and 764 fuel bundles and determined that a flux of1.7 x 10' n/cm -sec is conservative. In addition, sufficient data from similar plants exists to address the concern of flux wire saturation. Therefore, the station can be confident that the P-T curves developed using 2
2 the predicted flux of1.7 x 10' n/cm -see for rated power and 1.9 x 10' n/cm -sec for i
rerated power are conservative.
l i
l ECR LG 96-01057 Page 71 BJB-9710 11/20/97 Baseline Predicted Flux and Fluence The baseline flux was determined using a generic one dimensional Sn calculation in 1983 i
for BWR/4&Ss with a 251" vesselID and 764 fuel bundles. The calculated peak vesselID 2
surface flux at energies greater than 1MeV is 8.5 x 10'n/cm -sec. A safety factor of 2.0 was conservatively applied to establish a baseline predicted flux and end-of-life (EOL) 2 fluence (flux = 1.7 x 10'n/cm:- sec and EOL fluence = 1.7 x 10n/cm ). Since the lead factors (ratic of capsule flux to vessel ID surface flux) for plants of this size and fuel l
configuration range between 0.94 and 1.01, then this flux corresponds to a capsule flux of 2
2
~1.6 x 10'n/cm - sec (i.e.,1.7 x 10'n/cm - sec*0.94).
l l
Note that in the Reference 1 letter, the predicted flux for some plants was reported to be l
1.3 x 10'n/cm -sec. This flux was determined from the same one dimensional Sn 2
calculation, but using a safety factor of 1.5 instead of 2.0. The flux of 1.1 x 10'n/cm -sec
)
2 corresponds to the 1/4T flux.
i Comparison of Predicted to Measured Data l
Table 1 shows nineteen flux measurements taken between 1978 and 1996; ten ricasurements were from first cycle dosimeters, while the remaining nine were from capsules withdrawn between 6 and 9 EFPY. The population represents BWR 4 and 5 plants with 764 bundles and a thermal power rating of 3293 and 3323 MWth for the 2
BWR/4 and 5, respectively. The average value of that data is 6.7 x 10'n/cm - sec and the standard devhtion is 1.9 x 10'n/cm - sec. Therefore, the predicted flux used to develop 2
- the Limerick 1 and 2 P-T cur,es is well above two standard deviations of the measured 2
2 data (i.e., 6.7 x 10'n/cm - sec + 2(1.9 x 10'n/cm sec) = 1.1 x 10'n/cm - sec, measured 2
compared to 1.6 x 10'n/cm - sec, predicted capsule flux). There is a factor of approximately 1.5 between the measured data with two standard deviations and the value used to develop the P-T curves. Also, the predicted capsule flux is 34% grea'er than the 2
2 largest measured value (i.e.,1.2 x 10'n/cm - sec
- 1.34 = 1.6 x 10'n/cm - sec,. Iaerefore, the P-T curves are conservative.
Consideration of Power Rerate 2
For power rerate the original fluence (1.7 x 10 'n/cm ) was assumed to increase proportional to the increase in power. To be conservative, it was assumed that the power increased from the first day of commercial operation rather than when the power rerate was implemented. Therefore, for the proposed 10% power rerate the fluence was increased by 10% to 1.9 x 10n/cm. Since Limerick Units 1 and 2 are licensed for 5%
2 power rerate (from 3293 MWth to 3458 MWth), the fluence used to develop the P-T curves for power rerate are additionally conservative.
I
{
2-
ECR LG 90-01857 BJB-9710 Page 72 11/20/97 10CFR50 Appendix H and ASTM E185 The material surveillance program for both Limerick Units 1 and 2 are designed to meet the intent of ASTM E185-73. There are no requirements with ASTM E185-73 to include first cycle dosimetry. 10CFR50 Appendix H permits the use of revisions of ASTM E185 e
up to and including the 1982 version. In ASTM E185-82, although there are no specific requirements for first cycle dosimetry, article 7.3.3 states that separate dosimeter capsules j
should also be used to monitor radiation conditions independent of the specimen capsules i
ifit is expect.ed that the withdrawal schedule will otherwise result in saturation of the 3
dosimeter activitiet.
2 The suryditaam capsules for I.imerick Units 1 and 2 contain iron and copper wires. The j
iron wire isotope, Mn-54, has a half-life of 312.5 days. 'Ihe copper isotope is Co-60, has a half-life of 5.27 years. GE experience with these types of wires, pulled after 10 years of operation has shown that, while the iron wires reached a saturated condition after 4 or 5 years, consistent results with the "non-saturated" copper wires are achieved as long as an j
accurate daily power history is known. Extending the copper wire dosimetry to a i
saturated (or nearly saturated) condition is covered in ASTM E261-90, paingraph 9.5, which is referenced from the copper wire dosimetry method, ASTM E523-87. Paragraph 9.5 authorizes the use of the power history in the saturation condition, which will allow I
measurement of any saturated wire beyond the time of saturation (i.e., - 3 years for the iron wire and ~15 years for the copper wire). Based on these methods, saturation is not an issue for Unit I and 2 dosimetry. Therefore, there is no requirement nor recommendation for first cycle flux data based on 10CFR50 Appendix H, l
4 ASTM E185-73, and ASTM E185-82.
Conclusion The data from similar plants demonstrate there is sufficient conservatism in the predicted fluence for both rated and rerated power. Conservatism in the fluence assures that the P-T curves are not impacted by the fact that the first cycle flux wire data is unavailable for both Limerick Units 1 and 2.
ECR LG 99-01857 Page 73 BJB-9710 11/20/97 TABLE 1 - Flux Wire Measurements from 251" Vessel ID BWR/4&5's With 764 Fuel Bundles Plant ID Measured Capsule Flux (N/cm -s)
First Cycle Measurements at Capsule Location P
7.9E+8 J
8.2E+8 Y
1.2E%
AM 1.0E+9 AW 6.3E+8 AX 4.7E+8 AZ 5.0E+8 AK 4.8E+8 AH 4.9E+8 AY 6.2E+8 Surveillece Capsub Flux Measurcmcists P
7.50E+08 J
6.80E+08 Y
5.90E+08 AW 6.60E+08 AT 6.70E%8 AX 4.41E+08 AZ 5.22EM8 i
AK 6.85E%8 AY 7.49 EMS i
Minimum 4.4E+08 Maximum 1.2E+09 Average 6.7E+08 Standard Deviation 1.9E+08 j
i 251" VesselID BWR/4&5's Mux With 764 Fuel Bundles (N/cm'-s)
Predicted ID surface Flux with Safety Factor of 2.0 1.7E+09 Predicted Capsule Flux 1.6E+09 j
3 (ID surface flux capsule lead factor) a (1.7E+09
- 0.94)
Measured Data Average plus 2e 1.1E+09 Maximum Measured Value 1.2E+09..