RA-20-0324, Response to Request for Additional Information for Proposed Alternative for RPV Nozzle-to-Vessel Weld and Inner Radii Examination Requirements

Response to Request for Additional Information for Proposed Alternative for RPV Nozzle-to-Vessel Weld and Inner Radii Examination Requirements

ML20301A475
Person / Time
Site:	Brunswick
Issue date:	10/27/2020
From:	Ratliff J Duke Energy Progress
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
RA-20-0324
Download: ML20301A475 (16)

Text

Jay Ratliff Plant Manager Brunswick Nuclear Plant 8470 River Rd SE Southport, NC 28461 o: 910.832.3480 Serial: RA-20-0324 10 CFR 50.55(a)

October 27, 2020 United States Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 BRUNSWICK STEAM ELECTRIC PLANT, UNIT NOS. 1 AND 2 RENEWED FACILITY OPERATING LICENSE NOS. DPR-71 AND DPR-62 DOCKET NOS. 50-325 AND 50-324

SUBJECT:

Response to Request for Additional Information for Proposed Alternative for RPV Nozzle-to-Vessel Weld and Inner Radii Examination Requirements

REFERENCES:

1. Duke Energy Letter RA-19-0447, Proposed Alternative for RPV Nozzle-to-Vessel Weld and Inner Radii Examination Requirements in Accordance with 10 CFR 50.55a(z)(1),

dated June 23, 2020 (ADAMS Accession Number ML20181A004).

2. Duke Energy Letter RA-20-0247, Supplement to Proposed Alternative for RPV Nozzle-to-Vessel Weld and Inner Radii Examination Requirements in Accordance with 10 CFR 50.55a(z)(1), dated July 30, 2020 (ADAMS Accession Number ML20212L731).

3. Email from Andy Hon to Art Zaremba, Request for Additional Information - Brunswick Request for Alternate Examination of Reactor Vessel Nozzles (EPID: L-2020-LLR-0091) dated October 1, 2020 (ADAMS Accession Number ML20275A297).

Ladies and Gentlemen:

By letter dated June 23, 2020 (Reference 1), as supplemented by letter dated July 30, 2020 (Reference 2), Duke Energy Progress, LLC. (Duke Energy) submitted Relief Request RA 0447 in accordance with 10 CFR 50.55a(z)(1) to the requirements of American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) for the Brunswick Steam Electric Plant, Units 1 and 2.

By email dated October 1, 2020, (Reference 3), the NRC requested additional information required to complete its review. The Enclosure to this letter provides Duke Energys response to the request. The Attachment contains a markup of Attachment A of Westinghouse report LTR-REA-20-109 which provides supporting information for Duke Energys response to RAI 5 of the Enclosure.

This document contains no new Regulatory Commitments.

U.S. Nuclear Regulatory Commission Page 2 Serial: RA-20-0324 Should you have any questions concerning this letter, or require additional information, please contact Art Zaremba, Director - Nuclear Fleet Licensing, at 980-373-2062.

Sincerely, Jay Ratliff Plant Manager Brunswick Steam Electric Plant

Enclosure:

Duke Energy Response to Request for Additional Information Attachment

1. Attachment A of LTR-REA-13-19, Rev. 0, Additional Request for RPV Fast Neutron Fluence at Brunswick Units 1 and 2 cc : L. Dudes, Regional Administrator USNRC Region II Mr. Andrew Hon, Project Manager Mr. Gale Smith, NRC Senior Resident Inspector Chair - North Carolina Utilities Commission Mr. W. Lee Cox, Ill Section Chief, Radiation Protection Section, NC DHHS

RA-20-0324 Enclosure Page 1 of 7 Enclosure Duke Energy Progress, LLC Brunswick Steam Electric Plant Units 1 and 2 Duke Energy Response to Request for Additional Information

RA-20-0324 Enclosure Page 2 of 7

RAI 1

Issue:

The licensee states in Section 4 of Enclosure 1 to the supplement dated July 30, 2020 that 2 million realizations were performed as part of the probabilistic fracture mechanics (PFM) analysis (see fifth paragraph of Section 4). However, the licensee states in Section 5 of to the Supplement dated July 30, 2020 that no failures occurred for any path in 1 million simulations A similar statement is found in Enclosure 4 to letter dated June 23, 2020. The staff takes the term simulation to mean realization, in this context. There is an apparent discrepancy in the submitted documents on the number of realizations performed as part of the PFM analysis.

Request:

Clarify how many realizations were performed as part of the PFM analysis.

Response

In the context used, it is confirmed that simulation and realization are interchangeable. One million simulations were performed for the PFM analysis, and probabilities of failure are calculated for one million simulations. In the fifth paragraph of Section 4 of Enclosure 1 to the supplement dated July 30, 2020, 2 million times is a typographical error and should read one million times.

RAI 2

Issue:

The licensee states in Section 5 of Enclosure 1 to the supplement dated July 30, 2020 that the probability of failure is estimated as 1 failure / 1 million realizations / 60 years = 1.67 x 10-8 per year. This calculation implies that a converged solution was reached in the PFM. The uncertainty in the mean failure probability is not addressed in the licensees PFM analysis. This uncertainty may be important when comparing the mean failure probability to the chosen acceptance criterion of 5x10-6 per year, if the acceptance criterion is within two standard deviations of the mean failure probability.

Request:

Provide (1) a discussion of the uncertainty on the mean failure probability in relation to the acceptance criterion and (2) a discussion of solution convergence.

RA-20-0324 Enclosure Page 3 of 7

Response

(1, 2)

The uncertainty in the mean failure probability is addressed by determining the error associated with the Monte Carlo simulation. In the Monte Carlo simulation, errors in the estimated failure probability resulting from a given number simulations/realizations (i.e., sample size) can be evaluated as [1]:

% Error = 200 where is the probability of failure and n is the sample size (number of realizations). It should be noted that the error in the estimation of the probability of failure is due to the limited sample size which also affects the convergence. For example, when an infinite number of realizations is performed (which leads to total convergence), the error in the estimation of mean probability of failure would be zero. The error (and the lack of convergence) increases as the sample size is decreased from infinity.

As detailed in the response to RAI 4 below, there were no failures in the nozzle-to-shell welds and nozzle inner radius sections using one million iterations. The conditional cumulative (mean) probability of failure for a Low Temperature Overpressure (LTOP) event is therefore less than 1.0x10-6. The estimated error using the above relation is 200%. Thus, it is 95% (approximately two standard deviations) likely that the actual probability would be within 1.0x10-6 +/- 2

• 1.0x10-6.

Taking the upper bound of this will result in a conditional cumulative probability of failure for an LTOP event less than 3.0x10-6. This corresponds to a conditional probability of failure less than 5x10-8 per year after dividing by 60 years.

Considering the LTOP occurrence of 1x10-3 per year, the upper bound probability of failure during an LTOP occurrence is less than 5x10-11 per year. This upper bound probability of failure for an LTOP event bounds the Brunswick LTOP results in Table 12 of Enclosure 1 (See table in response to RAI 4) and is at least five orders of magnitude less than the acceptance criterion of 5x10-6. Because of the relatively low probability of failure compared to the acceptance criterion for an LTOP event, any uncertainty (even an order of magnitude) should not affect the conclusion of the PFM evaluation and accounts for any uncertainty in solution convergence.

Furthermore, convergence studies were performed in Section 8.2.1 of BWRVIP-05 [2] for the VIPER Code (the predecessor to VIPER-NOZ) to show that the use of 1 million realizations is adequate to achieve convergence. All important features such as solution convergence developed for the VIPER code in BWRVP-05 were maintained in the VIPER-NOZ code.

Reference

1. A. H-S. Ang, W.H. Tang, Probability Concepts in Engineering and Design, Vol. II:

Decision, Risk and Reliability, John Wiley & Sons, 1984.

2. BWRVIP-05: BWR Vessels and Internals Project, Technical Basis for the Reduction of Inspection Requirements for the Boiling Water Reactor Nozzle-to-Shell Welds and Nozzle Blend Radii, EPRI, Palo Alto, CA 2002. 1003557 (ADAMS Accession Number ML023330203).

RA-20-0324 Enclosure Page 4 of 7

RAI 3

Issue:

On page 6 of Enclosure 1 to the letter dated June 23, 2020, the licensee states that they will perform either a volumetric exam or VT-1 and that the VT-1 examination is outlined in Code Case N-648-2, Alternative Requirements for Inner Radius Examinations of Class 1 Reactor Vessel Nozzles. The staff notes that the NRCs condition on Code Case N-702 in Regulatory Guide 1.147, Revision 19 requires the use of Code Case N-648-2 if VT-1 is used in place of the volumetric exam.

Request:

Confirm that Code Case N-648-2 will be used for VT-1 exams, in accordance with NRCs condition on Code Case N-702.

Response

When a VT-1 examination is used in place of a volumetric examination, ASME Code Case N-684-2, Alternative Requirements for Inner Radius Examinations of Class 1 Reactor Vessel Nozzles, will be used for the VT-1 examination in accordance with required conditions specified in Regulatory Guide 1.147, Revision 19.

RAI 4

Issue:

The staff noted that the probability of failure at the blend radius of the Brunswick recirculation outlet nozzle due to low temperature overpressure (LTOP), as shown in Table 12 of Enclosure 1 of the July 30, 2020 supplement, is much lower compared to the probability of failure at the blend radius in one of the nozzles (Columbia) analyzed in BWRVIP-241, BWR Vessel and Internals Project: Probabilistic Fracture Mechanics Evaluation for the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Blend Radii. The staff noted that the transient cycles for the Brunswick recirculation outlet nozzle (Table 5 of Enclosure 1 to the supplement) are similar to the transient cycles for the Columbia recirculation outlet nozzle (Table 5-5 of BWRVIP-241). The staff also compared stresses and noted that the stresses at the Brunswick recirculation outlet nozzle blend radius are higher (Figures 10 and 11 of Enclosure 1 to the supplement) than those at the Columbia recirculation outlet nozzle blend radius (Figures 4-44 through 4-47 of BWRVIP-241). Given that the random parameters are the same for both cases (Table 11 of Enclosure 1 to the supplement for Brunswick and Table 5-1 of BWRVIP-241, Case 5, for Columbia), the staff expected that the probability of failure for Brunswick would be slightly higher than the probability of failure for Columbia. However, the LTOP probability of failure at the blend radius for Brunswick is five orders of magnitude lower than for the corresponding case for Columbia (Table 12 of Enclosure 1 to the supplement for Brunswick compared to Table 5-9 of BWRVIP-241 for Columbia).

RA-20-0324 Enclosure Page 5 of 7 Request:

Explain why the probability of failure at the nozzle blend radius of the Brunswick recirculation outlet nozzle due to LTOP is so much lower than the corresponding probability of failure for the Columbia recirculation outlet nozzle.

Response

The inside surface stresses for the Brunswick recirculation outlet nozzle blend radius (Figures 10 and 11 of Enclosure 1 to the supplement) are higher than those for the Columbia recirculation outlet nozzle blend radius (Figures 4-44 through 4-47 of BWRVIP-241). However, in determining the crack driving force (the stress intensity factor (K)) through the thickness of the blend radius, the shape of the through-wall stress distribution is an important factor. It can be seen from the through-wall stress distributions for the thermal transients that the stresses for Brunswick although higher on the inside surface, decays at a faster rate than those at Columbia and this could alter the K distribution through the wall. Using the through-wall stress intensity factor distributions calculated in the VIPER-NOZ output files for Columbia from BWRVIP-241-A and from the Brunswick evaluation, the K distribution for the limiting transient at Columbia (Pressure + SRV Blowdown) is compared to the limiting case (Path 1) at Brunswick (Pressure +

Improper Start of Recirc. Loop) in the figure below for the nozzle blend radius.

Stress Intensity Factor of Limiting Transient + Unit Pressure 140 120 100 K (ksi-in) 80 60 40 Columbia - SRV + P 20 Brunswick - ISCRL-M + P 0

0 1 2 3 4 5 6 7 8 9 10 Depth (inch)

RA-20-0324 Enclosure Page 6 of 7 It can be seen from this figure that the maximum K for Columbia (127.5 ksiin) is higher than that for Brunswick (110.1 ksiin) by 15%. For fatigue crack growth where the K is raised to a power of 2.927, this translates into crack growth of 50% higher for Columbia, and for stress corrosion crack growth where the K is raised to a power of 4.0, this translates into crack growth of 80% higher for Columbia. It is for this reason that the LTOP probability of failure for Columbia is higher than that for Brunswick.

The probabilities of failure for the Columbia recirculation outlet nozzle reported in Table 5-9 of BWRVIP-241-A are the conditional probabilities of failure. To calculate the probability of failure due to an LTOP event, these conditional probabilities are multiplied by the probability of an LTOP event occurrence (1x10-3). For normal operating condition, the probability of failure is the same as the conditional probability of failure. For no failure (NF) results without tabulated values, the conditional probability of failure is calculated as less than 1 failure in the total number of simulations (<1 failure / 1,000,000 simulations / 40 years = 2.5 x 10-8). The total number of simulations is assumed to be one million, which is consistent with recirculation outlet nozzle simulations in Table 5-4 of BWRVIP-108-A.

The table below compares the probability of failure for an LTOP event and the probability of failure for normal operation at the nozzle blend radius and the nozzle-to-shell welds for the recirculation outlet nozzles for Columbia from Table 5-9 of BWRVIP-241-A and Brunswick in Table 12 of Enclosure 1 of the July 30, 2020 supplement. The number of failures in the simulations are also provided in the table, and both evaluations performed one million simulations.

For the nozzle-to-shell welds, there were no failures in the recirculation outlet nozzle for both Columbia and Brunswick, and the probabilities of failure are comparable and are dependent on the total number of simulations (no failures in 1 million simulations) and evaluated plant life (40 years and 60 years for Columbia and Brunswick, respectively).

For the nozzle inside radius, the higher LTOP probability of failure of Columbia with failures compared to Brunswick with no failures is due to the reason provided above.

Number of Failures Probability of Failure (1 million simulations) (per year)

Crack Flaw Component a/l Condition Columbia Brunswick Columbia Brunswick Model Density Recirculation Recirculation Recirculation Recirculation Outlet Outlet Outlet Outlet Nozzle LTOP Not specified NF 6.83x10-9 <1.67x10-11 Blend Inside NA 0.1 Radius NO NF NF <2.5x10-8 <1.67x10-8 Radius LTOP NF NF <2.5x10-11 <1.67x10-11 Axial 1/2 1 Nozzle-to- NO NF NF <2.5x10-8 <1.67x10-8 Shell Weld LTOP NF NF <2.5x10-11 <1.67x10-11 Circ. 1/2 1 NO NF NF <2.5x10-8 <1.67x10-8 NF = No failure in the simulations NO = Normal Operating Condition

RA-20-0324 Enclosure Page 7 of 7

RAI 5

Issue:

The licensee stated that the neutron fluence projections or evaluation for the BSEP reactor pressure vessel nozzle-to-vessel welds and inner radii over the period of extended operation (54 effective full power years) can be found from RR, WCAP-17660 (Reference 3) and Pressure-Temperature Limits report elated documents. The staff uses the neutron fluence values as reported in WCAP-17660 to infer the neutron fluence at the recirculation outlet nozzles below:

Neutron Fluence Projections at 54 EFPY (WCAP-17660)

Unit 1 Unit 2 Girth Weld FG 1.0x1018 ~ 2.8x1018 n/cm2 0.9x1018 ~ 2.9x1018 n/cm2 (254 AVO*) (Table 2.2-17) (Table 2.2-42)

H6A 4.7x1017 ~ 1.9x1018 n/cm2 4.7x1017 ~ 2.0x1018 n/cm2 (183 AVO) (Table 2.2-6) (Table 2.2-31)

H6B 1.8x10 ~ 7.2x10 n/cm 17 17 2 1.8x1017 ~ 7.5x1017 n/cm2 (179 AVO) (Table 2.2-7) (Table 2.2-32)

Girth Weld GH 0.8x1012 ~ 2.3x1012 n/cm2 0.8x1012 ~ 2.3x1012 n/cm2 (110 AVO) (Table 2.2-16) (Table 2.2-41)

• AVO = Above Vessel Zero Based on Figure 2.1-4 of WCAP-17660, the recirculation outlet nozzles are located between the elevations of Girth Weld GH and Girth Weld FG. The actual elevation for recirculation outlet nozzle center is 161 AVO (Reference 4). By adding up the outer radius of the nozzle, 24, to the center elevation, it appears that the upper 25% of the nozzle would be exposed to a neutron fluence > 1.0x1017 n/cm2 at 54 EFPY.

Request:

Provide additional information or justification for the conclusion that the neutron fluence for the welds between the BSEP reactor pressure vessel and nozzles is less than 1.0x1017 n/cm2 at 54 EFPY.

Response

Additional information related to the fluence analysis for the BSEP reactor pressure vessel and nozzles is provided in Westinghouse Letter LTR-REA-13-19 (Attachment 1), Additional Request for RPV Fast Neutron Fluence at Brunswick Unit 1 and 2, February 27, 2013. This Westinghouse letter provides the life-time fluence for 54 Effective-Full-Power-Years (EFPY) of operation at the recirculation outlet nozzles (N1) at 0° and 180° azimuths; the top elevations of these nozzles are at 192.9 above vessel zero. The azimuthal profiles (since the reactor exhibits quadrant symmetry only a 0° to 90° sector is depicted) clearly indicate that the life-time fluence at N1 nozzles are below the extended beltline fluence threshold of 1.0x1017 n/cm2 (E >

1MeV) for Brunswick Unit 1 and Unit 2.

RA-20-0324 Attachment 1 Attachment 1 Attachment A of LTR-REA-13-19, Rev. 0, Additional Request for RPV Fast Neutron Fluence at Brunswick Units 1 and 2 (6 pages)

Westinghouse Non-Proprietary Class 3 Attachment A of Westinghouse Proprietary Class 2 LTR-REA-20-109, Rev. 0 Westinghouse Electric Company 1000 Westinghouse Drive Cranberry Township, Pennsylvania 16066 USA Mike Alford Direct tel: 412-374-5639 Major Projects Direct fax: 724-940-8565 Progress Energy - Brunswick Nuclear Plant e-mail: wangs@westinghouse.com 8470 River Road South Port, NC 28461 Our ref: LTR-REA-13-19 Date: February 27, 2013 Additional Request for RPV Fast Neutron Fluence at Brunswick Units 1 and 2

References:

1 - Westinghouse Report WCAP-17660-NP, Revision 0, Neutron Exposure Evaluations for Core Shroud and Pressure Vessel Brunswick Units 1 and 2, November, 2012.

2 - D. Alford to S. Wang et al, Additional Design Info Needed to Evaluate BNP Vessel Fracture Toughness, February 19, 2013.

The purpose of this transmittal is to respond to the request of additional information described in Reference 2. The information requested are related to the Brunswick Units 1 and 2 fluence analysis performed by Westinghouse as documented in Reference 1.

The following graphic presentations are attached to this transmittal:

Figure 1 displays the life-time fluence for 54 Effective-Full-Power-Years (EFPY) of operation at the recirculation outlet nozzles (N1) at 0° and 180° azimuths; the top elevations of these nozzles are at 192.9 above vessel zero. The azimuthal profiles in Figure 1 clearly indicate that the life-time fluence at N1 nozzles are below the extended beltline fluence threshold of 1e17 n/cm2 for Brunswick Unit 1 and Unit 2.

Figure 2 displays similar life-time fluence profiles at the recirculation inlet nozzles (N2) located at 30°, 60° and 90° of each quadrant; the top elevations of these nozzles are at 196.4 above vessel zero. Figure 2 demonstrates that the life-time fluences at N2 nozzles are also below the extended beltline fluence threshold of 1e17 n/cm2 for both Units 1 and 2.

Figure 3 illustrates the fluence profile above girth weld EF, axially along the vertical welds E1 and E2 at 75° and 255° azimuth, respectively. The elevations of girth welds as well as the life-time fluence at each girth weld have been documented in Reference 1. Figure 3 indicates that slightly (less than 1 cm) above girth weld EF or at 481 cm above the bottom of active fuel (BAF), the life-time fluences of vertical welds E1 and E2 will drop below the extended beltline threshold of 1e17 n/cm2.

This document is the property of and contains Proprietary Information owned by Westinghouse Electric Company LLC and/or its subcontractors and suppliers. It is transmitted to you in confidence and trust, and you agree to treat this document in strict accordance with the terms and conditions of the agreement under which it was provided to you.

©2012 Westinghouse Electric Company LLC All Rights Reserved

• • • This record was final approved on 10/22/2020 6:59:04 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 Westinghouse Proprietary Class 2 Attachment A of LTR-REA-20-109, Rev. 0 Page 2 of 6 Our ref: LTR-REA-13-19 Figure 4 presents the axial location above the girth weld EF where the life-time fluence is expected to drop below the beltline threshold of 1e17 n/cm2. Since the fast neutron flux/fluence level at the RPV surface varies with azimuth; the axial elevation where the extended beltline threshold occurs also varies with azimuth, as demonstrated in Figure 4.

Please contact the undersigned if there are any questions or comments.

Author: Reviewer:

Sylvia S. Wang

• Stanwood L. Anderson*

Radiation Engineering and Analysis Radiation Engineering and Analysis Approved By Laurent P. Houssay

• Manager, Radiation Engineering and Analysis

• Electronically Approved Records Are Authenticated in the Electronic Document Management System

• • • This record was final approved on 10/22/2020 6:59:04 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 Westinghouse Proprietary Class 2 Attachment A of LTR-REA-20-109, Rev. 0 Page 3 of 6 Our ref: LTR-REA-13-19 Brunswick 54 EFPY RPV Fluence (n/cm2 E > 1 MeV) at 192.9" above Vessel 0 1.2E+17 U1 U2 1.0E+17 8.0E+16 Fast Neutron Fluence (E > 1 MeV) 6.0E+16 N1 nozzle 4.0E+16 2.0E+16 0.0E+00 0 10 20 30 40 50 60 70 80 90 Azimuth Degree Figure 1 Brunswick Units 1 & 2 Lifetime Fluence at N1 Recirculation Outlet Nozzle

• • • This record was final approved on 10/22/2020 6:59:04 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 Westinghouse Proprietary Class 2 Attachment A of LTR-REA-20-109, Rev. 0 Page 4 of 6 Our ref: LTR-REA-13-19 Brunswick 54 EFPY RPV Fluence (n/cm2 E > 1 MeV) at 196.4" above Vessel 0 1.8E+17 U1 U2 1.6E+17 1.4E+17 1.2E+17 Fast Neutron Fluence (E > 1 MeV) 1.0E+17 8.0E+16 N2 nozzle N2 nozzle 6.0E+16 N2 nozzle 4.0E+16 2.0E+16 0.0E+00 0 10 20 30 40 50 60 70 80 90 Azimuth Degree Figure 2 Brunswick Units 1 & 2 Lifetime Fluence at N2 Recirculation Inlet Nozzle

• • • This record was final approved on 10/22/2020 6:59:04 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 Westinghouse Proprietary Class 2 Attachment A of LTR-REA-20-109, Rev. 0 Page 5 of 6 Our ref: LTR-REA-13-19 Brunswick Fluence (n/cm2 E > 1 MeV) at E1 & E2 Welds 75° Azimuth 1.2E+17 U1 U2 1.0E+17 54 EFPY Fluence (E > 1 MeV) 8.0E+16 6.0E+16 4.0E+16 480 485 490 495 Distance above BAF (cm)

Figure 3 Brunswick Units 1 & 2 Lifetime Fluence at Vertical Weld of Upper-Intermediate Shell

©2012 Westinghouse Electric Company LLC All Rights Reserved

• • • This record was final approved on 10/22/2020 6:59:04 AM. (This statement was added by the PRIME system upon its validation)

Westinghouse Non-Proprietary Class 3 Westinghouse Proprietary Class 2 Attachment A of LTR-REA-20-109, Rev. 0 Page 6 of 6 Our ref: LTR-REA-13-19 Brunswick Upper Shell Elevation with 54 EFPY Fluence Under 1e17 n/cm2 486 U1 U2 484 482 Distance above BAF (cm) 480 478 476 474 472 0 10 20 30 40 50 60 70 80 90 Azimuth Degree Figure 4 Brunswick Units 1 & 2 Upper Shell Elevation with Beltline Threshold Fluence

• • • This record was final approved on 10/22/2020 6:59:04 AM. (This statement was added by the PRIME system upon its validation)