ML061650080

Update to NRC Reactor Vessel Integrity Database and Exemption Request for Alternate Material Properties Basis Per 10 CFR 50.60(b)

ML061650080
Person / Time
Site:	Surry
Issue date:	06/13/2006
From:	Grecheck E Virginia Electric & Power Co (VEPCO)
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
06-434 BAW-2494, Rev 1
Download: ML061650080 (70)

Text

VIRGINIA ELECTRIC AND POWER COMPANY RICHMOND, VIRGINIA 2326 1 June 13, 2006 U.S. Nuclear Regulatory Commission Serial No.06-434 Attention: Document Control Desk NL&OS/GDM RO Washington, D.C. 20555 Docket Nos. 50-280 50-281 License Nos. DRP-32 DRP-37 VIRGINIA ELECTRIC AND POWER COMPANY SURRY POWER STATION UNITS 1 AND 2 UPDATE TO NRC REACTOR VESSEL INTEGRITY DATABASE AND EXEMPTION REQUEST FOR ALTERNATE MATERIAL PROPERTIES BASIS PER 10 CFR 5;0.60(b)

Virginia Electric and Power Company (Dominion) has prepared an update to the NRC's Reactor Vessel Integrity Database to document the results of the most recent 10 CFR !50.61 Pressurized Thermal Shock (PTS) screening calculations, Nil Ductility Transition Reference Temperature (RTNDT)values, and Upper Shelf Energy values.

The calculations utilize revised initial (unirradiated) RTNDTvalues for Linde 80 weld materials based on Topical Report BAW-2308, Revision 1-A. The calculations assume fluence v8aluesapplicable to the current 60-year license period that accommodate the use of Westinghouse Integral Fuel Burnable Absorber (IFBA) in Surry Units 1 and 2 and removal of Flux Suppression Inserts (FSls) for Surry Unit 1. 10 CFR 50.61 PTS screening1 criteria and 10 CFR 50 Appendix G Upper Shelf Energy criteria continue to be met for Surry Units 1 and 2 reactor vessel materials. The existing Reactor Coolant System (RCS) Pressureflemperature (Pfl) operating limits, Low Temperature Overpressure Protection System (LTOPS) setpoints, and LTOPS enabling temperature presently in the Surry Units 1 and 2 Technical Specifications continue to be valid and conserva1:ive through their period of applicability (i.e., 28.8 EFPY and 29.4 EFPY for Surry Units 1 and 2, respectively).

Also, pursuant to 10 CFR 50.1 2 and 10 CFR 50.60(b), Dominion requests an exemption from the requirements of 10 CFR 50.61 and 10 CFR 50 Appendix G to revise certain Surry reactor pressure vessel material initial (unirradiated) properties using Framatome ANP Topical Report BAW-2308, Revision 1-A. The Topical Report provides revised initial (unirradiated) reference temperatures for the Linde 80 weld materials present in the reactor pressure vessels of Surry Units 1 and 2 and was approved by the NRC in August 2005.

In May 2005, Dominion participated in a teleconference with the NRC Materials Branch staff to discuss Dominion's pending exemption request and aspects of the NRC's review

Serial No.06-434 Docket Nos. 50-280, 281 Page 2 of 3 of Topical Report BAW-2308, Revision 1. Regarding the timing of the exemption request review, the NRC staff cited 10 CFR 50.61 (b)(3) and provided an interpretation that could be used to allow making significant core design changes while the exemption request was being reviewed by the NRC. The attached exemption request is submitted based on the interpretation that the schedule for implementing flux suppression measures can take into account the timing of the submittal, NRC review, and approval of calculations that apply new analysis techniques. The Surry core design changes related to FSI removal and IFBA implementation are being made with the realistic expectation of obtaining NRC staff approval. This interpretation is reasonable, since the relevant (effects of these changes on reactor vessel integrity are very long-term in nature, thus providing sufficient time for the necessary regulatory interactions.

Furthermore, the NRC has already approved the new analysis techniques described in Topical Report BAW-2308, Revision 1-A.

A discussion of the proposed changes to the reactor vessel materials evaluations for Surry Power Station is provided in Attachment 1, and the request for exemption is included in Attachment 2. An update to the NRC Reactor Vessel Integrity Database (RVID) far Surry is included as Attachment 3. Framatome ANP Report BAW-2494, Revision 1, "Low Upper-Shelf Toughness Fracture Mechanics Analysis of Reactor Vessel of Surry Units 1 and 2 for Extended Life through 48 Effective Full Power Years,"

dated September 2005, is provided as Attachment 4.

If you ha.ve any questions or require additional information regarding this submittal, please contact Mr. Gary Miller at (804) 273-2771.

Sincerely, E. S. Grecheck Vice President - Nuclear Support Services Attachments:

1. Proposed Changes to Reactor Vessel Materials Evaluations

2. Regulatory Basis and Request for Exemption

3. Reactor Vessel Materials Data Tables

4. Framatome ANP Report BAW-2494, Revision 1

Serial No.06-434 Docket Nos. 50-280, 281 Page 3 of 3 Commitment made in this letter:

1. Sections of the Surry Units 1 and 2 Updated Final Safety Analysis Report (UI'SAR) will be revised to reflect implementation of the revised design basis analyses described herein. Following NRC approval of the exemption request associated with this submittal, a UFSAR revision will be made in accordance with the requirements of 10 CFR 50.71 (e).

cc: U.S. Nuclear Regulatory Commission Region II Sam Nunn Atlanta Federal Center 61 Forsyth Street, SW Suite 23T85 Atlanta, Georgia 30303 Mr. N. P. Garrett NFlC Senior Resident Inspector Surry Power Station Mr. S. R. Monarque NFlC Project Manager U. S. Nuclear Regulatory Commission On~eWhite Flint North 11:555 Rockville Pike Mail Stop 8-HI2 Rockville, MD 20852

Serial No.06-434 Docket Nos. 50-280,281 Attachment 1 P ~ O D OChanaes S ~ ~ to Reactor Vessel Materials Evaluations Surry Power Station Units 1 and 2 Virginia Electric and Power Company (Dominion)

Serial No.06-434 Docket Nos. 50-280, 281 Attachment 1 Pro~osedChanaes to Reactor Vessel Materials Evaluations Surrv Power Station Units 1 and 2 An update to the NRC's Reactor Vessel Integrity Database (RVID) has been prepared to docurrient the results of Virginia Electric and Power Company's (Dominion's) most recent 10 CFR 50.61 Pressurized Thermal Shock (PTS) screening calculations, Nil Ductility 'Transition Reference Temperature (RTNDT)values, and Upper Shelf Energy values for Surry Units 1 and 2. The calculations utilize revised initial (unirradiated)

R T N ~villues T for Linde 80 weld materials based on Framatome ANP Topical Report BAW-2308, Revision 1-A, "Initial RTNDTof Linde 80 Weld Materials" (Reference 1). The NRC approved topical Report BAW-2308, Revision 1-A in August 2005 (Reference 2). The calculations assume fluence values applicable to the current 60-year license period that accommodate the use of Westinghouse Integral Fuel Burnable Absorber (IFBA) in Surry Units 1 and 2, and removal of Flux Suppression Inserts (F-Sls) for Surry Unit 1. 10 CFR 50.61 PTS screening criteria and 10 CFR 50 Appendix G Upper Shelf Energy criteria continue to be met for all Surry Units 1 and 2 reactor vessel materials. The existing RCS Pressurefremperature (PTT) limits, Low Temperature Overpressure Protection System (LTOPS) setpoints, and LTOPS enabling temperatlure presently in the Surry Units 1 and 2 Technical Specifications continue to be valid and conservative through their period of applicability [i.e., 28.8 effective full power years (EFPY) and 29.4 EFPY for Surry Units 1 and 2, respectively].

The use of Topical Report BAW-2308, Revision 1-A requires the submittal of an exemption request pursuant to 10 CFR 50.12 and 10 CFR 50.60(b) from the requiremlents of 10 CFR 50.61, and 10 CFR 50 Appendix G. An exemption request for Surry Uniits 1 and 2 is provided in Attachment 2.

2.0 Background Beginning with Surry Unit 1 Cycle 13, FSls were placed into certain Surry Unit 1 peripheral fuel assembly locations to reduce the neutron fluence at the limiting reactor pressure vessel (RPV) weld locations. FSls were implemented to ensure that the limiting Surry Unit 1 RPV beltline weld materials would continue to meet the applicable PTS screening criteria provided in 10 CFR 50.61. The NRC was notified of Dominion's plan to use FSls in a Surry Unit 1 letter dated December 10, 1991 (Reference 4).

The most recent update to the Reactor Vessel Integrity Database (RVID) for Surry Units 1 and 2 was provided to the NRC in Dominion letter dated March 27, 2003 (Reference 3). This RVID update reported RTPTSvalues based on RPV neutron fluence estimates corresponding to the end of the original 40-year license period. The RPV neutron fluence analyses that supported the PTS assessment of Reference 3 were based on the assumption that future operating reactor cores would contain FSls for Surry Unit 1, and that discrete burnable poison rods would be used for excess reactivity control in both Surry units.

Page 1 of 11

Serial No.06-434 Docket Nos. 50-280, 281 Attachment 1 The FSIs are qualified to remain in the Surry Unit 1 core through the end of Cycle 21, after which they would require replacement. The FSls have a design lifetime of 12 EFPY'. Dominion has elected to remove the FSls at the end of Surry Unit 1 Cycle

20. Surry Unit 1 Cycle 21, which began operation in the Spring of 2006, does not contain F'Sls. The removal of FSls precludes the need to procure at least two more sets of FSls during the remaining life of the plant, which will reduce lifetime radioactive waste. Pdso, FSI removal reduces fuel handling requirements during refueling outages and disposal operations. Surry Unit 2 is unaffected by this change, since it has never used FSls.

Dominion implemented the use of Westinghouse IFBA in Surry Unit 1 Cycle 21 (Spring 2006), arid plans to implement the use of IFBA for Surry Unit 2 Cycle 21 (Fall 2006).

The IFBA transition will reduce, and eventually replace, the current practice of using discrete lburnable poison rods in the Surry core reload designs. IFBA core designs involve loading fresh fuel nearer the peripheral core locations than previous core designs using discrete burnable poison rods. The removal of FSls from Surry Unit 1, and the use of IFBA in reload core designs for Surry Units 1 and 2, will result in increased fast neutron fluence on reactor pressure vessel beltline materials.

3.0 Discussion of Changes to Previously Reported Information 3.1 E ' V Neutron Fluence Proiections RPV fast neutron fluence (E > 1.0 MeV) projections for the current 60-year renewed license period have been performed by Framatome ANP to account for FSI removal at Surry Unit 1, and IFBA reload patterns at Surry Units 1 and 2. These neutron fluence projectior~sare valid for cumulative core exposures of 48 EFPY for both Surry Units 1 and 2. Future operating cycles are assumed to be equilibrium fuel cycles with IFBA and without FSls at a capacity factor of 95%. The 60-year fast neutron fluence projections were calculated in accordance with the requirements of Regulatory Guide (RG) 1.I90 (Reference 5), as described in Framatome ANP Topical Report BAW-2241P (Reference 6). An NRC staff review of Topical Report BAW-2241P concluded that the methodollogy is acceptable for referencing in licensing applications for determining the RPV fluerice of Westinghouse designed reactors (Reference 7).

The revised fluence analyses also provide peak reactor vessel fluence results at a cumulative core exposure of 32 EFPY for each Surry unit. The fluence values at 32 EFPY can be considered bounding with respect to the original 40-year operating period for Surry Units 1 and 2. The existing Technical Specification RCS P/T limits, LTOPS setpoint, and LTOPS enabling temperature are valid and conservative through their period of applicability (i.e., 28.8 EFPY and 29.4 EFPY for Surry Units 1 and 2, respectively.) Therefore, the 32 EFPY fluence projections provided in the revised analysis represent a conservative point of comparison to previously docketed fluence analyses based on the original 40-year license period. Results from the revised fluence analyses, considering FSI removal and IFBA operating cycles, are shown below.

Page 2 of 11

Serial No.06-434 Docket Nos. 50-280, 281 Attachment 1 Table 3.1 - Revised RPV Fluence Projections for Surry Unit 1 Surry Uniit 1 I Neutron Fluence (E > 1.0 MeV)

Location Material Fluence at 32 Fluence at EFPY (n/cm2) 48 EFPY (n/cm2) lntermediate and Vessel Wid1 Inner Surface (0') 3.80E+1 5.66E+19 Lower Plates Lower Shell Longitudinal SA-149418T1554 6.40E+1 1.04E+19 Weld, L1 & L2 SA-15261299L44 Intermediate Shell SA-149418T 1554 6.78E+18 1.08E+19 Longitudinal Weld, L3 & L4 Intermediate to Lower Shell SA-1585172445 3.74E+19 5.61 E+19 Circumferential Weld, W05 SA-1650172445 Nozzle to lntermediate Shell J72612501 5.27E+18 7.75E+18 Circumferential Weld, W06 Table 3.2 - Revised RPV Fluence Projections for Surry Unit 2 p i j m i t2 I Neutron Fluence (E > 1.0 tLocation Lower Shell Longitudinal Material lntermediate and Lower Plates W F-418T1762 3.64E+1 7.62E+18 MeV)

Fluence at 32 EFPY (n/cm2)

Fluence at 48 EFPY (n/cm2) 5.38E+19 1.14E+19 W F-8I8Tl762 SA-1585172445 W F-418T1762 7.63E+18 1.14E+19 R300810227 3.62E+19 5.37E+19 J73714275 Circumferential Weld, W06 4.00E+18 6.32E+18 3.2 b i d e 80 Weld Material Pro~erties Framatorne ANP Topical Report BAW-2308, Revision I - A (Reference I ) , provides an alternate method for determining the unirradiated and adjusted RTNDTfor the Linde 80 weld materials present in the beltline region of the RPVs at Surry Unit 1 and 2. Topical Report BAW-2308, Revision 1-A, also provides revised initial (unirradiated) RTNDT values and associated uncertainties for these materials.

Page 3 of 11

Serial No.06-434 Docket Nos. 50-280, 281 Attachment 1 The NR'C approved Topical Report BAW-2308, Revision 1-A in August 2005 (Reference 2). Table 3 of the NRC Safety Evaluation Report (SER) (Reference 2) contains the revised initial reference temperature (IRTTo)and initial margin (01) values for Linde 80 weld materials that are approved by the NRC for the purpose of RPV material property determination. The approved values from Reference 2 are shown below.

Table 3.3 NRC: Staff-Accepted Initial RTToand a,Values for Linde 80 Weld Materials 1 Linde 80 Weld Material Initial RTTO( O F ) 1 Initial Margin a (OF) I

-72.5 12.3 W t s Generic Value) -47.6 17.2 The following Linde 80 weld materials are contained in the Surry Unit 1 reactor vessel: 8T1554, 299L44, and 72445.

The following Linde 80 weld materials are contained in the Surry Unit 2 reactor vessel: 8T1762 and 72445.

Note that for any Linde 80 material not specifically included in the table above, the inputs for "All Heats (Generic Value)" are to be used.

The following is stated as Condition and Limitation (4) in the NRC's Safety Evaluation for Topical Report BAW-2308, Revision 1-A:

"Any licensee who wants to utilize the methodology of TI? BA W-2308,Revision I as outlined in items (1) through (3)above, must request an exemption, per 10 CFR 50.12,from the requirements of Appendix G to 10 CFR Part 50 and I0 CFR 50.61to do so."

In the above quotation, Condition and Limitation (1) pertains to NRC-accepted values of initial (un~irradiated)reference temperature, IRTTo, and the corresponding uncertainty term, a[, for Linde 80 weld materials based on the Master Curve methodology using direct testing of fracture toughness in accordance with ASTM Standard Test Method E-1921 and ASME Code Case N-629. These values are provided in Table 3 of the NRC SER (Reference 2)) as shown above in Table 3.3.

Page 4 of 11

Serial No.06-434 Docket Nos. 50-280, 281 Attachment 1 Condition and Limitation (2) requires that a minimum chemistry factor of 167.0°F be applied when the methodology of RG 1.99, Revision 2, is used to assess the shift in initial properties due to irradiation.

Conditior~ and Limitation (3) requires that a value of OA = 28.0°F be used in the determination of the margin term, as defined in Topical Report BAW-2308, Revision 1-A, and RG 1.99, Revision 2 (Reference 12). As noted in Reference 2, the NRC staff has concluded that the use of OA = 28.0°F in conjunction with the IRTToand a1 values based on Master Curve testing, and material property shifts based on the models in RG 1.99, Revision 2, with a minimum chemistry factor of 167"F, provides an acceptable basis for RPV Linde 80 weld assessment. The Conditions and Limitations specified in Reference 2 have been met for the 10 CFR 50.61 PTS assessment and Adjusted Reference Temperatures (ART) at 114-T and 314-T, for the Surry Linde 80 weld materials addressed in this submittal.

There are no changes being proposed in this submittal for the material properties of non-Linde 80 weld materials. Chemical composition, chemistry factors, uncertainty terms, and overall margin terms are unchanged for the non-Linde 80 materials, and remain consistent with the information previously reported in Reference 3.

A summary of the material property data relevant to this submittal is included as . This data is provided as an update to the NRC's Reactor Vessel Integrity Database (RVID).

3.3 msssurized Thermal Shock Assessment Dominion has performed a PTS assessment for all Surry RPV beltline materials. The revised neutron fluence projections corresponding to the end of the current 60-year operating licenses (48 EFPY) were utilized, as provided in Section 3.1 above. The unirradiated R T N Dvalues

~ and associated uncertainties presented in Topical Report BAW-2308, Revision 1-A were used for the Linde 80 weld materials. Unirradiated RTNDTvalues and associated uncertainties for non-Linde 80 materials are unchanged from those previously provided in Reference 3. The results of the PTS screening calculations are provided in Attachment 3.

For Surry Unit 1, the limiting materials in terms of absolute value of R T ~ T are s the Intermediate to Lower Shell Circumferential Welds SA-1585172445 and SA-1650172445.

For these materials, the value of RTPTSis 2265°F versus the PTS screening criterion of 300°F for circumferential welds. This value represents a 73.5"F margin to the applicable PTS screening criterion for these materials.

For Surry Unit 1, the limiting material in terms of margin to the applicable PTS screening criterion, is the Lower Shell Longitudinal Weld SA-15261299L44. For this material, the value of RTPTSis 201.8"F versus the PTS screening criterion of 270°F for plates, forgings, and axial welds. This represents a 68.2"F margin to the applicable PTS screening criterion for this material.

Page 5 of 11

Serial No.06-434 Docket Nos. 50-280, 281 Attachment 1 For Surry Unit 2, the limiting material in terms of the absolute value of R T ~ Tand s margin to the applicable PTS screening criterion is the Intermediate to Lower Shell Circumferential Weld R300810227. For this material, the value of RTpTs is 236.4"F versus the PTS screening criterion of 300°F for circumferential welds. This represents a 63.6"F miargin to the applicable PTS screening criterion for this material.

In summary, when the revised 60-year fluence projections are considered in conjunction with the alternate initial RTNDTmethodology as described in Topical Report BAW-2308, Revision I-A, Surry Units 1 and 2 reactor vessel beltline materials meet the 10 CFR :50.61 PTS screening criteria through the end of the current 60-year operating license period.

10 CFR 50.61 Screenina Calculations without Use of BAW-2308, Revision 1-A Dominiori has performed sensitivity calculations to determine when the 10 CFR 50.61 screening criteria would be reached by the Linde 80 weld materials without employing the alternate methodology as described in BAW-2308, Revision I - A . The results of these sensitivity calculations show that the 10 CFR 50.61 PTS screening criterion of 270°F wlould be reached by Surry Unit 1 weld material SA-15261299L44 at a fast neutron fluence of 0.809E19 n/cm2.

To convert this neutron fluence value to an equivalent cumulative core burnup (effective full power year; EFPY) value, a simple linear interpolation is employed using the fluence data presented in Section 3.1, Table 3.1. At a neutron fluence of 0.809E19 n/cm2, the correspo~ndingcumulative core exposure for Surry Unit 1 is approximately 38.8 EFPY.

This cumulative core exposure is estimated to occur in approximately the year 2022 based on current projections of future operating cycles, including a 95% capacity factor.

3.4 u ) ~ eShelf r Enerav The results of Upper Shelf Energy (USE) calculations performed by Dominion are provided in Attachment 3. The percentage drops in USE values at the 114-T location within the reactor vessel wall were calculated using the RG 1.99, Revision 2, Position 1.2 methodology.

Equivalent Margin Analyses (EMAs) are required for the Surry Units 1 and 2 reactor vessel beltline materials for which either (1) initial (unirradiated) USE values were not known, alr (2) initial unirradiated USE values were available and the beltline materials USE at 1:he end of the licensed period of operation were projected to fall below the 50 ft-lb criterion specified in Section IV.A.l of 10 CFR Part 50, Appendix G.

For those weld materials which meet either of these criteria, the summary table provided in Attachment 3 displays a value of 'EMA' in place of the calculated values for USE at 1/4-T, Unirradiated USE, % Drop in USE at 1/4-T, and % Drop in USE Method.

Framatorne ANP has performed the EMAs for those Surry RPV materials that meet Page 6 of 11

Serial No.06-434 Docket Nos. 50-280,281 Attachment 1 either of the criteria (Reference 8). The EMAs are provided in Attachment 4.

The EMA documented in Reference 8 states that Surry Unit 1 beltline weld material SA-1526j299L44 is the limiting weld for the reactor vessels at Surry. The previous EMA, which is documented in Framatome Report BAW-2323 (Reference 9), determined that the rninimum ratio of material J-resistance to applied J-integral, Jo,l/J1, occurred for Surry Unit 1 beltline longitudinal weld material SA-1526. Table 6-1 of the most recently performed EMA (Reference 8), which is based on the revised 60-year fluence projectioris that include consideration of IFBA implementation and FSI removal, reaffirms that this limiting ratio occurs for Surry Unit 1 beltline longitudinal weld material SA- 1526..

The upper shelf energy EMA documented in Reference 8 includes conservative margin relative to the reactor vessel fluence projections described in Section 3.1. At Dominiorl's request, the EMA documented in Reference 8 includes reactor vessel fluence margins of 10% for circumferential welds, 20% for longitudinal welds, and 100%

for circurnferential welds above the active core height to provide a degree of retained margin in the analyses.

The EMA documented in Reference 8 concludes that the limiting reactor vessel beltline weld for Surry Units 1 and 2 satisfies the acceptance criteria of Appendix K to Section XI of the ASME Code for ductile flaw extensions and tensile stability using projected low upper-shelf Charpy impact energy levels at 48 EFPY of plant operation.

It is noted that USE values are provided in the summary table of Attachment 3 for Surry Unit 2 weld materials R3008 and SA-1585, even though the EMA documented in Reference 8 is applicable to these materials. The EMA is not cited in the summary table because 10 CFR Part 50 Appendix G calls for using the EMA if the USE value does not meet the 50 ft-lb criterion. Since Surry Unit 2 weld materials R3008 and SA-1585 continue to be above 50 ft-lbs, the USE value is reported in Attachment 3 instead of the EMA.

In summary, when the revised 60-year neutron fluence values are considered, Surry Units 1 and 2 reactor vessel beltline materials meet the 10 CFR Part 50 Appendix G requirem~entsby satisfying the 50 ft-lb USE limit, or by an EMA that demonstrates acceptable margins of safety against fracture for projected low upper-shelf Charpy impact energy levels at 48 EFPY.

3.5 Bchnical S~ecificationRCS Pressurefrem~eratureLimits The current Surry Units 1 and 2 Technical Specifications RCS P/T limits and LTOPS setpoint i ~ based e on a limiting 114-thickness (114-T) RTNDTof 228.4 OF and a limiting 314-thickness (314-T) RTNDTof 189.5OF (References 10 and 11). When the current Technica.1 Specifications RCS PIT limits and LTOPS setpoint were developed, these values of RTNDTwere determined to bound Surry Units 1 and 2 reactor vessel beltline materials at end-of-original 40-year license fluences corresponding to 28.8 EFPY and Page 7 of 11

Serial No.06-434 Docket Nos. 50-280, 281 Attachment 1 29.4 EFP'Y for Surry Units 1 and 2, respectively.

The summary table provided in Attachment 3 shows the results for the 114-T RTNDTand the 314-T' RTNDTvalues considering the 60-year fluence projections corresponding to 48 EFPY', and the Linde 80 weld material properties per Topical Report BAW-2308, Revision 1-A. At fluence projections corresponding to 48 EFPY, the limiting 114-T RTNDT value is 2225°F for Surry Unit 2 lntermediate to Lower Shell Circumferential Weld material R300810227, and the limiting 314-T RTNDTvalue is 188.6"F for Surry Unit 2 lntermediate to Lower Shell Circumferential Weld material R300810227.

Sensitivity calculations were also performed to determine the 114-T RTNDTand the 314-T RTNDT values at 32 EFPY, without crediting the alternate initial RTNm methodology described in Topical Report BAW-2308, Revision I-A. A cumulative core exposure of 32 EFPY exceeds the core exposure cited on the current Technical Specifications RCS PIT limits curves, which indicate cumulative exposure applicability limits of 28.8 EFPY for Surry Unit 1 and 29.4 EFPY for Surry Unit 2. Therefore, the 32 EFPY fluences represent a conservative point of comparison relative to the current Technical Specifications RCS PIT limits.

At fluencie projections corresponding to 32 EFPY, and without crediting the Linde 80 weld material properties per Topical Report BAW-2308, Revision 1-A, the limiting 114-T R T N D value

~ is 225.7"F for Surry Unit 1 Lower Shell Longitudinal Weld SA-15261299L44. At fluence projections corresponding to 32 EFPY and without crediting the Linde 80 weld material properties per Topical Report BAW-2308, Revision 1-A, the limiting 314-T RTNDTvalue is 188.9"F for Surry Unit 1 lntermediate to Lower Shell Circumferential Welds SA-1585172445 and SA-1650172445.

In summary, when the revised 60-year fluence projections at 48 EFPY are used with the revised Linde 80 weld initial RTNDTvalues, the limiting 114-T and 314-T RTNDTvalues remain less than those used in the development of the current PIT limits and LTOPS setpoints. Additionally, when revised fluence projections at 32 EFPY are used without the revised Linde 80 weld material properties, the limiting 114-T and 314-T RTNDTvalues remain less than those used in the development of the current P/T limits and LTOPS setpoint. Therefore, the existing Surry Technical Specifications RCS P/T limits, LTOPS setpoint, and LTOPS enabling temperature remain valid and conservative for their period of applicability, corresponding to 28.8 EFPY and 29.4 EFPY for Surry Units 1 and 2, respectively.

3.6 E ' V Material Surveillance Proaram Der 10 CFR 50 Amendix H Current reactor vessel material surveillance monitoring requirements for Surry are based on1 the predicted shift in Charpy V-notch 30 ft-lb energy (AT30). The alternate methodology described in Topical Report BAW-2308, Revision 1-A does not rely on obtaining direct fracture toughness measurements (i.e. in accordance with ASTM-1921) in the irradiated condition for the purposes of monitoring changes due to irradiation in the Linde 80 weld materials. Topical Report BAW-2308, Revision I - A also confirmed Page 8 of 11

Serial No.06-434 Docket Nos. 50-280, 281 Attachment 1 that the irradiation-induced shift in Charpy V-notch 30 ft-lb energy (AT30)conservatively overpredicted the Master Curve AT0 test data for Linde 80 weld materials. Therefore, the current reactor vessel material surveillance program at Surry Power Station is not affected (i.e. current monitoring requirements are based on predicted shift in Charpy V-notch 30 ft-lb energy (ATSO).

Dominion expects to submit for NRC review and approval a revised surveillance capsule withdrawal schedule for Surry, which will be valid for the current 60-year operating license period. The revised surveillance capsule withdrawal schedule will incorporate the guidance of NUREG-1801 (GALL Report).

4.0 Surry Units 1 and 2 Technical Specifications As notecl in Section 3.5, the current Surry Technical Specifications RCS P/T limits (TS 3.1 .E3), LTOPS setpoint, and LTOPS enabling temperature (TS 3.1 .G.l .c) are not affected by this submittal and remain valid and conservative for their period of applicability. Therefore, there are no changes to Surry Technical Specifications proposed in this submittal. However, Dominion expects to submit, at a later date, a Technical Specifications change request to provide revised RCS PTT Limits, LTOPS setpoint, and LTOPS Enable Temperature basis that will be effective through the end of the Surry Units 1 and 2 60-year operating licenses.

5.0 Affected UFSAR Sections Sections of the Surry Units 1 and 2 UFSAR will be revised to reflect implementation of the revised design basis analyses described herein. Following NRC approval of the exemption request associated with this submittal, a UFSAR revision will be made in accordance with the requirements of 10 CFR 50.71 (e).

6.0 Reactor Vessel Integrity Database (RVID)

Attachment 3 of this submittal contains a RVlD update based on the alternate material properties basis for Linde 80 weld materials as provided in Topical Report BAW-2308, Revision I-A. The RVID update also includes the revised RPV neutron fluence values, which include the effects of FSI removal and IFBA core designs. The revised fluence values support plant operation for 48 EFPY for Surry Units 1 and 2, corresponding to the end olf the current 60-year operating licenses.

7.0 Conclusions The proposed changes to the Surry Power Station Units 1 and 2 PTS assessment per 10 CFR 50.61 are valid to the end of the current 60-year license period. As described in this assessment, the PTS screening criteria per 10 CFR 50.61 are met for Surry Units 1 and 2 FlPV beltline materials. The assessment employs alternate initial R T N ~ ~

methodology for the Linde 80 weld materials as described in approved Topical Report BAW-23O8, Revision 1-A. The assessment also employs revised RPV neutron fluence Page 9 of 11

Serial No.06-434 Docket Nos. 50-280, 281 Attachment 1 values that appropriately consider l FBA core designs and FSI removal.

The Upper Shelf Energy values for Surry Units 1 and 2 beltline materials meet the 50 ft-lb acceptance criterion of 10 CFR 50 Appendix G at the end of the current 60-year operating license period, or an EMA demonstrates acceptable margins of safety against fracture for projected low upper-shelf Charpy impact energy levels at 48 EFPY.

The 114 and 314-T RTNDTvalues used in the development of the current Surry Units 1 and 2 Technical Specifications PfT limits, LTOPS setpoint, and LTOPS enabling temperature remain valid and conservative for the period of applicability.

References Framatome ANP Topical Report BAW-2308, Revision I-A, "Initial RTNDTof Linde 80 Weld Materials," approved August 2005.

NRC Letter from H. N. Berkow to J. S. Holm (Framatome ANP), "Final Safety Evaluation for Topical Report BAW-2308, Revision I,'Initial RTNDTof Linde 80 Weld Materials' (TAC No. MB6636)," dated August 4, 2005.

Letter from L. N. Hartz (Virginia Electric and Power Company) to USNRC, "Virginia Electric and Power Company, Surry Power Station Unit 2, Evaluation of Surry 2 Capsule Y Data," dated March 27, 2003.

Letter from W. L. Stewart (Virginia Electric and Power Company) to USNRC, "Virginia Electric and Power Company, Surry Power Station Units 1 and 2, North Anna Power Station Units 1 and 2, Revision to 10 CFR 50.61 Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events," dated December 10, 1991.

NRC Regulatory Guide 1.190, Revision 0, "Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence," dated March 2001.

Framatome ANP Report BAW-2241P-A, Rev. 1, "Fluence and Uncertainty M~!thodologies,"December 1999.

NFC Letter from S. A. Richards to J. J. Kelly (B&WOG), "Acceptance for Referencing of Licensing Topical Report BAW-2241P, Revision 1, Fluence and Uncertainty Methodologies (TAC No. M98962)," dated April 5,2000.

Framatome ANP Report BAW-2494, Revision 1, "Low Upper-Shelf Toughness Fracture Mechanics Analysis of Reactor Vessel of Surry Units 1 and 2 for Extended Life through 48 Effective Full Power Years," September 2005.

Page 10of 11

Serial No.06-434 Docket Nos. 50-280, 281 Attachment 1

9. Framatome Report BAW-2323, "Low Upper-Shelf Toughness Fracture Mechanics Analysis of Reactor Vessels of Surry Units 1 and 2 for Extended Life through 48 Effective Full Power Years," June 1998.

10.Letter from R. F. Saunders (Virginia Electric and Power Company) to USNRC, "Virginia Electric and Power Company, Surry Power Station Units 1 and 2, Request for E.xemption - ASME Code Case N-514, Proposed Technical Specifications Change, Revised Pressurefremperature Limits and LTOPS Setpoint,"

Serial No.95-197, June 8, 1995.

11. Letter from USNRC to J. P. O'Hanlon (Virginia Electric and Power Company), "Surry Units 1 and 2 - Issuance of Amendments Re: Surry, Units 1 and 2 Reactor Vessel Heatup and Cooldown Curves (TAC NOS. M92537 and M92538)," dated December 28,1995.

12. NRC I'legulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials," dated May 1988.

Page 11 of 11

Serial No.06-434 Docket Nos. 50-280,281 Attachment 2 Reaulatorv Basis and Reauest for Exem~tion Surry Power Station Units 1 and 2 Virginia Electric and Power Company (Dominion)

Serial No.06-434 Docket Nos. 50-280, 281 Attachment 2 Reaulatorv Basis and Request for Exem~tion 1.0 Introduction In accordance with the provisions of 10 CFR 50.60(b) and 10 CFR 50.12, Virginia Electric and Power Company (Dominion) is submitting a request for exemption from certain requirements of 10 CFR 50.61, "Fracture Toughness Requirements for Protection Against Thermal Shock Events," and 10 CFR 50, Appendix G, "Fracture Toughness Requirements." The requested exemption would allow use of an alternate method, as described in Framatome ANP Topical Report BAW-2308, Revision 1-A, for determining the adjusted RTNDT(reference nil-ductility temperature) of the Linde 80 weld materiala present in the beltline region of the Surry Units 1 and 2 reactor pressure vessels.

10 CFR 50.61 (a)@) and 10 CFR 50, Appendix G (II)(D)(i), require that the pre-service or unirradiated condition RTNDTbe evaluated according to the procedures in the ASME Code, Section Ill, Paragraph NB-2331, from Charpy V-notch impact tests and drop weight tests.

Framatorne ANP Topical Report BAW-2308, Rev. I - A provides an NRC-approved alternate method for determining the adjusted R T N(reference

~ ~ nil-ductility temperature) of the Linde 80 weld materials present in the beltline region of the reactor pressure vessels at Surry Power Stations Unit 1 and 2. BAW-2308, Revision 1-A, also provides revised initial (unirradiated) R T N Dvalues

~ for the Linde 80 weld materials present in the reactor pressure vessels of Surry Units 1 and 2.

The following Condition and Limitation is stated in the NRC's Safety Evaluation for Topical Report BAW-2308, Rev. 1:

'Any licensee who wants to utilize the methodology of TR BAW-2308, Revision I as: outlined in items (I) through (3) above, must request an exemption, per 10 CIrR 50.12, from the requirements of Appendix G to 10 CFR Part 50 and 10 CFR 50.61 to do so."

In the above quotation, Condition and Limitation (1) pertains to NRC-accepted values of initial (unirradiated) reference temperature, IRTTo, and the corresponding uncertainty term, 01, for Linde 80 weld materials based on the Master Curve methodology using direct testing of fracture toughness in accordance with ASTM Standard Test Method E-1921.

Condition and Limitation (2) requires that a minimum chemistry factor of 167.0°F be applied when the methodology of Regulatory Guide 1.99, Revision 2, is used to assess the shift in nil-ductility transition temperature due to irradiation.

Page 1 of 4

Serial No.06-434 Docket Nos. 50-280, 281 Attachment 1 Condition and Limitation (3) requires that a value of OA = 28.0°F be used to determine the margin term, as defined in Topical Report BAW-2308, Revision 1-A, and Regulatory Guide 1.99, Revision 2.

The exemption requested by Dominion addresses portions of the following regulations:

(1) Appendix G to 10 CFR Part 50, which sets forth fracture toughness requirements for ferritic materials of pressure-retaining components of the reactor coolant pressure boundary of light water nuclear power reactors to provide adequate margins of safety during any condition of normal operation, including anticipated operational occurrences and system hydrostatic tests, to which the system may be subjected over its service lifetime; (2) 10 CFR 50.61, which sets forth fracture toughness requirements for protection against pressurized thermal shock (PTS).

The exemption from Appendix G to 10 CFR 50 is to replace the required use of the existing Charpy V-notch and drop-weight-based methodology with the use of an alternate methodology that incorporates the use of fracture toughness test data for evaluating the integrity of the Linde 80 weld materials present in the Surry Units 1 and 2 reactor pressure vessel beltline regions. The alternate methodology employs direct fracture toughness testing per the Master Curve methodology based on use of ASTM Standard Method E 1921 (1997 and 2002 editions), and ASME Code Case N-629. The exemption is required since Appendix G to 10 CFR 50 requires that for the pre-service or unirracliated condition, RTNDTbe evaluated by Charpy V-notch impact tests and drop weight tests according to the procedures in the ASME Code, Paragraph NB-2331.

The exemption from 10 CFR 50.61 is to use an alternate methodology to allow the use of direct fracture toughness test data for evaluating the integrity of the Linde 80 weld materials present in the Surry Units 1 and 2 RPV beltline regions, based on the use of ASTM E 1921 (1997 and 2002 editions) and ASME Code Case N-629. The exemption is required because the methodology for evaluating RPV material fracture toughness in 10 CFR 50.61 requires that the pre-service or unirradiated condition be evaluated using Charpy V-notch impact tests and drop weight tests according to the procedures in the ASM E Code, Paragraph NB-2331.

Additionally, the NRC's Safety Evaluation for Topical Report BAW-2308, Revision 1, conclude!; that an exemption is required to address issues related to 10 CFR 50.61 inasmuch as the methodology presented in Topical Report BAW-2308, Revision 1, as modified and approved by the NRC staff, represents a significant change to the methodollogy specified in 10 CFR 50.61 for determining the PTS reference temperature (RTPTs)value for Linde 80 weld material. The changes in the methodology described in BAW-2308, Revision 1-A, with respect to the methodology per 10 CFR 50.61, include Page 2 of 4

Serial No.06-434 Docket Nos. 50-280,281 Attachment 1 the requirements for use of a minimum chemistry factor of 167°F and a value of an = 28.0°F for Linde 80 weld materials.

10 CFR !50.12 states that the Commission may grant an exemption from requirements contained in 10 CFR 50 provided that: 1) the exemption is authorized by law, 2) the exemption will not result in an undue risk to public health and safety, 3) the exemption is consister~twith the common defense and security, and 4) special circumstances, as defined in 10 CFR 50.1 2(a)(2) are present. The requested exemption to allow the use of Topica.1 Report BAW-2308, Revision I-A, as the basis for the Linde 80 weld material initial properties at Surry Units 1 and 2 satisfy these requirements as described below.

1. The roauested exem~tionis authorized bv law.

No law exists which precludes the activities covered by this exemption request.

10 CFR 50.60(b) allows the use of alternatives to 10 CFR 50, Appendix G when an exemption is granted by the Commission under 10 CFR 50.1 2.

In addition, 10 CFR 50.61 permits other methods for use in determining the initial material properties provided such methods are approved by the Director, Office of Nuclear Reactor Regulation.

2. The reauested exem~tiondoes not Dresent an undue risk to the ~ u b l i chealth and safetv.

The proposed material initial properties basis described in Topical Report BAW-2308 Revisiion 1-A represents an NRC-approved methodology for establishing weld wire specific and generic lRTTo values for Linde 80 welds. Topical Report BA-2308, Revisiion 1-A, includes appropriate conservatisms to ensure that use of the proposed initial material properties basis does not increase the probability of occurrence or the consequences of an accident at Surry Unit 1 or 2, and will not create the possibility for a new or different type of accident that could pose a risk to public health and safety.

The use of this proposed approach ensures that the intent of the requirements specified in 10 CFR 50 Appendix G and 10 CFR 50.61 are satisfied.

The requested exemption is consistent with the NRC staff requirements specified in the Safety Evaluation for the approved Topical Report BAW-2308, Revision I-A; consequently, the exemption does not present an undue risk to the public health and safety.

3. The roauested exem~tionwill not endanaer the common defense and securitv.

The requested exemption is specifically concerned with RPV material properties and is corisistent with NRC staff requirements specified in the Safety Evaluation for Page 3 of 4

Serial No.06-434 Docket Nos. 50-280, 281 Attachment 1 approved Topical Report BAW-2308, Revision I-A. Consequently, the requested exemption will not endanger the common defense and security.

4. SDeciid circumstances are Dresent which necessitate the reauest for an exem~tion to the reaulations of 10 CFR 50.61 and 10 CFR 50 Amendix G.

Pursuant to 10 CFR 50.12(a)(2), the NRC will not consider granting an exemption to the regulations unless special circumstances are present. The requested exemption meets. the special circumstances of paragraph 10 CFR 50.1 2(a)(2)(ii) since application of these regulations in this particular circumstance is not necessary to achieve the underlying purpose of the regulations.

The underlying purpose of 10 CFR 50.61 and 10 CFR 50 Appendix G is to protect the integrity of the reactor coolant pressure boundary by ensuring that each reactor vessel material has adequate fracture toughness. Application of paragraph NB-2331 of ASME Section Ill in the determination of initial material properties was conservatively developed based on the level of knowledge existing in the early 1970s concerning reactor pressure vessel materials and the estimated effects of operation. Since the early 1970s, the level of knowledge concerning these topics has greatly expanded. This increased knowledge level permits relaxation of the ASME Ill NB-2331 requirements via application of Topical Report BAW-2308, Revision I-A, while maintaining the underlying purpose of the ASME Code and NRC regulations to ensure an acceptable margin of safety is maintained.

This submittal presents the reactor vessel integrity assessments for Surry Power Station Units 1 and 2 utilizing the methodology of Topical Report BAW-2308, Revision 1-A for Linde 80 weld materials. The assessment documents the integrity of the reactor pressure vessel for Surry Units 1 and 2 relative to the requirements and u~iderlyingpurpose of 10 CFR 50.61 and 10 CFR 50 Appendix G.

Therefore, the intent of 10 CFR 50.61 and 10 CFR 50 Appendix G will continue to be satisfiled for the proposed change in reactor vessel material initial properties basis, thus justifying the exemption request. Issuance of an exemption from the criteria of these regulations to permit the use of Topical Report BAW-2308, Revision I - A for Surry Units 1 and 2 will not compromise the safe operation of the reactors, and will ensure that RPV integrity is maintained.

Page 4 of 4

Serial No.06-434 Docket Nos. 50-280. 281 Attachment 3 Reactor Vessel Materials Data Tables Surry Power Station Units 1 and 2 Virginia Electric and Power Company (Dominion)

Serial No.06-434 Docket Nos. 50-280,281 Attachment 3 PTS Summary Based on 60-Year Fluences and Linde 80 Weld Material Properties per BAW-2308, Rev. 1 Surry Unit 1

.Facility:

vessei ivianuiaciurer: B&'w and Fioiieraam Docicyara

+

114-T ART value of 228 4 F was used in the determination of PiT llmits (Approved by NRC on 12/28/95)

" 314-T ART value of 189.5 F was used in the determlnatlon of PiT llmits (Approved by NRC on 12/28/95)

Note- Shaded cells indlcate a changed value relat~veto Dominion's must recent update to the NRC's Reactw Vessel Integr~tyDatabase (RVID) (Last Update on 3/27/03)

Facility: Surry Unit 2 Vessel Manufacturer: B&W and Rotterdam Dockyard 114-T ART value of 228 4 F was used in the determ~nattonof PK h t s (Approved by NRC on 12/28/95)

" 314-T ART value of 189 5 F was used in the determ~nationof P/T llmlts (Approved by NRC on 12/28/95)

Note: Shaded cells ind~cateachangedvalue relat~veto Domin~on'smost recent update to the NRC's Reactw Vessel Integrity Database (RVID) (Last Update on 3/27/03).

Page 1 of 2

Serial No.06-434 Docket Nos. 50-280, 281 Attachment 3 Upper Shelf Energy Summary Based on 60-Year Fluences CvUSE Values Facility: Surry Unit 1 Vessel Manufacturer: B&W and Rotterdam Dockyard RPV Weld Wlre Heat or Material ID 122V109VA1 C4326-1 Location Nozzle Shell Forg~ng lntermed~ateShell Forg~noor FluxType SA508, CI 2 SA533. Gr. 6 1 USE O 1 / 4 T 69.2 84.4 lld-T

( X IE19) 0.473 3.453

.F .l ~ ~. o n r e Un~rrad~ated 83.0 115.0 U S E Untrrad~atedUSE Metho, MeasuredIMTEB 5-2 Measured

%Drop in USE O EOL O 114 I I

%Drop ~n C4326-2 Intermediate Shell SA533. Gr. 6 1 68.9 3.453 94.0 MeasuredIMTEB 5-2 4415-1 Lower Shell SA533, Gr. 6 1 I 76.6 3.453 103.0 Measured 441 5-2 Lower Shell SA533, Gr. 6 1 60.9 3.453 83 0 MeasuredlMTEB 5-2 J726R5017 Nozzle to Int Shell Circ Weld SAF 89 EMA 0.473 EM A Est~rnate SA-I585172445 Int to Low Sh Circ (ID 40%) Llnde 80 EMA 3.423 EMA Measured SA-I650172445 Int to Low Sh. Clrc (OD 60%) Llnde 80 EMA 3.423 EMA Measured SA-149418T1554 Int Shell Long Welds L3 a L4 Linde 80 EMA 0.659 EMA Estimate SA-149418T 1554 Lower Shell Long. Weld L1 Linde 80 EMA 0.634 EMA Estimate SA-15261299L44 Lower Shell Long. Weld L2 Linde 80 EMA 0.634 EMA Measured Note Shaded cells indicate a changed value relative to Dom~nion'smost recent update to the NRC's Reactor Vessel Integrity Database (RVID) (Last Update on 3/27/03)

CvUSE Values Facility: Surry Unit 2 Vessel Manufacturer: B & W and Rotterdam Dockyard Note Shaded cells indicate a changed value relative to Dominion's most recent update to the NRC's Reactor Vessel Integrity Database (RVID) (Last Update on 3/27/03)

Page 2 of 2

Serial No.06-434 Docket Nos. 50-280,281 Attachment 4 Framatome ANP R e ~ o r BAW-2494.

t Revision 1 Low Up~er-ShelfTouahness Fracture Mechanics Analvsis of Reactor Vessel of Surq Units 1 and 2 for Extended Life throuah 48 Effective Full Power Years Se~tember2005 Surry Power Station Units 1 and 2 Virginia Electric and Power Company (Dominion)

BAW-2494, Revision I September 2005 Low Upper-Shelf Toughness Fracture Mechanics Analysis of Reactor Vessel of Surry Units 1 and 2 for Esxtended Life through 48 Effective Full Power Years AREVA Document No. 77-2494-01 (See Section 1 1 for document signatures.)

Prepared for Dominion Generation Prepared by Framatome ANP, Inc.

An AREVA and Siemens company 3315 Old Forest Road P. 0. Box 10935 Lynchburg, Virginia 24506-0935

BAW-2494, REV. 1 EXECUTIVE

SUMMARY

Dominion Generation is considering withdrawal of the flux suppression inserts (FSls) from the core of SUI-ryUnit 1 in Cycle 21 and implementation of Integral Fuel Burnable Absorbers (IFBA) in the feed fuel for both Surry Units 1 and 2 in their respective Cycle 21. As a result of these changes, projected fluence values at 48 effective full power years (EFPY) of plant operation have changed. It must be ensured that these changes do not affect the plant adversely from a regulatory compliance point of view. One of the compliance issues is Appendix G to 10 CFR Part 50 where low upper-shelf toughness is addressed. An equivalent margins assessment has to be made for material toughness when the upper-shelf Charpy energy level falls below 50 ft-lb.

This report addresses this particular compliance issue regarding low upper-shelf toughness only.

The Charpy upper-shelf value of reactor vessel beltline weld materials at Surry Units 1 and 2 may be leas than 50 ft Ib at 48 EFPY. In order to demonstrate that sufficient margins of safety against fracture remain to satisfy the requirements of Appendix G to 10 CFR Part 50, a low upper-shelf toughness fracture mechanics analysis has been performed. The limiting welds in the beltline region have been evaluated for ASME Levels A, B, C, and D Service Loadings based on the evaluation acceptance criteria of the ASME Code,Section XI, Appendix K.

The analysis presented in this report demonstrates that the limiting reactor vessel beltline weld at Surry Units 1 and 2 satisfies the ASME Code requirements of Appendix K for ductile flaw extensions and tensile stability using projected low upper-shelf Charpy impact energy levels for the weld material at 48 EFPY.

AREVA

BAW-2494, REV. 1 RECORD OF REVISIONS Revision Affected Paqes Description Date 0 All Original release May 2005 1 All Include analysis of Surry Unit 2. September 2005 Re-evaluate Surry Unit 1 using updated fluence values.

iii h AREVA

BAW.2494. REV. 1 TABLE OF CONTENTS 1.0 Introduction ....................................................................................................................1-1 2.0 Acceptance criteria......................................................................................................... 2-1 2.1 Levels A and B Service Loadings (K-2200) ........................................................2-1 2.2 Level C Service Loadings (K-2300) .................................................................... 2-1 2.3 Level D Service Loadings (K-2400)....................................................................2-2 3.0 Material Properties and Reactor Vessel Design Data ....................................................3-1 3.1 J-Integral Resistance Model for Mn-Mo-NiILinde 80 Welds ...............................3-1 3.2 Reactor Vessel Design Data .............................................................................. 3-2 3.3 Mechanical Properties for Weld Material............................................................ 3-2 3.4 J-Integral Resistance for Linde 80 Weld Material............................................... 3-3 4.0 Analytical Methodology .................................................................................................. 4-1 4.1 Procedure for Evaluating Levels A and B Service Loadings ..............................4-1 4.2 Procedure for Evaluating Levels C and D Service Loadings ..............................4-3 4.3 Temperature Range for Upper-Shelf Fracture Toughness Evaluations .............4-5 4.4 Effect of Cladding Material ................................................................................. 4-5 5.0 Applied Loads ................................................................................................................ 5-1 5.1 Levels A and B Service Loadings....................................................................... 5-1 5.2 Levels C and D Service Loadings ...................................................................... 5-1 6.0 Evaluation for Levels A and B Service Loadings ...........................................................6-1 7.0 Evaluation for Levels C and D Service Loadings ........................................................... 7-1 8.0 Summary of Results .......................................................................................................8-1 10.0 References.................................................................................................................. 1 11.0 Certification ..................................................................................................................11-1 AREVA

BAW.2494. REV. 1 LIST OF TABLES Me'chanicalProperties for Base and SA-1526 Weld Material of Surry Unit 1 ................3-3 Selected Welds and Properties...................................................................................... 3-4 J-Integral Resistances for Levels A and B Service Loadings.........................................3-4 J-Integral Resistances for Levels C and D Service Loadings ........................................3-5 Flaw Evaluation for Levels A and B Service Loadings...................................................6-2 J-Integral vs. Flaw Extension for Levels A and B Service Loadings . SA-1526 ............6.3 J-F;ICurves for Evaluation of Levels A and B Service Loadings.....................................6-4 K/ 11s. Crack Tip Temperature for SLB ............................................................................ 7-3 KIcat '1, 0 Wall Thickness ................................................................................................ 7-4 KJ, at Wall Thickness with Aa = 0.10 in ....................................................................7-5 J-Integral vs . Flaw Extension for Levels C and D Service Loadings ..............................7-6 J-A! Curves for Evaluation of Levels C and D Service Loadings .................................... 7-7 Level D Service Loadings - Internal Pressure at Tensile Instability ...............................7-8 A

AREVA

BAW.2494. REV. 1 LIST OF FIGURES Reactor Vessel of Surry Unit 1.......................................................................................1-2 Reactor Vessel of Surry Unit 2 .......................................................................................1-3 Reactor Vessel Beltline Region with Postulated Longitudinal Flaw ...............................2-3 Reactor Vessel Beltline Region with Postulated Circumferential Flaw .......................... 2-4 SLB transient - Reactor Coolant Temperature and Pressure vs . Time .........................5-2 J-Integral vs . Flaw Extension for Levels A and B Service Loadings .............................6-5 K, crs. Crack Tip Temperature for Levels C & D Service Loadings .................................7-9 J-Integral vs . Flaw Extension for Levels C & D Service Loadings................................7-10 Dt AREVA

BAW-2494, REV. 1 1.0 Introduction One consideration for extending the operational life of reactor vessels beyond their original licensing period is the degradation of upper shelf Charpy impact energy levels in reactor vessel materials due to neutron radiation. Appendix G to 10 CFR Part 50, "Domestic Licensing of Production and Utilization Facilities," states in Paragraph IV.A.1.a that, "Reactor vessel beltline materials must have Charpy upper shelf energy ... of no less than 75 ft Ib initially and must maintain C:harpy upper shelf energy throughout the life of the vessel of no less than 50 ft Ib, unless it is demonstrated in a manner approved by the Director, Office of Nuclear Reactor Regulation, that lower values of Charpy upper shelf energy will provide margins of safety against fracture equivalent to those required by Appendix G of Section XI of the ASME Code."

Materials vvith Charpy upper shelf energy below 50 ft Ib are said to have low upper shelf fracture toughness. Fracture mechanics analysis is necessary to satisfy the requirements of Appendix G to 10 CFR Part 50 for reactor vessel materials with upper shelf Charpy impact energy levels that have dropped, or that are predicted to drop, below the 50 ft Ib requirement.

Dominion generation is considering withdrawal of the flux suppression inserts (FSls) from the core of Surry Unit 1 in Cycle 21 and implementation of Integral Fuel Burnable Absorbers (IFBA) in the feed fuel for both Surry Units 1 and 2 in their respective Cycle 21. This document assesses the effect of these proposed changes on the upper-shelf fracture toughness of the reactor vessels at Surry Units 1 and 2. The base metal and weld materials used in the beltline regions of the Surry reactor vessels are identified in Figures 1-1 and 1-2. The B&W Owners Group (B&.WOG)fracture toughness model was used in the low upper-shelf toughness fracture mechanics; analyses of the reactor vessels of the B&WOG Reactor Vessel Working Group

( R W G ) which includes the Surry Units 1 and 2 reactor vessels. The low upper-shelf toughness analysis for all reactor vessels of the B&WOG R W G for Levels A & B Service Loadings was documented in BAW-2192PA [I]. An additional fracture mechanics analysis for Levels C 8; D Service Loadings was carried out for all these reactor vessels and documented in BAW-217ElPA [2]. Both these reports have been accepted by the NRC. An additional low upper-shelf toughness analysis covering end-of-license and end-of-license renewal fluence values was performed in 1998 for Surry Units Iand 2 [3]. For the current planned changes, the effect on the Surry Units 1 and 2 reactor vessel materials' upper-shelf toughness is assessed in this report.

The present analysis addresses ASME Levels A, B, C, and D Service Loadings. For Levels A and B Service Loadings, the low upper-shelf toughness analysis is performed according to the acceptance criteria and evaluation procedures contained in Appendix K to Section XI of the ASME Code [4]. The evaluation also utilizes the acceptance criteria and evaluation procedures prescribed in Appendix K for Levels C and D Service Loadings. Levels C and D Service Loadings are evaluated using the one-dimensional, finite element, thermal and stress models and linear elastic fracture mechanics methodology of Framatome ANP's PCRlT computer code to determine stress intensity factors for a worst case pressurized thermal shock transient.

AREVA

BAW-2494, REV. 1 Figure 1-1 Reactor Vessel of Surry Unit 1 J726 (Rotterdam) Weld Weld S A - 1494 Intermediate Shell (Plate) C4326- 1 &

C4326-2 Weld SA-1585 Inside 40%

SA- 1650 Outside 60%

Weld SA-1494 Weld SA-1526 Lower Shell (Plate) C4415-1 &

C4415-2 A

AREVA

BAW-2494, REV. 1 Figure 1-2 Reactor Vessel of Surry Unit 2 L737 (Rotterdam) Weld Weld SA- 1585 Weld SA- 1585 Inside 50%

WF-4 Outside 50%

Intermediate Shell (Plate) C4208-2 &

R3008 (Rotterdam) Weld C4339- 1 Weld WF-4

+-

Weld WF-4 Inside 63%

WF-8 Outside 37%

Lower Shell (Plate) C433 1-2 8 C4339-2 A

AREVA

BAW-2494, REV. 1 2.0 Acceptance Criteria Acceptance criteria for the assessment of reactor vessels with low upper shelf Charpy impact energy levels are prescribed in Article K-2000 of Appendix K to the ASME Code,Section XI [4].

These criteria are summarized below as they pertain to the evaluation of reactor vessel weld metals.

2.1 Levels A and B Service Loadings (K-2200)

(a) When evaluating adequacy of the upper shelf toughness for the weld material for Levels A and B Service Loadings, an interior semi-elliptical surface flaw with a depth % of the wall thickness and a length six times the depth shall be postulated, with the flaw's major axis oriented along the weld of concern and the flaw plane oriented in the radial direction. Two criteria shall be satisfied:

(1) The applied J-integral evaluated at a pressure 1.15 times the accumulation pressure (Pa)as defined in the plant specific Overpressure Protection Report, with a factor of safety of 1.O on thermal loading for the plant specific heatup and cooldown conditions, shall be less than the J-integral of the material at a ductile flaw extension of 0.10 in.

(2) Flaw extensions at pressures up to 1.25 times the accumulation pressure (Pa)shall be ductile and stable, using a factor of safety of 1.0 on thermal loading for the plant specific heatup and cooldown conditions.

(b) The J-integral resistance versus flaw extension curve shall be a conservative representation for the vessel material under evaluation.

2.2 Level C Service Loadings (K-2300)

(a) When evaluating the adequacy of the upper shelf toughness for the weld material for Level C Service Loadings, interior semi-elliptical surface flaws with depths up to 'lI0 of the base metal wall thickness, plus the cladding thickness, with total depths not exceeding 1.0 in., and a surface length six times the depth, shall be postulated, with the flaw's major axis oriented along the weld of concern, and the flaw plane oriented in the radial direction. Flaws of various depths, ranging up to the maximum postulated depth, shall be analyzed to determine the most limiting flaw depth. Two criteria shall be satisfied:

(1) The applied J-integral shall be less than the J-integral of the material at a ductile flaw extension of 0.10 in., using a factor of safety of 1.0 on loading.

(2) Flaw extensions shall be ductile and stable, using a factor of safety of 1.0 on loading.

(b) The J-integral resistance versus flaw extension curve shall be a conservative representation for the vessel material under evaluation.

AREVA

BAW-2494, REV. 1 2.3 Level D Service Loadings (K-2400)

(a) When evaluating adequacy of the upper shelf toughness for Level D Service Loadings, flaws as specified for Level C Service Loadings shall be postulated, and toughness properties for the corresponding orientation shall be used. Flaws of various depths, ranging up to the maximum postulated depth, shall be analyzed to determine the most limiting flaw depth. Smaller maximum flaw sizes may be used when justified. Flaw extensions shall be ductile and stable, using a factor of safety of 1.0 on loading.

(b) The J-integral resistance versus flaw extension curve shall be a best estimate representation for the vessel material under evaluation.

(c) The extent of stable flaw extension shall be less than or equal to 75% of the vessel wall thickness, and the remaining ligament shall not be subject to tensile instability.

A AREVA

BAW-2494, REV. 1 Figure 2-1 Reactor Vessel Beltline Region with Postulated Longitudinal Flaw Semi-Elliptical Flaw (Not to Scale)

A AREVA

BAW-2494, REV. 1 Figure 2-2 Reactor Vessel Beltline Region with Postulated Circumferential Flaw Semi-Elliptical Flaw (Not to Scale) h AREVA

BAW-2494, REV. 1 3.0 Material Properties and Reactor Vessel Design Data An upper-shelf fracture toughness material model is discussed below, as well as mechanical properties for the weld material and reactor vessel design data.

3.1 J-Integral Resistance Model for Mn-Mo-NiILinde 80 Welds A model lor the J-integral resistance versus crack extension curve (J-R curve) required to analyze low upper-shelf energy materials has been derived specifically for Mn-Mo-NilLinde 80 weld materials. A previous analysis of the reactor vessels of B&W Owners Group RVWG [I]

described the development of this toughness model from a large database of fracture specimens. Using a modified power law to represent the J-R curve, the mean value of the J-integral is given by:

with In(Cl)=al + a , C u ( ( ~ ~ +a3

) ~ ~ T + a 4 In(B,)

C, = dl + d2In(C,) + d3In(B,)

C3 = d4+ d5 In(C,) + d, In(BN)

C4 = -0.4489 where Aa = crack extension, in.

Cu = copper content, wt%

& = fluence at crack tip, 1018 n/cm2 T = temperature, O F BN = specimen net thickness = 0.8 in.

and A

AREVA

BAW-2494, REV. 1 A lower bound (-2S,) J-R curve is obtained by multiplying J-integrals from the mean J-R curve by 0.699 ['I]. It was shown in a previous low upper-shelf fracture toughness analysis performed for B&W Owners Group plants [5] that a typical lower bound J-R curve is a conservative representation of toughness values for reactor vessel beltline materials, as required by Appendix K [4] for Levels A, B, and C Service Loadings. The best estimate representation of toughness required for Level D Service Loadings is provided by the mean J-R curve.

3.2 Reactor Vessel Design Data Pertinent design data for upper-shelf flaw evaluations in the beltline region of the reactor vessel are provided below for Surry Units 1 and 2.

Design Pressure, Pd = 2485 psig (use 2500 psig)

Inside radius, Ri = 78.95 in.

Vessel thickness, t = 8.08 in.

Nominal cladding thickness, t, = 0.16 in.

Reactor coolant inlet temperature, Tc = 543°F 3.3 Mechanical Properties for Weld Material The beltlin~eregion weld SA-1526 has been previously determined [3] to be the limiting weld for the reactor vessels at Surry. Mechanical properties for the base and SA-1526 weld materials are presented in Tables 3-1. Base metal properties are found in the ASME Code [6]. Weld metal tensile properties are taken from surveillance capsule data for the SA-1526 weld material tested at 70°F and 550°F at a fluence level of 1.60 x 10 n/cm2 [7]. Properties for the intermedia~tetemperatures are calculated by determining the relationship between the variation in yield strength of the base metal with temperature and applying a scaling factor based on the given yield strength values of the weld at the tested temperatures. Also, Poisson's ratio, v, is taken to be 0.3.

Reactor vessel base metal: SA-533, Grade B, Class 1 low alloy steel plate [8]

Description:

Mn-112Mo-1/2Ni Carbon content: c 0.30%

Description of weld material Weld wire: Mn-Mo-Ni Weld flux: Linde 80, SAF 89 Note: Although the J-R upper-shelf fracture toughness model was developed specifically for Linde 80 weld material, it is assumed that this material model may be used for all beltline welds, including the Rotterdam J726, L737 and R3008 weld materials A

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BAW-2494, REV. I Table 3-1 Mechanical Properties for Base and SA-1526 Weld Material of Surry Unit 1 Temp. Yield Strength (0,) Ultimate Strength (o,)*

Material:: Base Base Weld Base Weld Base Metal Metal SA- 1526 Metal SA-1526 Metal Source: Code Code Actual Code Actual Code

[Ref.]

- 161 161 [71 161 [71 [GI (OF) (ksi) (ksi) (ksi) (ksi) (ksi) (in/inI0F) 7.06E-06 7.06E-06 7.25E-06 7.43E-06 7.58E-06 7.70E-06 L1 7.76E-06 7.77E-06 7.83E-06

• Note: The ultimate strength values I of the base I and weld n calculations The ASMEi transition region fracture toughness curve for K,,, used to define the beginning of the upper-shelf toughness region, is indexed by the initial RTNDTof the weld material.

Initial RTmT = -7.0°F [9]

Margin = 69S°F [9]

3.4 J-Integral Resistance for Linde 80 Weld Material Values of J-integral resistance from the upper-shelf toughness model of Section 3.1 are dependent on the temperature and fluence at the crack tip location, the copper content of the weld material, and the size (thickness) of the fracture specimen. These parameters are listed below for the reactor vessels at Surry.

Crack tip temperature varies with plant operation. At 100% power normal operating conditions, the temperature at the crack tip, T, is taken to be the inlet temperature, or Crack tip temperature, T = Tc = 543°F Fluence aft the crack tip is determined using the attenuation equation from Regulatory Guide 1.99, Rev. 2 [ I 01:

A AREVA

BAW-2494,REV. 1 where

= attenuated fluence at crack tip, n/cm2 IS = fluence at inside surface, n/cm2 x = depth into the vessel wall, in.

Table 3-2lists the copper content of the weld materials and the fluence at the inside surface of the reactor vessel for all welds located within the innermost 40% of the beltline wall.

Table 3-2 Selected Welds and Properties Plant SA-1526 1 Longitudinal 0.34 12.48x 10' L737 1 Circumferential I 0.35 12.64x 1018 SA-1585 Longitudinal 0.22 13.68x 10" Surry 2 R3008 Circumferential 0.19 59.07 x 1018 WF-4 Longitudinal 0.19 13.68x 10" ence values are calculated in Ref. 1 1 (Surry Unit 1) and Ref. 12 (Surry Unit 2 These values are increased by a percentage value specified in Ref. 13 to account for future chanlges as requested by Dominion Generation.

Tables 3-:3and 3-4 provide mean and lower bound J-Integral Resistance, Jo.l, of the weld material at a ductile flaw extension of 0.10in. This data is provided for the beltline region weld locations at Surry Units 1 and 2, based on the following postulated flaw depths.

Service Flaw Depth Extension Total Depth Loading a Aa x=a+Aa Condition (in.) (in.) (in.)

Level A&B tf4 = 2.02 0.1 2.12 Level C&D tll0 = 0.808 0.1 0.908 AREVA

BAW-2494, REV. 1 Table 3-3 J-Integral Resistances for Levels A and B Service Loadings SA- 1585 L 8.22 934 653 Surry 2

~ R3008 WF-4 C

L 35.51 8.22 924 973 646 680 Table 3-4 J-Integral Resistances for Levels C and D Service Loadings C Plant Weld ID Weld Orientation Fluence at Extended Crack Depth (x 10" n/cm2)

Mean J0.l (Iblin)

Lower Bound J0.l (Iblin)

A SA-1585 L 11.OO 923 645 Surry 2 R3008 C 47.50 913 638 h

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BAW-2494, REV. 1 4.0 Analytical Methodology Upper-shelf toughness is evaluated through use of fracture mechanics analytical methods that utilize the acceptance criteria and evaluation procedures of Section XI, Appendix K [4], where applicable.

4.1 Procedure for Evaluating Levels A and B Service Loadings The applied J-integral is calculated per Appendix K, paragraph K-4210 [4], using an effective flaw depth1 to account for small scale yielding at the crack tip, and evaluated per K-4220 for upper-shelf toughness and per K-4310 for flaw stability, as outlined below.

(1) For an axial flaw of depth a, the stress intensity factor due to internal pressure is calculated with a safety factor (SF) on pressure using the following:

where (2:) For a circumferential flaw of depth a, the stress intensity factor due to internal pressure is calculated with a safety factor (SF) on pressure using the following:

where (3) For an axial or circumferential flaw of depth a, the stress intensity factor due to radial thermal gradients is calculated using the following:

0 I(CR) S 100°Flhour where CR = cooldown rate ("Flhr), and AREVA

BAW-2494, REV. 1 (4) The effective flaw depth for small scale yielding, a,, is calculated using the following:

(5) For an axial flaw of depth a,, the stress intensity factor due to internal pressure for small scale yielding is calculated with a safety factor (SF) on pressure using the following:

where (6) For a circumferential flaw of depth a,, the stress intensity factor due to internal pressure is calculated with a safety factor (SF) on pressure using the following where (7) For an axial or circumferential flaw of depth a,, the stress intensity factor due to radial thermal gradients is 0 I(CR) 1 100°Flhour where CR = cooldown rate ("Flhr), and (8) The J-integral due to applied loads for small scale yielding is calculated using the following:

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BAW-2494, REV. 1 where (9) Evaluation of upper-shelf toughness at a flaw extension of 0.10 in. is performed for a flaw depth, a = 0.23 + 0.10 in.,

using where Pa is the accumulation pressure for Levels A an(j B Service Loadings, such that where Jl = the applied J-integral for a safety factor of 1.15 on pressure, and a safety factor of 1.0 on thermal loading Jo.l= the J-integral resistance at a ductile flaw extension of 0.10 in.

(103 Evaluation of flaw stability is performed through use of a crack driving force diagram procedure by comparing the slopes of the applied J-integral curve and the J-R curve. The applied J-integral is calculated for a series of flaw depths corresponding to increasing amounts of ductile flaw extension. The applied pressure is the accumulation pressure for Levels A and B Service Loadings, Pa, and the safety factor (SF) on pressure is 1.25. Flaw stability at a given applied load is verified when the slope of the applied J-integral curve is less than the slope of the J-R curve at the point on the J-R curve where the two curves intersect.

4.2 Procedure for Evaluating Levels C and D Service Loadings Levels C and D Service Loadings are evaluated using the one-dimensional, finite element, thermal arid stress models and linear elastic fracture mechanics methodology of the PCRlT computer code to determine stress intensity factors. The beltline region weld identified in Section 3.:3is analyzed for the limiting Level D transient for Surry Units 1 and 2 which is a main steam line break (SLB) without offsite power transient. The pressurizer pressure is increased by 70 psi to account for the pressure difference between the pressurizer and the downcomer AREVA

BAW-2494, REV. 1 (i.e., the reactor vessel beltline region) during the time period of interest. This Level D transient is also useld to evaluate Level C Service Loadings since it bounds all Level C transients [2].

The transient considered appears in Figure 5.1. Transients are assumed to hold steady at the end of theiir definitions, and are held constant until the thermal gradient through the shell has developed fully and begins to dissipate.

The evaluation is performed as follows:

For each transient described above, utilize PCRlT to calculate stress intensity factors for a semi-elliptical flaw of depth 'I,,, of the base metal wall thickness, as a function of time, due to internal pressure and radial thermal gradients with a factor of safety of 1.0 on loading. The applied stress intensity factor, KI, calculated by PCRlT for each of these transients is compared to the KJ, limit of the weld. The transient that most closely approaches the KJ, limit is chosen as the limiting transient, and the critical time in the limiting transient occurs at the point where Kl most closely approaches the upper-shelf toughness curve.

At the critical transient time, develop a crack driving force diagram with the applied J-integral and J-R curves plotted as a function of flaw extension. The adequacy of the upper-shelf toughness is evaluated by comparing the applied J-integral with the J-R curve at a flaw extension of 0.10 in. Flaw stability is assessed by examining the slopes of the applied J-integral and J-R curves at the points of intersection.

Verify that the extent of stable flaw extension is no greater than 75% of the vessel wall thickness by determining when the applied J-integral curve intersects the mean J-R curve.

Verify that the remaining ligament is not subject to tensile instability. The internal pressure p shall be less than PI, where PI is the internal pressure at tensile instability of the remaining ligament. Equations for PI are given below for the axial and circumferential flaws [14]. These equations first appear in the 2001 Edition of the ASME Section XI code that is cited.

(a) For an axial flaw, where A

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BAW-2494, REV. 1 and

! = surface length of crack, six times the depth, a R,,, = mean radius of vessel This equation for PI includes the effect of pressure on the flaw face.

(b) For a circumferential flaw, This equation for PI includes the effect of pressure on the flaw face. This equation is valid for internal pressures not exceeding the pressure at tensile instability caused by the applied hoop stress acting over the nominal wall thickness of the vessel. This validity limit on pressure for the circumferential flaw equation for PI is 4.3 Temperature Range for Upper-Shelf Fracture Toughness Evaluations Upper-shelf fracture toughness is determined through use of Charpy V-notch impact energy versus temperature plots by noting the temperature above which the Charpy energy remains on a plateau, maintaining a relatively high constant energy level. Similarly, fracture toughness can be addresssed in three different regions on the temperature scale, i.e. a lower-shelf toughness region, a transition region, and an upper-shelf toughness region. Fracture toughness of reactor vessel steel and associated weld metals are conservatively predicted by the ASME initiation toughness curve, Klc, in the lower-shelf and transition regions. In the upper-shelf region, the upper-shellf toughness curve, KJ,, is derived from the upper-shelf J-integral resistance model described in Section 3.1. The upper-shelf toughness then becomes a function of fluence, copper content, temperature, and fracture specimen size. When upper-shelf toughness is plotted versus temperature, a plateau-like curve develops that decreases slightly with increasing temperatwe. Since the present analysis addresses the low upper-shelf toughness issue, only the upper-shelf temperature range, which begins at the intersection of K,, and the upper-shelf toughness curves, KJ,, is considered.

4.4 Eff~ectof Cladding Material The PCRIT code utilized in the flaw evaluations for Levels C and D Service Loadings does not consider stresses in the cladding when calculating stress intensity factors for thermal loads. To account for this cladding effect, an additional stress intensity factor, Klclad,is calculated separately and added to the total stress intensity factor computed by PCRIT.

The contribution of cladding stresses to stress intensity factor was examined previously [2]. In this low upper-shelf toughness analysis performed for B&W Owners Group Reactor Vessel AREVA

BAW-2494, REV. 1 Working Group plants, the Zion-I WF-70 weld using thermal loads from the Turkey Point SLB was deterrnined to be the bounding case. The Zion-I vessel was as thick as or thicker than any other vess,el. The thickness of the Surry reactor vessels is 8.08" whereas the Zion vessel is 8.44". Frolm a thermal stress perspective, it is conservative to consider the thicker vessel. For the Zion vessel, the maximum value of KIclad,at any time during the transient and for any flaw depth, was determined to be 9.0 ksidin. This bounding value is therefore used as the stress intensity factor for KIcladin this low upper-shelf toughness analysis.

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BAW-2494, REV. 1 5.0 Applied Loads The Levels A and B Service Loadings required by Appendix K are an accumulation pressure (internal prressure load) and a cooldown rate (thermal load). Since Levels C and D Service Loadings are not specified by the Code, Levels C and D pressurized thermal shock events are reviewed and a worst case transient is selected for use in flaw evaluations.

5.1 Levels A and B Service Loadings Per paragraph K-1300 of Appendix K [4], the accumulation pressure used for flaw evaluations should not: exceed 1.1 times the design pressure. Using 2.5 ksi as the design pressure, the accumulatiion pressure is 2.75 ksi. The cooldown rate is also taken to be the maximum required by Appendlix K, 1OO°F/hour.

5.2 Levels C and D Service Loadings As discussed in Section 4.2, the SLB transient is evaluated using the computer code PCRIT.

Pressure and temperature time histories for the SLB transient are shown in Figure 5-1. The pressurizer pressure is increased by 70 psi to account for the pressure difference between the pressurizer and the downcomer (i.e., the reactor vessel beltline region) during the time period of interest.

h AREVA

BAW-2494, REV. 1 Figure 5-1 SLB transient - Reactor Coolant Temperature and Pressure vs. Time 200 300 400 Transient Time (sec) 100 0 L 0 100 200 I

300 Transient Time (sec)

I 400 500 600 h

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BAW-2494, REV. 1 6.0 Evaluation for Levels A and B Service Loadings Initial flaw depths equal to % of the vessel wall thickness are analyzed for Levels A and B Service Loadings following the procedure outlined in Section 4.1 and evaluated for acceptance based on values for the J-integral resistance of the material from Section 3.4. The results of the evaluation are presented in Table 6-1, where it is seen that the minimum ratio of material J-integral resistance (Jo.,) to applied J-integral (J,) is 1.26 which is higher than the minimum acceptable value of 1.O.

The flaw evaluation for the controlling weld (SA-1526) is repeated by calculating applied J-integrals for various amounts of flaw extension with safety factors (on pressure) of 1.15 and 1.25 in Table 6-2. The results, along with mean and lower bound J-R curves developed in Table 6-3, are plotted in Figure 6-1. An evaluation line at a flaw extension 0.10 in. is also included to confirm the results of Table 6-1 by showing that the applied J-integral for a safety factor of 1.15 is less than the lower bound J-integral resistance of the material. The requirerne~ntfor ductile and stable crack growth is also demonstrated by Figure 6-1 since the slope of the applied J-integral curve for a safety factor of 1.25 is considerably less than the slope of the lower bound J-R curve at the point where the two curves intersect.

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Table 6-1 Flaw Evaluation for Levels A & B Service Loadings Dimensional data: Material data:

Ri = 78.95 in. T= 543 OF t= 8.08 in. E= 27042 ksi a, = 2.0200 in. v = 0.3 Aa = 0.1000 in. E' = 29716 ksi a= 2.1200 in.

aft = 0.2624 ( 0.2 1 alt 1 0.5 )

Loading data: Geometry factors for initial flaw depth (wlo plasticity correction):

Pd = 2.50 ksi Fl = 1.0513 for pressure loading and axial flaws Pa = 2.75 ksi FP= 0.9699 for pressure loading and circumferential flaws SF= 1.15 F3 = 1.O624 for thermal loading and both flaw types CR = 100 OF/hr Plant Weld Orient. KlP Kt SY ae adt F1' or F2' F3' KIP' Kt' J~ Jo.1 at ~4 Jo.1 IJ1 (ksidin) (ksidin) (ksi) (in.) (ksidin) (ksidin) (Iblin) (Iblin)

J726 C 46.59 19.72 85.30 2.1521 0.2663 0.9715 1.0622 47.02 19.71 150 556 3.71 SA-1494 L 92.41 19.72 85.30 2.2117 0.2737 1.0574 1.0613 94.94 19.70 442 710 1.61 Surry 1 SA-1585 C 46.59 19.72 85.30 2.1521 0.2663 0.9715 1.0622 47.02 19.71 150 613 4.09 SA-1526 L 92.41 19.72 85.30 2.2117 0.2737 1.0574 1.0613 94.94 19.70 442 556 1.26 L737 C 46.59 19.72 85.30 2.1521 0.2663 0.9715 1.0622 47.02 19.71 150 548 3.65 SA-1585 L 92.41 19.72 85.30 2.2117 0.2737 1.0574 1.0613 94.94 19.70 442 653 1.48 Surry 2 R3008 C 46.59 19.72 85.30 2.1521 0.2663 0.9715 1.0622 47.02 19.71 150 646 4.31 W F-4 L 92.41 19.72 85.30 2.2117 0.2737 1.0574 1.0613 94.94 19.70 442 680 1.54 10 P

(0 P

Table 6-2 J-Integral versus Flaw Extension for Levels A & B Service Loadings - SA-1526 Ri = 78.95 in. Pa= 2.75 ksi t= 8.08 in. CR = 100 OF/hr a, = 2.0200 in. o,= 85.30 ksi S F = 1.15 SF = 1.25 Aa a KIP Kit ae KIpl Kt' J1 KIP Kit ae KIP' Kit' J1 (in.) (in.) (ksidin) (ksidin) (in.) (ksidin) (ksidin) (Iblin) (ksidin) (ksidin) (in.) (ksidin) (ksidin) (Iblin) 0.000 2.02 89.66 19.71 2.1072 92.06 19.72 420 97.46 19.71 2.1201 100.45 19.72 486 0.025 2.045 90.35 19.72 2.1333 92.78 19.71 426 98.21 19.72 2.1464 101.24 19.71 492 0.050 2.07 91.04 19.72 2.1594 93.50 19.71 431 98.95 19.72 2.1727 102.03 19.71 499 0.075 2.095 91.73 19.72 2.1856 94.22 19.70 437 99.70 19.72 2.1990 102.82 19.70 505 0.100 2.12 92.41 19.72 2.2117 94.94 19.70 442 100.45 19.72 2.2253 103.60 19.69 512 0.125 2.145 0.150 2.17 0.175 2.195 0.200 2.22 0.225 2.245 0.250 2.27 0.275 2.295 0.300 2.32 0.325 2.345 0.350 2.37 0.375 2.395 0.400 2.42 0.425 2.445 0.450 2.47 0.475 2.495 0.500 2.52

BAW-2494, REV. 1 Table 6-3 J-R Curves for Evaluation of Levels A and B Service Loadings Weld: !;A-1 526 T= 543 OF t= 8.08 in.

a, = 2.02 in.

+t IS = 12.48 x 10 nlcm2@ inside surface Cu = 0.34 wt%

B, = 0.80 in Aa a 41 InCl c1 c2 c3 J-R (Iblin)

(in.) (in.) (lo1* n/cm2) Mean Low 0.001 2.0210 7.6836 0.24083 1.27230 0.08892 -0.09502 83 58 AREVA

BAW-2494, REV. 1 Figure 6-1 J-Integral vs. Flaw Extension for Levels A and B Service Loadings Lower Bound J-R Curve

--- Japp w/ SF=1.25

. - - - -Japp

. w/ SF=1.15 E v a l u a t i o n Line for SF=1. I 5 0.00 0.05 0.10 0.15 0.20 0.25 Flaw Extension, Aa (in.)

h AREVA

BAW-2494, REV. 1 7.0 Evaluation for Levels C and D Service Loadings A flaw depth of 'llo of the base metal wall thickness is used to evaluate the Levels C and D Service Loadings. Based on the results of Table 6-1 for Levels A and B Service Loadings and flaw depths equal to % of the wall thickness, the controlling weld for Levels C and D Service Loadings is the SA-1526 longitudinal weld.

Table 7-1 presents applied stress intensity factors, Kl, from the PCRIT pressurized thermal shock analysis of the steam line break transient described in Section 5.2, along with total stress intensity factors after including a contribution of 9.0 ksidin from cladding, as discussed in Section 4.4. The stress intensity factor calculated by the PCRIT code is the sum of thermal, residual stress, deadweight, and pressure terms. Table 7-1 also shows the variation of crack tip temperature with time for the SLB event. To determine the critical time in the transient for the Levels C and D flaw evaluation, allowable stress intensity factors are calculated for both the transition and upper-shelf toughness regions. Transition region toughness is obtained from the ASME Section XI equation for crack initiation [15],

using an RTNDT value of 284.8"F from PCRIT for a flaw depth of 'Il0 of the wall thickness, where:

Klc = transition region toughness, ksidin T = crack tip temperature, OF The RTNDTvalue of 284.8"F calculated in PCRIT is based on the Initial RT;VDT value of -7.0°F and Margin term of 69S°F. Use of the newly-approved Topical Report BAW-2308, Rev. 1, [16, 171 would allow usage of an Initial RTNDTvalue of -81.8"F and Margin term of 60.6"F for weld material S,A-1526. This would result in a much lower RTNDTvalue of approximately 201°F for use in the KlCequation shown above. It is therefore conservative to use the RTNDTvalue of 284.8"F.

Upper-shelf toughness is derived from the J-integral resistance model of Section 3.1 for a flaw depth of 'Il0of the wall thickness, a crack extension of 0.10 in., and a fluence value of 10.04 x 1018 n/cm2, as follows:

where KJ, = upper-shelf region toughness, ksidin Jo.,= J-integral resistance at Aa = 0.1 in.

Toughness values are given in Tables 7-2 and 7-3 for the transition and upper-shelf regions, respectively, as a function of temperature.

Figure 7-1 shows the variation of applied stress intensity factor, K,, transition range toughness, Klc, and upper-shelf toughness, KJc with temperature. The small triangles on the Kl curve indicate points in time at which PCRIT solutions are available. In the upper-shelf toughness range, the Kl curve is closest to the lower bound Kj, curve at 6.5 minutes into the transient. This h

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BAW-2494, REV. 1 time is selected as the critical time in the transient at which to perform the flaw evaluation for Levels C and D Service Loadings.

Applied J-integrals are calculated for the controlling weld (SA-1526) for various flaw depths in Table 7-4 using stress intensity factors from PCRlT for the steam line break transient (at 7.0 min.) and adding 9.0 ksidin to account for cladding effects. Stress intensity factors are converted to J-integrals by the plain strain relationship, Table 7-4 lists flaw extensions vs. applied J-integrals. As the Surry vessels are 8.08 in. thick, the initial flaw depth of ' l l oof the wall thickness is 0.808 in. Flaw extension from this flaw depth is calculated by subtracting the initial flaw depth of 0.808 in. from the built-in PCRlT flaw depths.

The results, along with mean and lower bound J-R curves developed in Table 7-5, are plotted in Figure 7-2. An evaluation line is used at a flaw extension 0.10 in. to show that the applied J-integral is less than the lower bound J-integral of the material, as required by Appendix K for Level C Service Loadings [4]. The requirements for ductile and stable crack growth are also demonstrated by Figure 7-2 since the slope of the applied J-integral curve is considerably less than the sllopes of both the lower bound and mean J-R curves at the points of intersection.

Referring to Figure 7-2, the Level D Service Loading requirement that the extent of stable flaw extension be no greater than 75% of the vessel wall thickness is easily satisfied since the applied J-integral curve intersects the mean J-R curve at a flaw extension that is only a small fraction of the wall thickness (less than 1%).

The last requirement is that the internal pressure p shall be less than P,, the internal pressure at tensile instability of the remaining ligament. Table 7-6 gives the results of the calculations for PI for flaw depths up to 1.29 in. As the internal pressure p is less than PI, the remaining ligament is not subject to tensile instability.

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BAW-2494, REV. 1

'able 7-1 K, vs. Crack Tip Temperature for SLB a/t =I110 a = 0.808 in.

PCRlT Clad Total Time Temp K I S ~ KI KI h

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BAW-2494, REV. 1 Table 7-2 K,,at 1110 Wall Thickness KI, Curve at a = 1110T RTNDT = 284.8 OF T T-RTNDT K I ~

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BAW-2494, REV. 1 Table 7-3 Kj, at 1/10 Wall Thickness with Aa = 0.10 in.

Kj, Curve with Aa = 0.10 in.

Fluence = 12.48 x 10" nlcm2at inside surface

= 10.04 x 10 n/cm2at t/10 + 0.1" Aa = 0.10 in.

Cu= 0.34 wtOh E = 27042 ksi v = 0.30 Cq = -0.4489 Lower Lower Mean Bound Mean Bound T lnCl C1 C2 C3 Jo.1 Jo.1 KJ~ KJ, (ksidin) (ksidin) 183.0 153.0 h

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BAW-2494, REV. 1 Table 7-4 J-Integral vs. Flaw Extension for Level C and D Service Loadings rime = 6.50 min E= 27042 ksi

rack tip at t110 t= 8.08 in. v= 0.3 (a/t)*40 a Aa Temp. KIW~ Kldad K~tota~ Japp (in.) (in.) (F) (Iblin) 1 0.2020 323.60 52.55 9.0 61.6 127 2 0.4040 339.20 73.31 9.0 82.3 228 At Aa = 0.10 in., Japp= 371 Iblin.

LEVEL D Jo.~lJapp= 2.54 LEVEL C Jo.llJapp= 1.77 A

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BAW-2494, REV. 1

'Table 7-5 J-R Curves for Evaluation of Levels C and D Service Loadings Weld: SA-1526 Time = 6.50 min.

T= 368.70 OF t= 8.08 in.

a, = 0.808 in.

4 IS = 12.48 x 10" nlcm2@ inside surface Cu = 0.34 wt%

B, = 0.80 in Aa a 41 lnCl c1 c2 c3 J-R (Iblin)

(in.) (in.) (10" nlcm2) Mean Low 0.001 0.8090 10.2776 0.47981 1.61577 0.1 1673 -0.09721 83 58 h

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BAW-2494, REV. 1 Table 7-6 Level D Service Loadings - Internal Pressure at Tensile Instability flaw depth a (in.) P, (ksi)

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BAW-2494, REV. 1 Figure 7-1. K, vs. Crack Tip Temperature for Levels C & D Service Loadings I

I

! I - - - Klc I I I

I I

-. .- KJc Mean

- -.- - - - .i I 4- .I.--

KJc Lower Bound

' / --.--..-. . -. -. Upper Shelf Limit

! I - - .-.

I A KI at a=tll 0 for SLB I I 1

i I I I

Evaluation point at 6.50min. into transient

/

/

/ I

/

/

/

/ I

/

f I

I Upper-Shelf Toughness Range 300 350 400 450 500 550 Crack Tip Temperature (OF)

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BAW-2494, REV. 1 Figure 7-2. J-Integral vs. Flaw Extension for Levels C & D Service Loadings L

- - - Mean

- J-R Curve Lower Bound J-R Curve

- - - Japplied for SLB at 6.50 Imin.

E v a l u a t i o n Line for Level C 0.00 0.05 0.10 0.15 0.20 0.25 0.30 0.35 0.40 0.45 0.50 Flaw Extension, Aa (in.)

A AREVA

BAW-2494, REV. 1 8.0 Summary of Results A low upper-shelf toughness fracture mechanics analysis has been performed to evaluate the reactor vessel weld at Surry Units 1 and 2 for projected low upper-shelf energy levels at 48 EFPY, considering Levels A, B, C, and D Service Loadings of the ASME Code.

Evidence that the ASME Code,Section XI, Appendix K [4] acceptance criteria have been satisfied for Levels A and B Service Loadings is provided by the following:

(1) The limiting weld is the axial weld SA-1526 of Surry Unit 1. Figure 6-1 shows that with factors of safety of 1. I 5 on pressure and 1.0 on thermal loading, the applied J-integral (J1) is less than the J-integral of the material at a ductile flaw extension of 0.10 in. (&). The ratio Jo.lIJ1= 1.26 is greater than the required value of 1.O.

(2) Figure 6-1 shows that with a factor of safety of 1.25 on pressure and 1.0 on thermal loading, flaw extensions are ductile and stable since the slope of the applied J-integral curve is less than the slope of the lower bound J-R curve at the point where the two curves intersect.

Evidence that the ASME Code,Section XI, Appendix K [4] acceptance criteria have been satisfied for Levels C and D Service Loadings is provided by the following:

Figure 7-2 shows that with a factor of safety of 1.0 on loading, the applied J-integral (J1) is less than the J-integral of the material at a ductile flaw extension of 0.10 in. (JO.~). From Tables 7-4 and 7-5, for Level C Service Loadings, the ratio Jo.lIJ1 = 6581371 = 1.77, and for Level D Service Loadings, the ratio Jo.llJl =

9411371 = 2.54. Both these margins are greater than the required value of 1.0.

Figure 7-2 shows that with a factor of safety of 1.0 on loading, flaw extensions are ductile and stable since the slope of the applied J-integral curve is less than the slopes of both the lower bound and mean J-R curves at the points of intersection.

Figure 7-2 shows that flaw growth is stable at much less than 75% of the vessel wall thickness. It has also been shown that the remaining ligament is sufficient to preclude tensile instability by a large margin.

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BAW-2494, REV. 1 9.0 Conclusion The limiting Surry Units 1 and 2 reactor vessel beltline weld satisfies the acceptance criteria of Appendix K to Section XI of the ASME Code [4] for projected low upper-shelf Charpy impact energy levels at 48 effective full power years of plant operation.

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BAW-2494, REV. 1 References BAW-2192PA, Low Upper-Shelf Toughness Fracture Mechanics Analvsis of Reactor Vessels of B&W Owners Reactor Vessel Working Group For Level A & B Service Loads, April 1994.

BAW-2178PA, Low Upper-Shelf Toushness Fracture Mechanics Analvsis of Reactor Vessels of B&W Owners Reactor Vessel Workina Group For Level C & D Service Loads, April 1994.

BAW-2323, Low Upper-Shelf Touahness Fracture Mechanics Analvsis of Reactor Vessels of Surrv Units 1 and 2 for Extended Life Through 48 Effective Full Power Years, June 1998.

ASME Boiler and Pressure Vessel Code,Section XI, 1992 Edition with Addenda through December 1993.

BAW-2275, Low Upper-Shelf Toughness Fracture Mechanics Analvsis of B&W Desinned Reactor Vessels for 48 EFPY, August 1996.

ASME Boiler and Pressure Vessel Code, Section Ill, Appendices, 1986 Edition with no Addenda.

BAW-2324, Analvsis of Capsule X Virginia Electric Surrv Unit No. 1 Reactor Vessel Material Surveillance Prosram, April 1998.

BAW-2150, Materials Information for Westinnhouse-Designed Reactor Vessels Fabricated bv B&W, December 1990.

USNRC Reactor Vessel Integrity Database Version 2.0.1 (RVID), U. S. Nuclear Regulatory Commission, July 2000.

10. Regulatory Guide 1.99, Revision 2, U.S. Nuclear Regulatory Commission, May 1988.

11. AREVNFANP Document No. 32-5068629-00, "Surry 1 Updated Fluence Calculation -

Cycles 1 to 19," August 2005.

12. AREVAIFANP Document No. 32-5070607-00, "Surry 2 Fluence Calculation for Cycles 1 to 18," August 2005.

13. AREVNFANP Document No. 38-5063040-00, released April 2005 and containing Dominion Generation Letter NFP-05-011-A "Surry Equilibrium Cycle Data to Support Fluence and Equivalent Margin Analyses," dated March 21, 2005.

14. ASME Boiler and Pressure Vessel Code, Appendix K,Section XI, 2001 Edition.

15. EPRl NP-719-SR, T.U. Marston, Flaw Evaluation Procedures: ASME Section XI, Electric Power Research Institute, Palo Alto, California, August 1978.

16. BAW-2308, Revision 1, Initial RTNDT of Linde 80 Weld Materials, August 2003.

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BAW-2494, REV. 1

17. Final Safety Evaluation for Topical Report BAW-2308, Revision 1, "Initial RTN~T of Linde 80 Weld Materials," (TAC No. MB6646), August 4, 2005.

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BAW-2494, REV. 1 11.0 Certification This report is an accurate description of the low upper-shelf toughness fracture mechanics analysis performed for the reactor vessels at Surry Units 1 and 2.

HIP. Gunawardane, Engineer Ill Date Structural and Fracture Mechanics Unit This report has been reviewed and found to be an accurate description of the low upper-shelf toughness fracture mechanics analysis performed for the reactor vessels at Suny Unit 1 and 2.

Date Structural and ~ra'ctureMechanics Unit Verification of independent review.

-'9/719 9 ' 4

1. b

@ 5//2/&5 B. Djazmati, Manager / Date Structural and Fracture Mechanics Unit This report is approved for release.

Analysis Services unit AREVA