CNRO-2006-00010, Revised Request for Alternative to 10 CFR 50.55a Examination Requirements of Category B1.11 Reactor Pressure Vessel Welds

Revised Request for Alternative to 10 CFR 50.55a Examination Requirements of Category B1.11 Reactor Pressure Vessel Welds

ML060610605
Person / Time
Site:	River Bend
Issue date:	02/08/2006
From:	Burford F Entergy Operations
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
CNRO-2006-00010, RBS-ISI-004
Download: ML060610605 (15)

Text

Entergy Operations, IIic.

- %7'- EnXi 1340 Echelon Parkway Jackson, Mississippi 39213-8298 Tel 60-1-368-5758 F. G. Elurford Acting Director Nuclear Safety & Licensing CNRO-2006-0001 0 February 8, 2006 U. S. Nuclear Regulatory Commission Attn.: Document Control Desk Washington, DC 20555-0001

SUBJECT:

River Bend Station Docket No. 50-458 License No. NPF-47 Revised Request for Alternative to 10 CFR 50.55a Examination Requirements of Category B1 .11 Reactor Pressure Vessel Welds

REFERENCE:

1. Request for Alternative to 10 CFR 50.55a Examination Requirements of Category B1.11 Reactor Pressure Vessel Welds dated August 24, 2005 Ladies and Gentlemen:

In Reference 1,. Entergy Operations, Inc. (Entergy) proposed Request for Alternative RBS-ISI-004 for River Bend Station (RBS). On January 9, 2006, Entergy and members of your staff held a call to discuss fluence methodologies used in support of the request. As a result of the call, NRC staff requested confirmation of the information discussed. Entergy has revised the request to include the information from the discussion; specifically, the information supporting the first bullet on page 4 has been expanded. The revised request is included as Attachment 1.

Entergy continues to request NRC approval by mid-August of 2006 to enable preparation for RBS's fall 2007 refueling outage. Should you have any questions regarding this submittal, please contact Bill Brice at (601) 368-5076.

No new commitments are made in this letter.

Very truly yours, FGB/NBB/bal

Attachment:

Request for Altemative RBS.-ISI-004 cc: (see next page)

WaDco

CNRO-2006-0001 0 Page 2 of 2 cc: Mr. W. A. Eaton (ECH)

Mr. P. D. Hinnenkamp (RBS)

Dr. Bruce S. Mallett U. S. Nuclear Regulatory Commission Region IV 611 Ryan Plaza Drive, Suile 400 Arlington, TX 76011 NRC Senior Resident Inspector P. 0. Box 1050 St. Francisville, LA 70775 U. S. Nuclear Regulatory Commission Attn: Mr. B. J. Benney MS 0-7D1 Washington, DC 20555-0001

CNRO 2006-00010 ATTACHMENT REQUEST FOR ALTERNATIVE RBS-ISI-004

ENTERGY OPERATIONS, INC.

RIVER BEND STATION REQUEST FOR ALTERNATIVE RBS-ISI-004

1. COMPONENT / EXAMINATION I)ENTIFICATION Component / Number: Reactor / Q1B13DO01 Code Class: 1

References:

1. ASME Section Xi, 1992 Edition, IWB-2500

2. 10 CFR Sections 50.55a(a)(3)(i) and I30.55a(g)(6)(ii)(A)(2)

3. 13WRVIP-05, "BWR Reactor Pressure Vessel Shell Weld Inspection"

4. 13WRVIP Response to NRC Request for Additional Information (RAI) on BWRVIP-05, 12/22197

5. NRC Generic Letter 98-05, "Boiling Water Reactor Licensees Use of the BWRVIP-05 Report. to Request Relief From Augmented Examination Requirements on Reactor Pressure Vessel Circumferential Shell Welds"

6. NRC Final Safety Evaluation of the BWR Vessel and Internals Project BWRVIP-05 Report to (TAC No. M93925), dated July 28, 1998

7. NRC letter to Entergy Operations, Inc., dated March 26, 1999 (TAC No. MA0621)

Examination Category: B-A Item No.: B1.1 1 Examination Required: Volumetric Examination of Welds and Adjacent Base Materials

Description:

Circumferential Shell Welds in Reactor Vessel Unittinspection Interval: River Bend Station (RBS) / Second (2nd) 10-year Interval Page 1 of 12

II. CODE REQUIREMENTS The ASME code of record for the second 10-year interval at RBS is the 1992 Edition with portions of the 1993 Addenda. ASME Section Xl, 1992 Edition, IWB-2500 requires the reactor pressure vessel (RPV) circumferential shell welds and associated base material to be volumetrically examined once each interval. The examinations are to be dispersed over the three periods of the interval within the limits specified by IWB-2412-1. Deferral of the examinations until the end of the interval is permissible; however, IWB-2420 requires the sequence of examinations established in the first interval to be repeated during subsequent intervals to the extent practical.

Ill. PROPOSED ALTERNATIVE Pursuant to the provisions of 10 CFR 50.55a(a)(3)(i) and consistent with guidance provided in NRC Generic Letter 98-05 (Reference 5), Entergy Operations, Inc. (Entergy) requests NRC approval to use an alternative from the examination of RPV circumferential welds required by ASME Section Xl, IWB-2500, Examination Category B-A, Item No. BI. 11 as described below.

As a result of this request, circumferential welds AA, AB, AC, and AD of the RPV shell will not require the examinations currently specified by ASME Section Xl for the remainder of the current operating license.

A. Examination Scope Examination of the longitudinal (axial) RPV shell welds (Examination Categ Dry B-A, Item No. B1.12) shall be performed for 100% of the welds. Axial weld examination shall also include that portion of the circumferential weld that intersects each axial weld, or approximately 2% to 3% of the intersecting circumferential weld.

The procedure and personnel employed for these examinations will meet the requirements of ASME Section Xi, Appendix Vill as required by 10 CFR 50.55a(g)(6)(ii)(C).

B. Successive Examination of Flaws For flaws detected in the circumferential weld (Item No. B .11), the successive examinations required by IWB-2420 are not required for non-threatening flaws (e.g.,

embedded flaws from material manufacturing or vessel fabrication which experience negligible or no growth during the design life of the vessel), provided the following conditions are met:

(1) The flaw is characterized as "subsurface" in accordance with BWRVIP-05, Figure 9-1.

(2) The Non-Destructive Examination (NDE) technique and evaluation that detected and characterized the flaw as. originating from the material manufacture or vessel fabrication is documented in a flaw evaluation report.

(3) The vessel weld containing the flaw is acceptable for continued service in accordance with IWB-3600 and the flaw is demonstrated acceptable for the intended service life of the RPV.

Page 2 of 12

Successive examination of flaws detected in the axial welds (Item No. B1.12), shall be performed as required by IWB-2420.

C. Additional Examinations of Flaws For flaws detected in the circumferential weld (Item No. B13.1 1), the additional examinations required by ASME Section Xl, IWB-2430 are not required provided the following conditions are met:

(1) If the detected flaw is characterized as "subsurface", additional examinations are not required.

(2) If the flaw is not characterized as 'subsurface', then an engineering evaluation shall be performed addressing the following (at a minimum):

(a) A determination of the root cause of the flaw, (b) An evaluation of any potential failure mechanisms, (c) An evaluation of service conditions which could cause subsequent failure, (d) An evaluation per IWB-3600 demonstrating that the vessel is acceptable for continued service.

If the flaw meets the criteria of IWB-3600 for the intended service life of the vessel, then additional examinations may be limited to those welds subject to the same root cause conditions and failure mechanisms, up to the number of examinations required by IWB-2430(a). If the engineering evaluation determines that there are no additional welds subject to the same root cause conditions, or if no failure mechanism exists, then no additional examinations are required.

For flaws detected in axial welds (Item No. B13.12), additional examinations shall be performed as required by IWB-2430.

Examination of the circumferential welds shall be performed if axial weld examinations reveal an active, mechanistic mode of degradation. The NRC shall approve the timing and scope of these examinations.

IV. BASIS FOR ALTERNATIVE The basis for this request is documented in the report BWR Vessel and InternalsProject, BWMR Reactor Pressure Vessel Shell Weld Inspection Recommendations (BNRVIP-05)

(Reference 3) that was transmitted to the NRC in September 1995 and BWRVIP Response to NRC RAI on BWRVIP-05 (reference 4) that was transmitted to the NRC on December 18, 1997. The NRC staff approved BWRVIP-05 in a letter dated July 28, 1998 (Reference 6).

The BWRVIP-05 report provides the technical basis for eliminating examinations of BWR RFPV circumferential shell welds. The BWRVIP-05 report concludes that the probability of failure of the BWR RPV circumferential shell welds is orders of magnitude lower than that of the axial shell welds. Additionally, the NRC safety evaluation dated July 28, 1998 demonstrates that examination of BWR RPV circumferential shell welds does not Page 3 of 12

measurably affect the probability of failure. Therefore, the NRC evaluation supports the conclusions of BWRVIP-05.

The NRC evaluation of BWRVIP-05 utilized the FAVOR code to perform a probabilistic fracture mechanics (PFM) analysis to estimate RPV failure probabilities. Three key assumptions in the PFM analysis are:

(1) The neutron fluence was estimated to be end-of-life (EOL) mean fluience (2) The chemistry values are mean values based on vessel types, and (3) The potential for beyond-design-basis events is considered.

Although BWRVIP-05 provides the technical basis supporting this altemative, the following information is provided to show the conservatisms of the NRC plarit-specific analysis as they apply to RBS:

• The region of the vessel that corresponds to the top and bottom of active fuel is designated as the beltline. The RBS beltline does not contain circumferential welds. The closest two weld seams are AB, which is approximately 7 inches below the beltline, and AC, which is approximately 19 inches above the beltline region. These two seams are the closest circumferential weld seams to the active fuel core and are, therefcre, considered the limiting seams for this evaluation.

The calculated 32 effective full power years (EFPY) EOL peak fluence (E>1.0 MeV) is 4.98E+18 n/cm2 . This peak is located at approximately 1:3 inches above the mid-plane of the vessel beitline region at the clad/base metal interface. The peak value decreases about 80% to 9.96E+1 7 n/crn2 at the top and decreases slightly more at the bottom of the beltline region. For purposes of this evaluation, a conservative mean fluence value has been calculated for the two limiting seams by averaging the peak fluence with the top of beltline fluence. This mean fluence value is 2.988E+18 n/cm2 . It is used to determine the Delta Reference Temperature for nil-ductility temperature (ARTNDT) values and subsequently the mean RTNDT values for each weld heat in the two circumferential weld seams AB and AC. This value is conservative because beltline region fluences are used for non-beltline region materials.

The fluence was determined based on the vessel fluence analysis results reported in "Neutron Transport Analysis for River Bend Station" (MPM-904779). This analysis was performed by MPM Technologies using methods consistent with the requirements of Regulatory Guide 1.190. MPMI methods were previously described in "Benchmarking of Nine Mile Point Unit 1 and Unit 2 Neutron Transport Calculations" (MPM-402781 revision 1) and approved in "Nine Mile Point Nuclear Station, Unit 1 - Issuance of Amendment RE:

Pressure-Temperature Limit Curves and Tables", (TAC NO. MB6687).

Additionally, a RBS specific application of these methods is described in "BWR Vessel and Internals Project River Bend 183 Degree Surveillance Capsule Report" (BWRVIP-1 13) which was provided to the NRC by letter of June 10, 2003. The River Bend fluence calculations were updated in November, 2004 (MPM-904779) to include shroud fluence data to support scheduled inspections as described below. Both RBS fluence analyses (BWRVIP-1 13 Page 4 of 12

and MPM-904779) were performed by MPM Technologies and use the same transport methods. The more recent calculation (MPM-904779) used more detailed source term data. The data was revised in the following four areas:

1) MPM-904779 includes eight separate transport calculations spread through out the cycle 1 core's operation to support the cycle 1 dosimetry evaluation.

BWRVIP-1 13 uses a single cycle average calculation since the initial cycle dosimetry was not evaluated.

2) Both MPM-904779 and BWRVIP-1 13 evaluate each cycle using cycle average input to the transport calculations except for cycles just prior to dosimetry removal. The BWRVIP-1 1:3 calculation used the 3D moderator density distribution from the middle of cycle point with the power distribution closest to the cycle average distribution. MPM-904779 calculated a weighted average moderator density distribution from 3D moderator distributions calculated throughout core life.

3) MPM-904779 included actual operational data for cycles 1 - 12 while BWRVIP-113 only used operational data for cycles 1 - 9. MPM-904779 used a slightly more conservative fuel cycle to extrapolate the fluence to end of plant life.

4) BWRVIP-1 13 was performed in 1/8 core symmetry while MPM-904779 used %

core symmetry. This change more correctly models the jet pumps which are 1/4 core symmetric. The 183 degree capsule dosimetry is not impacted since the asymmetry occurs approximately 90 degrees from the capsule location. At the position of the highest fluence the asymmetry is less than 4%.

The refinements included in MFPM-904779 result in an increase in accuracy as illustrated by Table I and a slight increase in the maximum vessel fluence. The 32 EFPY maximum inside radius fluence increased from 4.38E+18 to 4.98E+18. This difference is less than the uncertainties of either calculation and, therefore, the methods used in the reports are considered equivalent. However, the MPM-904779 results are preferred based on the improved accuracy.

Table 1: DosImetry Comparisons Calculated/Measured Ratio for River Bend BWRVIP-1 13 MPM-904779 Cycle 1 Flux Wire Fe Total N/A 1.04 Cycle 9 Cu Total 0.803 0.872 Fe Total 0.914 1 Cycle 9 Capsule Total 0.858 0.936 Page 5 of 12

• For RPVs fabricated by Chicago Bridge & Iron (CB&I), the EOL neutron fluence for the circumferential weld used in the NRC plant-specific analysis is 0.510E+19 n/cm 2 (see Table 2.6-4 of Reference 6). The maximum (peak) inner surface fluence used for the RPV limiting non-beltline region at the EOL (32 E-FPY) is predicted to be 0.2988E+19 n/ cm]. Thus, the effect of fluence on embrittlement is lower, and the NRC analysis as described in the NRC safety evaluation (Reference 6) is conservative for RBS.

• Review of the RPV fabrication records also reveals that Heat No. 7691 6 (SMAW 5/32" E8018-G) was used only for weld pick-ups on Seam AB and is, therefore, considered to be a very small contributor to the overall weld metal properties for Seam AB. However, as shown in Table 2, the attributes of this speciRc weld heat causes Seam AB to be the limiting circumferential weld.

As described above, there is conservatism in the already low circumferential-weld-failure probabilities as related to RBS. Other RPV shell weld information that compares to the information used in the NRC plant-specific analysis is provided in Table 2.

As shown in Table 2, the limiting circumferential weld in the RBS RPV is Seam AB with a maximum calculated shift in RTNDT (i.e., ARTNDT) of 90'F and a mean reference temperature (i.e., mean RTNDT) of 301F at 32 EFPY (EOL). The values used by the NRC in the plant-specific analysis for ARTNDTis 109.50 F with a resulting mean RTNDT of 44.51F.

Therefore, it is evident that the values used by the NRC in their plant-specific analysis are bounding and provide additional assurance that the RPV welds are also bounded by the BWRVIP-05 report.

An added margin is also provided at RBS by the NDE of the RPV welds. The examination coverage for all welds, except for circumferential weld AA, exceeded 90%

coverage of the full volume during the first 10-year examination. Weld AA examination coverage was limited to 62% weld volume. RPV geometric limitations created by the change in vessel diameter at the transition to the bottom head prevented greater than 90% weld volume coverage. This limitation was evaluated and approved by the NRC staff in a letter to RBS dated March 26, 1999 (TAC No. MA0621) (Reference 7).

In previous evaluations, the NRC concluded that beyond design basis events occurring during plant shutdown could lead to cold over-pressure events that could challenge RPV integrity. As indicated by the NRC safety evaluation, each licensee requesting relief is to demonstrate the implementation of operator training and the existence of procedures that limit the frequency of cold over-pressure events to the amount specified in the NRC plant-specific evaluation.

Entergy has reviewed the BWRVIP's response and concurs that the conditions and events are accurately depicted and that the procedures and personnel training at RBS are comparable to those described by the BWRVIP and are adequate to prevent a cold over-pressure transient event. Consequently, the probability of a cold over-pressure transient is considered to be less than or equal to that used in the NRC plant-specific analysis described in the NRC safety evaluation (Reference 6) and is conservative for RBS.

Page 6 of 12

Review of Potential Injection Sources That Could Cause an RPV Cold Over-Pressurization Event:

The Reactor Core Isolation Cooling (RCIC) system is one of the high pressure make-up systems at RBS. The RCIC system utilizes a steam turbine-driven pump. RCIC injection during cold shutdown is not possible since no steam is available to drive the RCIC turbine.

The High Pressure Core Spray (HPCS) system is another high pressure make-up system at RBS. The HPCS pump is motor operated, so it can be operated when the reactor is in cold shutdown. However, the HPCS system would require manual initiation, inadvertent initiation, or manual startup to start and inject into the RPV. Also, there is a high RPV water level interlock for the HPCS injection valve to prevent overfilling the RPV. This high level interlock is not normally overridden (this is done only for valve stroke testing with the HPCS pump breaker racked out). Even if the HPCS system is inadvertently started, it would not overfill and pressurize the reactor due to the high level interlock.

The Standby Liquid Control System ('SBLC) is a high pressure system used to shut down the reactor if the control rods fail to insert. SBLC has no auto start function so a spurious start is unlikely. SBLC must be manually initiated by a key lock switch. A Control Room Senior Reactor OperatDr (SRO) maintains custody of the keys. SBLC is a low flow rate system (about 42 gpm per pump) and is limited to the amount of water that is contained in the storage tank (about 5000 gallons). Even if the SBLC was manually initiated and not monitored, there is nDt enough water in the storage tank to Fill the RPV from normal water level and would not, therefore, pressurize it.

The Reactor Feed Pumps (RFP) provide high pressure make-up during normal operation. RBS has three 33% motor-driven pumps that are fed by the condensate pumps (CPs). These pumps provide full flow at pressures of about 1400 psig to the reactor. RBS has three 50% capacity CPs that supply the RFPs during normal operation. The CPs have a discharge pressure of about 550 psig. This corresponds to about 340 psig suction pressure at the RFP. The CPs have a shutoff discharge pressure of about 575 psig. During operation of the CPs, sufficient temperature margin is provided to ensure that the Technical Specification for the reactor pressure and temperature (P/T) is not exceeded. This is accomplished by plant procedures dictating when CP operation is allowed. When the plant is in cold shutdown, reactor temperature is maintained above 700F per Technical Specifications.

For the reactor P/T limit to be exceeded, a CP would have to be manually started and manually lined up for injection. Then the injection would have to be ignored by the operating crew and allowed to continue until the reactor is pressurized above the P/T limits. The operating crew would have numerous indications that condensate was injecting (feed flow indicators and recorders, check valve indication) and that reactor level and pressure were increasing (upset and shutdown level indication and recorders, narrow and wide range pressure indicators and recorders). Because of the number of operator errors that would have to occur and the number of indications that would have to be ignored, the probability of this event is very low.

The Low Pressure Core Spray (LPCS) and Low Pressure Core Injection (LPCI) systems are low pressure emergency core cooling systems (ECCS). LPCI is actually a mode of the Residual Heat Removal system. Technical Specifications for the reactor P/T limit permit pressures from about 95 psig up to about 313 psig at temperatures from 70 up to Page 7 of 12

approximately 120'F. Between 120'F and 1500 F, the pressure permitted by Technical Specifications remains constant at about 313 psig. At 1500 F, the allowable pressure increases immediately to roughly 540 psig and thereafter increases rapidly as temperature increases. LPCS has a discharge pressure of about 500 psig and LPCI has a discharge pressure of about 329 psig. During refueling outages, there is; typically only a very short period of time during RPV head de-tensioning and following head re-tensioning in which an overpressurization event could occur. As soon as the head is re-tensioned, plant procedures instruct the operators to begin heat-up for the RPV inservice leak test. Temperatures are normally maintained between 900F and 12001: during shutdown (temperatures are allowed to range from 700F to 2000 F in Mode 4 and from 700F to 140OF in Mode 5 with approval of the SRO). Therefore, the reactor bulk coolant temperature is normally well above 70'F. Procedural controls and the short period of time when the vessel coolant temperatures are low, make the probability for an overpressurization due to an inadvertent actuation of these systems very low. The automatic initiation signals for LPCI and LPCS are low reactor water level of -143" and high differential pressure between the drywell and containment of 1.68 psid. During the short periods around vessel head de-tensioning and re-tensioning, the reactor water level is maintained near the vessel flange which is 200" and the drywell head is removed making the Drywell and containment one area. Because of this, the possibility of an automatic initiation of these systems is very small.

The Control Rod Drive (CRD) system is a high pressure system used to operate the control rods. The CRD system is a low flow rate system with about 45 gprn flow rate to the reactor. During cold shutdown conditions, reactor water level is maintained with CRD (makeup) and the Reactor Water Cleanup System (RWCU) (reject). Per plant procedures, the RPV head vents are open when reactor coolant temperature is less than 1900 F. During cold shutdown conditions, the operators closely monitor reactor water level, pressure, and temperature. With the CRD flowrate low and the RPV head vents open, the operators should have sufficient time to react to regain control of reactor pressure, should any abnormalities occur.

Post Outage Primary System Hydrostatic Testing is a postulated over pressurization event. RBS has plant procedures as well as Technical Specifications that dictate parameters and steps when performing hydrostatic testing. Hydrostatic testing is considered an "Infrequently Performed Test or Evolution". This requires management oversight, crew briefs, review of industry events, and assigned responsibilities for the test to be performed. Reactor coolant is heated up to >1500F before reactor pressure is increased to test pressure. Reactor level is maintained with CRD (make-up) and/or RWCU (reject). Reactor pressure changes are limited to 50 psi per minute by plant procedures. All safety relief valves are required to be operable during the test by plant procedures. Because of these strict controls, the likelihood of an over pressurization event during a hydrostatic test is minimal.

Procedural Controls and Operator Traininq That Prevent Reactor Pressure Vessel Cold Over-Pressurization:

Plant procedures and Technical Specifications dictate bands at which reactor water level, pressure, and temperature are to be maintained that ensure an adequate level of safety during all modes of operation. Operation of RBS follows the steam saturation curve; therefore, the operating temperatures are expected to be well in excess of the minimum temperatures required by Technical Specifications. The Control Room operators are Page 8 of 12

required by procedure to maintain reactor parameters (i.e., water level, pressure, and temperature) within these bands and to frequently monitor those parameters. They are also required by procedure to report to the SRO anytime operation is outside of a prescribed band. The SRO is responsible to ensure that actions are taken to return those parameters back within the desired band. Also, as previously noted, plant procedures require pre-job briefings and contingency plans before infrequent tests or evolutions are performed. Training reinforces these requirements in both classroom and simulator training. Finally, plant conditions, status of plant equipment, special activities along with their potential effect on key plant parameters, and contingency planning are discussed with oncoming crews during shift turnover.

At RBS, work performed during an outage is scheduled by the Outage Management group. Outage Management includes SROs who provide oversight of the outage schedule development to avoid conditions that could adversely affect reactor water level, pressure, or temperature. The outage schedule is reviewed by the Outage Risk Assessment Team (ORAT) to assure a proper level of plant safety is maintained.

Significant emergent work or significant changes to the schedule are also reviewed by the ORAT. The outage is performed following this schedule.

General Operating Procedure (GOP), 'Power Decrease/Plant Shutdown", requires that the RPV head vents be opened when reactor coolant temperature is about 190'F during reactor cooldown and administrative control be established to ensure that the RPV remains vented. During hydrostatic testing, the Reactor Vessel In-Service L.eak Test procedure requires reactor coolant temperature be heated up to >150'F and all Safety Relief Valves to be operable prior to increasing reactor pressure. These help ensure the Technical Specifications requirements for reactor P/T limits are not exceeded.

V. CONCLUSION Based on BWRVIP-05, the BWRVIP's response to the NRC's Request for Additional Information, the plant-specific analysis performed by the NRC staff, and the discussions above, Entergy believes the proposed alternative to eliminate the described examinations is reasonable and provides an acceptable level of quality and safety.

Therefore, Entergy requests the NRC staff authorize the use of RBS-ISI-004 in accordance with 10 CFR 50.55a(a)(3)(i).

This proposed alternative is for the remaining term of operation under the existing license.

Page 9 of 12

- I Table 2

SUMMARY

OF RESULTS (see supplemental information)

Summary of Results for MEAN Reference Temperature (Mean RTNDT) Evaluation for the River Bend Reactor Vessel Non-Beltline Circumferential AB and AC Seam Weld Materials by Heat Number at the Vessel l.D. Surface Applicable to 32 EFPY with Power Uprates.

Material Description Chemical Inside Surface @ Vessel l.D.

Composition (Clad/Base Metal Interface)

Vessel Vesl32 Weld Material Identification Cu NI Initial Chemistry EFPY (EOL~

32uEFcY nlc RNT ARTNOT3 F enRNT Mean RTN3T, F Circ. Weld Fleenc Seam wt% wt% RTNDT Factor (CF) uen Mev ncrnat 32 EFPY at 32 EFPY River Bend Results Raco/NMM (Tandem Wire)

AB 4P7216 / Linde 124 / 0751 0.038 0.820 -80 51.4 2.988E+18 34.4 -45.6 Raco/NMM (Single Wire)

AB 4P7216 / Linde 124/ 0751 0.038 0.820 -50 51.4 2.988E+18 34.4 -15.6 Raco/NMM (Tandem Wire)

AB 4P7465 / Linde 124 / 0751 0.02 0.807 -60 27.0 2.988E+18 18.1 -41.9 Raco/NMM (Single Wire)

AB 4P7465 / Linde 124/0751 0.02 0.807 -60 27.0 2.988E+18 18.1 -41.9 (SMAW Process)

An I A A n fI~ a-

-6.4 If~ff I a- I - - - .A aflaI A^B (2RSMuW B57Pc)048 0.09 U.96 -60 12.0 2.988E+1 8 81.6 (SMAW Process)

AB 03L048 / B525B27AF l0.09 l0.96 l-60 l 122.0 l 2.988E+18 l 81.6 21.6 Page 10 of 12

Material Description Chemical Inside Surface @ Vessel l.D.

Composition (Clad/Base Metal Interface)

Vesl32 EFPY (EOLl RIT enRNT Vessel Weld Material Identification Cu NI Initial Chemistry FluenceF ncMean nr ARTNDT, F RTNnT, F Circ. Weld wt% wt% RTNDT Factor (CF) at 32 EFPY at 32 EFPY Seam (Em1.0 Mev.)

(SMAW Process)

AB 76916/ D516B27AE 0.10 0.95 -60 134.5 2.988E+18 [90.0] [30.0]

(SMAW Process)

AB L83978 / J414B27AD 0.02 1.06 -80 27.0 2.988E+18 18.1 -61.9 Raco/NMM (Tandem Wire)

AC 5P6771 t Linde 124/ 0342 0.034 0.934 -20 46.2 2.988E+18 30.9 10.9 Raco/NMM (Single Wire)

AC 5P6771 ! Llnde 124 / 0342 0.034 0.9 -30 46.2 2.988Eh18 30.9 0.9 (SMAW Process)

AC 640892 / J424B27AE 0.09 1.00 -60 122.0 2.988E+1 8 81.6 21.6 (SMAW Process)

AC 629865/ A421A27AD 0.05 1.10 -70 68.0 2.988E+18 45.5 -24.5 NRC Bounding Criteria for CB&I Group Results NRC Limiting Plant-Specific Circ. Weld Weld Analysis Parameters at 0.10 0.99 -65 134.9 5.10E+18 109.5 44.5 32 EFPY SER Table 2.6-4, l___ __ _CB&I Group I I II__ I i _ _ _ _I

'j JLirIItIIgI"utifii vaiue of the Shift in the Reference Temperatures (ARTNDT) and the Mean Reference Temperatures (Mean RTNOT).

Page 11 of 12

SUPPLEMENTAL INFORMATION RELATED to Table 2

1) RBS RPV beitline does not contain circumferential welds. Weld Seam AB is approximately 7 inches below the vessel beltline and Weld Seam AC is approximately 19 inches above the vessel beltline region. Weld Seams AB and AC are the two circumferential welds closest to the active fuel core and are therefore considered to be the limiting vessel girth welds in this evaluation.

2) The calculated 32 EFPY peak fluence value at the inner surface location is 4.98E+ 18 n/cm2 located at 13 inches above the mid-plane of the core and corresponding vessel beltline region. This peak fluence value decreases to 9.96E+17 n/cm 2 (peak reduced by 80%) at the top and bottom locations of the vessel beltline region.

Although the maximum inner surface fluence applicable outside of the vessel beltline region is 9.96E+1 7 n/cm2 or lower, a mean fluence value of 2.988E+1 8 n/cm 2 for the vessel beltline is established using an average of the peak and reduced values for the beltline region.

This alternative evaluation conservatively uses the 32 EFPY mean fluence value (2.988E+18 n/cm2 ) that applies to the vessel beltline region to determine the ARTNDT values and subsequently the Mean RTNDT values for each weld heat within the two circumferential seams AB and AC, which are non-beltline materials. In addition, no credit is taken for the attenuation caused by the RPV inner surface cladding.

3) Initial RTNDT determined in accordance with ASME Section III, NB-2300.

4) The largest 32 EFPY ARTNDT and Largest 32 EFPY mean RTNDT occurs in the SMAW portion (Ht No. 76916) of weld AB; therefore, this weld is the limiting of the two circumferential welds and is used to compare against the NRC's plant-specific analysis 32 EFPY parameters for vessels fabricated by CB&I. This Request for Alternative has compared the shift of Weld Seam AB to the shift predicted by the NRC for the CB& I Fabrication Group, and the RBS shift is less. As such, the NRC analysis is bounding and provides basis for this alternative.

5) Mean RTNDT is determined using Eq. 'I of Reg. Guide 1.99 excluding the Margin Term in Eq. 4. ARTNDT is determined using Eq. 2. Eq. 3 is not used because the evaluation is for the inner surface of the vessel (no attenuation into the vessel).

6) Based on BWRVIP-1 35, there are no surveillance data sets for any of the weld heats identified in Table 2. However, the best estimate chemistry values provided in Table D-1 of BWRVIP-135 were used to determine the Chemistry Factors used in Table 2 for weld heats 4P7216, 4P7465, and 5P6771.

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