ML24302A144
| ML24302A144 | |
| Person / Time | |
|---|---|
| Site: | Summer  |
| Issue date: | 10/24/2024 |
| From: | James Holloway Dominion Energy South Carolina |
| To: | Office of Nuclear Reactor Regulation, Document Control Desk |
| References | |
| 24-309 | |
| Download: ML24302A144 (1) | |
Text
Dominion Energy Souih Carolina, Inc.
5000 Dominion Boulevard, Glen Allen, VA 23060 DominionEnergy.com October 24, 2024 United States Nuclear Regulatory Commission Attention: Document Control Desk 11555 Rockville Pike Rockville, MD 20852 DOMINION ENERGY SOUTH CAROLINA, INC.
VIRGIL C. SUMMER NUCLEAR STATION UNIT 1 Dominion
~
Energy" Serial No.:
NRA/NDM:
Docket No.:
License No.:
10 CFR 50 10 CFR 51 10 CFR 54 24-309 RO 50-395 NPF-12 UPDATE TO SUBSEQUENT LICENSE RENEWAL APPLICATION (SLRA)
SUPPLEMENT 4 AND REQUESTED INFORMATION IN RESPONSE TO LIMITED AGING MANAGEMENT AUDIT By letter dated August 17, 2023 [Agencywide Documents Access and Management System (ADAMS) Package Accession No. ML23233A179], Dominion Energy South Carolina, Inc. (Dominion, Dominion Energy South Carolina, or DESC) submitted an application for the subsequent license renewal of Renewed Facility Operating License No. NPF-12 for Virgil C. Summer Nuclear Station (VCSNS) Unit 1.
The NRG staff conducted an Aging Management Audit from November 6, 2023, to March 21, 2024, as part of their review of the VCSNS SLRA. As a result of discussions with the NRC staff throughout the audit, additional information which is necessary for the NRC staff to complete their technical review was identified. Some of the associated updates to the SLRA were provided in the Supplement 1 letter, dated April 1, 2024 [ADAMS Accession No. ML24095A207], Supplement 2 letter, dated May 6, 2024 [ADAMS Accession No. ML24129A200], and Supplement 3 letter, dated May 30, 2024 [ADAMS Accession No. ML24155A146].
Subsequently, the NRG staff conducted a Limited Aging Management Audit beginning on April 24, 2024. As a result of discussions with the NRC staff throughout this audit, further information which is necessary for the NRC staff to complete their technical review was identified for one topic. Additional updates to the SLRA are provided as Supplement 4 in Enclosures 1 and 2 to this letter.
During the Limited Aging Management Audit, the NRC staff also indicated that additional information related to three topics was needed to support writing the Safety Evaluation Report. DESC agreed to provide the required additional information as part of this supplement letter. The three topics and information requested are provided in Enclosure
- 3. provides a description of the topic that requires the SLRA to be supplemented and identifies the affected SLRA section and table for Supplement 4. Enclosure 2 includes
Serial No.: 24-309 Docket No.: 50-395 Page 2 of 6 mark-ups of affected SLRA sections and tables for Supplement 4, as described in. Enclosure 3 provides NRG-requested information in response to the Limited Aging Management Audit.
To aid the staff in assessing changes, Enclosure 2 shows new text as underlined and deleted text as lined-through. Please note that change bars are shown for new text but are not shown for deleted text.
If there are any questions regarding this submittal or if additional information is needed, please contact Mr. Keith Miller at (804) 273-2569.
Sincerely, James E. Holloway Vice President - Nuclear Engineering and Fleet Support CRAIG D SLY COMMONWEAL TH OF VIRGINIA COUNTY OF HENRICO Notary Public
}
Commonwealth of Virgin16 i
Reg. # 7518653
?-Ii*
My Commission Expires December :3"i ~0- l
~
The foregoing document was acknowledged before me, in and for the County and Commonwealth aforesaid, today by James E. Holloway, who is Vice President - Nuclear Engineering and Fleet Support of Dominion Energy South Carolina, Inc. He has affirmed before me that he is duly authorized to execute and file the foregoing document on behalf of that Company, and that the statements in the document are true to the best of his knowledge and belief.
Acknowledged before me this 24-fh day of octbbe C ' 2024.
My Commission Expires: _1_t_,_/~3--11 /~z._'-/ ___ _
Commitments made in this letter:
The Licensee Commitments identified in Table A4.0-1 of Appendix A, FSAR Supplement, are proposed to support approval of the subsequent renewed operating licenses and may change during the NRC review period.
Enclosures:
- 1. Topic that Requires a SLRA Supplement - Supplement 4
- 2. SLRA Mark-ups - Supplement 4
- 3. Requested Information in Response to Limited Aging Management Audit
cc:
U.S. Nuclear Regulatory Commission, Region II Marquis One Tower 245 Peachtree Center Avenue, NE Suite 1200 Atlanta, Georgia 30303-1257 NRC Senior Resident Inspector Virgil C. Summer Nuclear Station Ms. Lauren Gibson NRC Branch Chief U. S. Nuclear Regulatory Commission One White Flint North Mail Stop 11 E11 11555 Rockville Pike Rockville, Maryland 20852-2738 Ms. Kim Conway NRC Project Manager U.S. Nuclear Regulatory Commission Two White Flint North Mail Stop A 1 OM 11555 Rockville Pike Rockville, Maryland 20852-2738 Mr. Ted Smith NRC Branch Chief U.S. Nuclear Regulatory Commission Two White Flint North Mail Stop B 72M 11555 Rockville Pike Rockville, Maryland 20852-2738 Ms. Marieliz Johnson NRC Project Manager U. S. Nuclear Regulatory Commission One White Flint North Mail Stop 7 E 1 11555 Rockville Pike Rockville, Maryland 20852-2738 Serial No.: 24-309 Docket No.: 50-395 Page 3 of 6
Mr. G. Edward Miller NRG Senior Project Manager U. S. Nuclear Regulatory Commission One White Flint North Mail Stop O-9E3 11555 Rockville Pike Rockville, Maryland 20852-2738 Mr. Gregory Lindamood Santee Cooper - Nuclear Coordinator Virgil C. Summer Nuclear Station P. 0. Box 88 Jenkinsville, SC 29065 Stephen R. Pelcher Santee Cooper - Deputy General Counsel One Riverwood Drive Moncks Corner, SC 29461 Elizabeth Johnson Director, Historical Services, D-SHPO South Carolina Department of Archives and History 8301 Parklane Road Columbia, SC 29233 Johnathan Leader State Archaeologist South Carolina Institute of Archaeology and Anthropology 1321 Pendleton St, 1st Floor, Suite 16 Columbia, SC 29208 Fran Marshall Director, Office of Environmental and Public Health Serial No.: 24-309 Docket No.: 50-395 Page 4 of 6 South Carolina Department of Health and Environmental Control 2600 Bull Street Columbia, SC 29201 Myra Reese Director, Office of Environmental Affairs South Carolina Department of Health and Environmental Control 2600 Bull Street Columbia, SC 29201
Serial No.: 24-309 Docket No.: 50-395 Chuck Hightower Water Quality Division South Carolina Department of Health and Environmental Control 2600 Bull Street Columbia, SC 29201 Lorianne Riggin Director of Environmental Programs South Carolina Department of Natural Resources 1000 Assembly Street Columbia, SC 29201-3117 Duane Parrish Director, South Carolina Parks, Recreation, and Tourism 1205 Pendleton St. Ste 248 Columbia, SC 29201 Florence Belser Chair, Public Service Commission of South Carolina 101 Executive Center Drive #100 Columbia, SC 29210 Christy Hall Secretary of Transportation South Carolina Department of Transportation 955 Park Street Columbia, SC 29201 Kim Stenson Director, South Carolina Emergency Management Division 2779 Fish Hatchery Rd West Columbia, SC 29172 Rhonda Thompson Chief, Bureau of Air Quality South Carolina Department of Health and Environmental Control 2600 Bull Street Columbia, SC 29201 Renee Shealy Chief, Bureau of Environmental Health Services South Carolina Department of Health and Environmental Control 2600 Bull Street Columbia, SC 29201 Page 5 of 6
Serial No.: 24-309 Docket No.: 50-395 Henry Porter Chief, Bureau of Land and Waste Management South Carolina Department of Health and Environmental Control 2600 Bull Street Columbia, SC 29201 Jennifer Hughes Chief, Bureau of Water South Carolina Department of Health and Environmental Control 2600 Bull Street Columbia, SC 29201 Chris Stout Chief, Office of Ocean & Coastal Resource Management South Carolina Department of Health and Environmental Control 1362 McMillan Avenue North Charleston, SC 29405 Page 6 of 6 TOPIC THAT REQUIRES A SLRA SUPPLEMENT SUPPLEMENT 4 Dominion Energy South Carolina, Inc.
(Dominion Energy South Carolina, or DESC)
Virgil C. Summer Nuclear Station Unit 1 Serial No.: 24-309 Docket No.: 50-395
The following topic requires the SLRA to be supplemented:
Serial No.: 24-309 Docket No.: 50-395, Page 2 of 2
- 1. Reactor Vessel Supports and Primary Shield Wall: SLRA Updates In response to the Limited Aging Management Audit, additional information is provided regarding reactor vessel (RV) supports and the primary shield wall (PSW). A new line for "steel elements (inaccessible/embedded): primary shield wall" is added to Tables 2.4.1-1 and 3.5.2-1 and associated new note 13 is added to reference the Section 3.5.2.2.2.6 analysis. New note 14 of Table 3.5.2-1 is added to clarify that the PSW concrete item includes grout associated with the RV supports. The concrete environment is added to Section 3.5.2.1.1 for the Reactor Building. The "reduction in fracture toughness" and "loss of intended function" aging effects are added to Section 3.5.2.1.15, Table 3.5.2-15 and its new note 4, and Sections A 1.32 and 82.1.32. A description of existing activities to monitor for potential degradation of the PSW is added to Section 3.5.2.2.2.6, and consideration of these activities is discussed in revised note 12 of Table 3.5.2-1. New note 5 is added to Table 3.5.2-15 for consistency with the new RV support inspection description in Section 3.5.2.2.2.6. An enhancement to inspect at least one reactor pressure vessel support every five years is referenced in Section 3.5.2.2.2.6 and added to Table A4.0-1 #32 and Section 82.1.32. Structures Monitoring Program (82.1.35) is added to the PSW line in Table 3.5.2-1.
Based on the above, the SLRA is supplemented as shown in Enclosure 2, to provide further information which is necessary for the NRG staff to complete their technical review, as shown in the following:
SLRA Section SLRA Table 3.5.2.1.1 2.4.1-1 3.5.2.1.15 3.5.2-1 3.5.2.2.2.6 3.5.2-15 A1.32 A4.0-1 #32 82.1.32 SLRA MARK-UPS SUPPLEMENT 4 Dominion Energy South Carolina, Inc.
(Dominion Energy South Carolina, or DESC)
Virgil C. Summer Nuclear Station Unit 1 Serial No.: 24-309 Docket No.: 50-395
Page 2 of 17 Supplement 4 Table 2.4.1-1 Reactor Building Structural Member Bolting Concrete elements Containment liner Equipment hatch, personnel access air lock, personnel escape air lock, and accessories (hinges, pins, closure mechanisms)
Moisture barriers Penetrations (electrical)
Penetrations (mechanical)
Primary shield wall Refueling cavity/fuel transfer canal liner Seals and gaskets Service Level I coatings Steel elements Steel elements (inaccessible/
embedded): 12rimarv shield wall Sump liners Tendon anchorage components Tendons See Table 2.1-1 for definitions of intended functions.
Virgil C. Summer Nuclear Station Application for Subsequent License Renewal Structures Intended Function(s)
Pressure Boundary, Structural Support Enclosure Protection, Fire Barrier, Flood Barrier, Jet Impingement Shield, Missile Barrier, Pressure Boundary, Structural Support Pressure Boundary, Structural Support Enclosure Protection, Fire Barrier, Missile Barrier, Pressure Boundary, Structural Support Enclosure Protection Pressure Boundary, Structural Support Pressure Boundary, Structural Support Enclosure Protection, Structural Support Water Barrier, Pressure Boundary, Structural Support Pressure Boundary Coating Integrity Enclosure Protection, Flood Barrier, Jet Impingement Shield, Missile Barrier, Structural Support Structural Support Pressure Boundary, Structural Support Structural Support Structural Support Page2-250
Supplement 4 Page 3 of 17 Virgil C. Summer Nuclear Station Application for Subsequent License Renewal Aging Management Review 3.5.2.1 Materials, Environments, Aging Effects Requiring Management and Aging Management Programs 3.5.2.1.1 Reactor Building Materials The materials of construction for the Reactor Building structural members are:
- Concrete
- Elastomer, rubber and other similar materials
- Reinforced concrete
- Stainless steel
- Steel Environment The Reactor Building structural members are exposed to the following environments:
- Air
- Air - indoor uncontrolled
- Air - outdoor
- Air with borated water leakage
- Concrete
- Groundwater
- Soil
- Treated borated water
- Water - flowing
- Water - standing Page 3-552
Supplement 4 3.5.2.1.15 Materials NSSS Supports Page 4 of 17 Virgil C. Summer Nuclear Station Application for Subsequent License Renewal Aging Management Review The materials of construction for the NSSS Supports subcomponents are:
- High-strength steel
- Lubrite
- Stainless steel
- Steel Environment The NSSS Supports subcomponents are exposed to the following environments:
- Air
- Air - indoor uncontrolled
- Air with borated water leakage Aging Effects Requiring Management The following aging effects, associated with the NSSS Supports subcomponents, require management:
- Cracking
- Loss of intended function
- Loss of material
- Loss of mechanical function
- Loss of preload
- Reduction in concrete anchor capacity
- Reduction in fracture toughness Aging Management Programs The following aging management programs manage the aging effects for the NSSS Supports subcomponents:
- ASME Section XI, Subsection IWF (82.1.32)
- 8oricAcid Corrosion (82.1.4)
- Structures Monitoring (82.1.35)
Page 3-577
Supplement 4 Page 5 of 17 Virgil C. Summer Nuclear Station Application for Subsequent License Renewal Aging Management Review potential radiation damage, the PSW demand is almost wholly dictated by the RV support load in this region.
The PSW contains embedded RV support anchorages. The maximum fluence at 72 EFPY on the RV support anchorage embedded steel (within the PSW), adjusted to add 20% for analytical uncertainty, is 8.82 x 1017 n/cm2 (E > 1.0 MeV). This is less than the damage threshold for the steel of 1.00 x 1019 n/cm2 {E > 1.0 MeV), per EPRI Report 3002013084.
The conservatisms in the evaluation were as follows:
- Exposures were based on 72 EFPY
- Future projections included a 10% positive bias on the peripheral and re-entrant corner assemblies on the projection fuel cycle.
- The loss of strength in the PSW concrete as a result of gamma dose incident on the PSW was assumed to apply to the full thickness to the point where the gamma dose falls below the NUREG-2192 damage threshold, when in reality the gamma dose effect would reduce in an approximately linear fashion from the outside surface.
- The latest research data presented in EPRI Report 3002011710 indicated that the threshold for damage to concrete from gamma dose may be higher than 1 x 108 Gy.
- The IR does not include any reduction in PSW demand to account for LBB implementation.
As evidenced by the evaluation described above, and considering the integrated effects of neutron fluence, gamma dose, and gamma heating, the PSW is capable of carrying the loads of the RV at the end of 80 years of plant operation. Therefore, the PSW will continue to satisfy its design criteria considering the long-term radiation effects and a plant specific AMP or enhancements to an existing AMP is not required. However, exisiting activities are also conducted to monitor for potential degradation as follows:
- VT-3 inspections of the six RV supports every ten years as directed by the ISi program.
- Review of the RV supports inspection results under the Structures Monitoring program for conditions that may require evaluation of the primary shield wall concrete and grout.
- Inspection of the incore pit room (area under the RV) twice per refueling outage. If evidence of degradation is noted (such as cementitious debris), a Condition Report is initiated in the Corrective Action Program for evaluation.
- Monitoring of the condition on the outside of the PSW concrete every five years under the Structures Monitoring program.
In addition, Dominion has made an enhancement to the IWF program to require that at least one RV support will be inspected every five years. Taken together, the analytical assessment of the Page 3-614
Supplement 4 Page6 of 17 Virgil C. Summer Nuclear Station Application for Subsequent License Renewal Aging Management Review primary shield wall along with these on-going activities ensure the effects of aging will be adequately managed through the subsequent period of extended operation.
Gamma Heating The impact of gamma heating on the PSW has been evaluated and it was concluded that the maximum PSW concrete temperature would be less than 145°F. This temperature is bounded by the long-term PSW concrete temperature limit of 150°F reported in Section 3.8.1.5.1.2 of the FSAR.
Therefore, gamma heating is not an issue for the PSW concrete.
Secondary Shield Wall Evaluation The neutron fluence and gamma dose threshold limits of NUREG-2192 are not exceeded beyond the first 10 inches of the PSW. The SSW is physically external to the PSW. The entire SSW is further from the core at all points compared to the PSW, meaning that the neutron fluence and gamma dose is higher in the PSW than the SSW. Therefore, since the NUREG-2192 threshold limits are not exceeded external to the PSW and the SSW is physically external to the PSW, the neutron and gamma dose limits in the SSW are lower than the NUREG-2192 threshold limits. Since the NUREG-2192 threshold limits are not exceeded in the SSW, separate analysis of the SSW is not required.
Reactor Vessel Steel Support Evaluation The NRG previously identified radiation embrittlement of the RV supports as a generic safety issue (GSl-15). The NRG resolved the issue, as documented in NUREG-1509 on the basis of a risk-informed evaluation, without imposing new requirements on licensees. The review concluded that loss of fracture toughness due to irradiation embrittlement will not affect the ability of the RV structural steel to perform its component intended functions through the original design life of the plant. However, this review was not performed for an 80-year plant life. Accordingly, a review of the aging effect of reduction in fracture toughness due to embrittlement from exposure to neutron fluence of the RV support steel was performed for subsequent license renewal in WGAP-18785-NP.
NUREG-1509, "Radiation Effects on Reactor Pressure Vessel Supports," provides a screening evaluation approach to evaluate the loss of fracture toughness of RV supports due to radiation effects for long term operation. The screening evaluation includes the following criteria for assessing the structural integrity of the RV steel supports:
- The initial nil ductility temperature (NOT) of the RV supports is well below the minimum operating temperature.
- The radiation exposure at the supports is low.
- The peak tensile stresses are 6 ksi or less.
Page 3-615
Supplement 4 Page 7 of 17 Virgil C. Summer Nuclear Station Application for Subsequent License Renewal Aging Management Review As discussed in Table 3.5.2.2.2.6-4, the support box plates and support shoe actual stresses are less than the critical stresses. As discussed in Table 3.5.2.2.2.6-5, the anchor bolts and hold down/guide pins critical flaw lengths are larger than the Section XI allowable flaw lengths.
The change in embrittlement from 42 EFPY to 72 EFPY for the support shoe and hold down/guide pins were considered to evaluate the change of radiation embrittlement with time. For the support plates and anchor bolts, the fracture mechanics analysis conservatively considers lower bound fracture toughness, which represents infinite embrittlement. Thus, only the support shoe and hold down/guide pins component specific fracture toughness was considered in the change of embrittlement evaluation. The evaluation demonstrates that the supports have sufficient flaw tolerance to not be impacted by neutron embrittlement from the original design life of 40 years to the SLR period of 80 years (see Section 8.2 of WCAP-18785-NP for additional discussion).
Based on these conclusions, the RV supports continue to be structurally stable (i.e., flaw tolerant) considering 80 years of radiation embrittlement effects on the supports. Additionally, no additional inspections or enhancements are required for aging management of the RV supports, and the current ASME Code,Section XI inspection requirements are sufficient; except that procedures will be revised to require that at least one RV support will be inspected every 5 years during the subsequent period of extended operation. Furthermore, the loads from the RV are transmitted to the PSW, and the PSW can accommodate the operating and design basis loads. Thus, the RV will continue to be adequately supported for 80 years of plant operation.
Operating Experience Per Section 4.3.1 of NUREG-1509, physical examination of the structural components is essential to the re-evaluation completed herein and an assessment of the overall condition of the RV support structure. Based on the RV support equipment specification, the structural steel components and welds had required examination per ASME Code, Section Ill, Appendix IX (radiography, liquid penetrant, magnetic particle, and ultrasonic testing). During initial fabrication, any unsatisfactory conditions were to be removed, re-welded, and re-examined. Thus, it is expected that the analyzed components are free from cracks after initial fabrication and after an extended period of time since crack growth mechanism are not present at the RV supports. The most recent RV support inspections were performed per ASME Code,Section XI, IWF guidance. During the inspections, evidence of dry boron residue was discovered based on remote visual inspection of the RV supports.
The boric acid deposits were dry and crystalline with no visual evidence of active leakage. This leakage was considered historical, as there is no active leakage from either the RCS pressure boundary or the refueling cavity. It was concluded that the visual inspection identified no indications, no significant surface degradations nor component damage on the RV support, and that no evidence appears rejectable per ASME Code,Section XI, IWF-3410. The RV supports are within the ASME Code,Section XI, ISi program and any further leaks and conditions that would affect the supports will be identified and periodically monitored.
Page 3-620
Page 8 of 17 Table 3.5.2-1 Containment Structure - Aging Management Evaluation Structural Intended Material Environment Aging Effect Requiring Aging Management Programs Member Function(s)
Management Penetrations PB:SS Dissimilar (E) Air - indoor Cracking 10 CFR Part 50, Appendix J (B2.1.33)
(mechanical) metal welds uncontrolled ASME Section XI, Subsection IWE (B2.1.30)
Cracking (CLB fatigue 10 CFR Part 50, Appendix J (B2.1.33) analysis does not exist)
ASME Section XI, Subsection IWE (B2.1.30)
Cumulative fatigue damage TLAA (Only if CLB fatigue analysis exists)
Loss of material 10 CFR Part 50, Appendix J (82.1.33)
ASME Section XI, Subsection IWE (82.1.30)
Stainless (E) Air-indoor Cracking 10 CFR Part 50, Appendix J (82.1.33) steel uncontrolled ASME Section XI, Subsection IWE (82.1.30)
Cracking (CLB fatigue 10 CFR Part 50, Appendix J (B2.1.33) analysis does not exist)
ASME Section XI, Subsection IWE (82.1.30)
Cumulative fatigue damage TLAA (Only if CLB fatigue analysis exists)
Steel (E) Air - indoor Cracking (CLB fatigue 10 CFR Part 50, Appendix J (B2.1.33) uncontrolled analysis does not exist)
ASME Section XI, Subsection IWE (82.1.30)
Cumulative fatigue damage TLAA (Only if CLB fatigue analysis exists)
Loss of material 10 CFR Part 50, Appendix J (82.1.33)
ASME Section XI, Subsection IWE (82.1.30)
(E) Air with borated Loss of material Boric Acid Corrosion (82.1.4) water leakage Primary shield EN:SS Concrete (E) Air-indoor Reduction of strength; loss of NooeStructures Monitoring (B2.1.35) wall uncontrolled mechanical properties Refueling BWl;PB:SS Stainless (E) Air Loss of material; cracking Structures Monitoring (82.1.35) cavity/fuel steel (E) Treated borated Cracking; loss of material Structures Monitoring (82.1.35) transfer canal water Water Chemistry (82.1.2) liner Virgil C. Summer Nuclear Station Page 3-647 Application for Subsequent License Renewal NUREG-2191 Table 1 Notes Item Item I1.A3.CP-38 3.5.1-010 A
I1.A3.CP-38 3.5.1-010 A
II.A3.CP-37 3.5.1-027 A
11.A3.CP-37 3.5.1-027 A
II.A3.C-13 3.5.1-009 A, 11 II.A3.CP-36 3.5.1-035 A
II.A3.CP-36 3.5.1-035 A
II.A3.CP-38 3.5.1-010 A
II.A3.CP-38 3.5.1 -010 A
II.A3.CP-37 3.5.1-027 A
II.A3.CP-37 3.5.1-027 A
11.A3.C-13 3.5.1-009 A, 11 II.A3.CP-37 3.5.1-027 A
II.A3.CP-37 3.5.1-027 A
II.A3.C-13 3.5.1-009 A, 11 I1.A3.CP-36 3.5.1 -035 A
II.A3.CP-36 3,5,1-035 A
111.B5.T-25 3,5,1-089 C
III.A4.T-35 3.5.1-097 6.,_
12,_
11 III.B2.T-37b 3,5,1-100 C
II1.AS.T-14 3.5.1 -078 A
II1.AS.T-14 3,5,1-078 A
Supplement 4
Page 9 of 17 Table 3.5.2-1 Containment Structure - Aging Management Evaluation Structural Intended Material Environment Aging Effect Requiring Aging Management Programs Member Function(s)
Management Seals and PB Elastomer, (E) Air - indoor Loss of sealing 10 CFR Part 50, Appendix J (B2.1.33) gaskets rubber and uncontrolled other similar materials Service Level I MCI Coatings (E) Air-indoor Loss of coating or lining Protective Coating Monitoring and Maintenance coatings uncontrolled integrity (B2.1.37)
Steel elements EN;FLB;JIS; Steel (E) Air-indoor Loss of material Structures Monitoring (B2.1.35)
MB;SS uncontrolled (E) Air with borated Loss of material Boric Acid Corrosion (B2.1 -4) water leakage Steel elements ss Steel (El Concrete Reduction in fracture Structures Monitoring (B2.1.35)
(inaccessible/
toughness; loss of intended embedded):
function grima(Y shield wall Sump liners PB;SS Dissimilar (E) Air-indoor Cumulative fatigue damage TLAA metal welds uncontrolled (Only if CLB fatigue analysis exists)
Stainless (E) Air-indoor Cumulative fatigue damage TLAA steel uncontrolled (Only if CLB fatigue analysis exists)
(E)Water-Cracking; loss of material ASME Section XI, Subsection IWE (B2.1.30) standing Tendon ss Steel (E) Air - outdoor Loss of material ASME Section XI, Subsection IWL (B2.1.31) anchorage components Tendons ss Steel (E) Air-indoor Loss of material ASME Section XI, Subsection IWL (B2.1.31) uncontrolled Loss of prestress TLAA Table 3.5.2-1 Plant-Specific Notes:
NUREG-2191 Table1 Notes Item Item II.A3.CP-41 3.5.1-033 A, 4 II.A3.CP-152 3.5.1-034 A
III.A4.TP-301 3.5.1-073 A
III.A4.TP-302 3.5.1-077 A, 5 III.B5.T-25 3.5.1-089 C, 5 None None H. 13 II.A3.C-13 3.5.1-009 A, 10 II.A3.C-13 3.5.1-009 A, 10 11I.A7.T-23 3.5.1-052 E, 9 II.A1.C-10 3.5.1-032 A
II.A1.C-10 3.5.1-032 A
II.A1.C-11 3.5.1-008 A
- 1.
Concrete elements include beams, columns. walls. slabs. curbs. foundations. pads. jet impingement barriers, missile barriers. tendon access gallery.
dome. ring girder and buttresses.
- 2.
Containment liner includes liner plates, liner leak chase channels, liner anchors, and integral attachments.
Virgil C. Summer Nuclear Station Page 3-648 Supplement 4 Application for Subsequent License Renewal
Page 10 of 17
- 3.
Moisture barriers includes elastomeric moisture barrier between Containment liner and internal concrete basement floor.
- 4.
Seals and gaskets include O-rings and other elastomer materials that are part of the Containment pressure boundary.
- 5.
Steel elements include beams, columns, baseplates, bracing, stairs, platforms, grating, decking, ladders, missile barriers, and embedded steel.
- 6.
Stainless steel bolting is associated with electrical penetrations.
- 7.
The ASME Section XI, Subsection IWE (B2.1.30) program and 10 CFR Part 50, Appendix J (82.1.33) program have been substituted for the Structures Monitoring (82.1.35) program to manage the applicable aging effect(s) for this component type, material, and environment combination.
- 8.
The ASME Section XI, Subsection IWL (82.1.31) program has been substituted for the Structures Monitoring (82.1.35) program to manage the applicable aging effect(s) for this component type, material, and environment combination.
- 9.
The plant-specific aging management program used to manage the applicable aging effect(s) for this component type, material, and environment combination is the ASME Section XI, Subsection IWE (82.1.30) program.
- 10. The sump liners are part of the containment pressure boundary.
- 11. Applicable to the main steam containment penetrations.
- 12. As discussed in Section 3.5.2.2.2.6, analysis has determined that reduction of strength and loss of mechanical properties due to irradiation will not impact the primary shield wall's ability to perform its intended function under design basis conditions. Nevertheless, several existing activities that are described in Section 3.5.2.2.2.6, are conducted to monitor for aging effects. The other aging effects that are assigned to "Concrete elements" with an air-indoor uncontrolled environment are also applicable to the primary shield wall concrete. As-4i-sst:J-&Sed in Section 3.5.2.2.2.6, analysis has determined that reductioA-Gf strength and loss o shield wall's intended function eleftm~RVW:~~- a+r--1-A(~->G
. ff-1:ls-~:f-F a-n:-~~w, 81
,,e~~~-Ed-WIR-',l,/--1-,11-r:G-1::i-meA1f-af:e-fHS(;l-al:}9fiK;a,9-10-ffi--H3/4H)!Rffiafll'-SFHe+G-IP.ia+1-GG~i:et1a,.
s that are as-signed to "Concrete
- 13. As discussed in Section 3.5.2.2.2.6, evaluation has determined that reduction in fracture toughness/loss of intended function of embedded steel due to neutron irradiation, and effects associated with reduction of strength of the concrete due to radiation will not impact the ability of the primary shield wall's inaccessible steel elements to perform their intended functions under design basis conditions. Therefore, aging management activities are not required to manage these aging effects. Should future information (e.g., industry and/or plant-specific operating experience) indicate the need for aging management of these aging effects, those activities would be addressed by the Structures Monitoring program.
Virgil C. Summer Nuclear Station Page 3-649 Supplement 4 Application for Subsequent License Renewal
Page 11 of 17 Table 3.5.2-15 Structures and Component Supports - NSSS Supports -Aging Management Evaluation Structural Intended Material Environment Aging Effect Requiring Aging Management Programs NUREG-2191 Member Function(s)
Management Item Bolling ss High-strength (E)Air Cracking ASME Section XI, Subsection IWF (62.1.32)
III.B1.1.TP-41 steel Loss of preload ASME Section XI, Subsection IWF (62.1.32) 111.61.1.TP-229 Stainless (E)Air Loss of material; cracking ASME Section XI, Subsection IWF (62.1.32) lll.81.1.T-36b steel Loss of preload ASME Section XI, Subsection IWF (62.1.32) 111.81.1.TP-229 Steel (E) Air - indoor Loss of material ASME Section XI, Subsection IWF (B2.1.32) 111.81.1.TP-226 uncontrolled Loss of preload ASME Section XI, Subsection IWF (B2.1.32) 111.81.1.TP-229 (E) Air with borated Loss of material Boric Acid Corrosion (82.1.4)
II1.B1.1.T-25 water leakage Grout ss Grout (E) Air - indoor Reduction in concrete Structures Monitoring (82.1.35) 111.81.1.TP-42 uncontrolled anchor capacity Sliding surfaces ss Lubrite (E) Air - indoor Loss of mechanical function ASME Section XI, Subsection IWF (B2.1.32)
II1.B1.1.TP-45 uncontrolled Spring hangers; ss Steel (E) Air - indoor Loss of mechanical function ASME Section XI, Subsection IWF (B2.1.32)
I11.B1.1.T-28 guides; stops uncontrolled Reduction in fracture ASME Section XI, Subsec!ion IWF (82.1.32)
None toughness, loss of intended function (E) Air with borated Loss of material Boric Acid Corrosion (82.1.4) 111.B1.1.T-25 water leakage Stainless steel ss Stainless (E) Air Loss of material; cracking ASME Section XI, Subsection IWF (B2.1.32)
III.B1.1.T-36b elements steel Steel elements ss Steel (E) Air - indoor Loss of material ASME Section XI, Subsection IWF (B2.1.32) 111.81.1.T-24 uncontrolled (E) Air with borated Loss of material Boric Acid Corrosion (B2.1.4) 111.B1.1.T-25 water leakage Table 3.5.2-15 Plant-Specific Notes:
- 1.
Stainless steel elements include support members.
- 2.
Steel elements include support members, bearing plates, baseplates, and connections.
Table1 Notes Item 3.5.1-068 6
3.5.1-087 6
3.5.1-099 8
3.5.1-087 8
3.5.1-081 8
3.5.1-087 8
3.5.1-089 A
3.5.1-055 A
3.5.1-075 6, 3 3.5.1-057 8
None
~
Q 3.5.1-089 A
3.5.1-099 8, 1 3.5.1-091 B, 2 3.5.1-089 A, 2
- 3.
Lubrite sliding surfaces have been used in ASME Class 1 pipe supports. Lubrite sliding surfaces are not used in the reactor vessel supports.
Virgil C. Summer Nuclear Station Page 3-690 Supplement 4 Application for Subsequent License Renewal
Page12of17
- 4.
Reduction in fracture toughness and loss of intended function due to irradiation is limited to the reactor vessel supports. The aging effects also include deformation, cracking, and misalignment.
- 5.
Degradation identified during the IWF inspections will also be evaluated under the Structures Monitoring program for potential degradation of the grout.
concrete, and embedded steel in the primary shield wall.
Virgil C. Summer Nuclear Station Page 3-691 Supplement 4 Application for Subsequent License Renewal
Page 13 of 17 Supplement 4 Virgil C. Summer Nuclear Station Application for Subsequent License Renewal Appendix A - FSAR Supplement Inspection results will be compared with prior recorded results in acceptance of components for continued service.
In conformance with 10 CFR 50.55a(g)(4)(ii), the containment inservice inspection program will be updated during each successive 120-month inspection interval to comply with the requirements of the latest edition and addenda of the ASME Code specified 12 months before the start of the inspection interval.
A1.31 ASME Section XI, Subsection IWL The ASME Section XI, Subsection IWL program is an existing condition monitoring program that manages the following aging effects for Containment concrete and the unbonded post-tensioning system:
Cracking Cracking; loss of bond; and loss of material (spalling, scaling)
Cracking; loss of material Increase in porosity and permeability; cracking; loss of material (spalling, scaling)
Increase in porosity and permeability; loss of strength Loss of material (spalling, scaling) and cracking Loss of material Loss of prestress This program also includes inspection of tendon and anchorage hardware surfaces, inspection and testing of tendon corrosion protection media, and measurement of tendon force and elongation. This program consists of periodic visual inspection of concrete surfaces for reinforced concrete containments for signs of degradation, assessment of damage, and corrective actions. The Subsection IWL requirements are supplemented to include quantitative acceptance criteria for concrete surfaces based on the "Evaluation Criteria" provided in Chapter 5 of ACI 349.3R.
A1.32 ASME Section XI, Subsection IWF The ASME Section XI, Subsection IWF program is an existing condition monitoring program that manages cracking, loss of material, loss of mechanical function, reduction in fracture toughness, loss of intended function, and loss of preload for ASME Class 1, 2, 3, and MC piping and component supports. This program consists of periodic visual examination of piping and component supports for signs of degradation, evaluation, and corrective actions. This program recommends additional inspections beyond the inspections required by the 10 CFR Part 50.55a ASME Section XI, Subsection IWF program. This includes a one-time inspection within five years prior to entering the subsequent period of extended operation of an additional 5% of the sample Page A-22
Page 14 of 17 Table A4.0-1 Subsequent License Renewal Commitments 32 Program ASME Section XI, Subsection IWF program 10 CFR50, 33 Appendix J program Commitment The ASME Section XI, Subsection IWF program is an existing condition monitoring program that will be enhanced as follows:
- 1. Procedure(s) will be revised to include class MC component supports in the scope of the program.
- 2. Procedure(s) will be revised to evaluate the acceptability of inaccessible areas (e.g., portions of supports encased in concrete, buried underground, or encapsulated by guard pipe) when conditions exist in accessible areas that could indicate the presence of, or result in, degradation to such inaccessible areas.
- 3. Procedure(s) will be revised to require ASTM A325 and ASTM A490 bolts and associated nuts and washers to be stored in closed containers to protect them from dirt and corrosion. Additionally, the closed containers will be required to be stored in a protected shelter (Storage Level B or C) until use.
- 4. Procedure(s) will be revised to specify a one-time inspection within five years prior to entering the subsequent period of extended operation of an additional 5% of the sample populations for Class 1, 2, and 3 piping supports. The additional supports will be selected from the remaining population of IWF piping supports and will include components that are most susceptible to age-related degradation.
- 5. Procedures will be revised to require that at least one reactor pressure vessel support will be inspected every 5 years during the subsequent period of extended operation. (Added - Supplement 4) a,. 2"..Procedure(s) will be revised to require that if a component support does not exceed the acceptance standards of IWF-3400 but is repaired to as-new condition, the sample will be increased or modified to include another support that is representative of the remaining population of supports that were not repaired. {Renumbered - Supplement 4) ee L.._Procedure(s) will be revised to include the additional unacceptable conditions indicated below that are not specified in IWF-3410(a) and to specify any unacceptable conditions may be accepted with a documented technical basis.
- a. Loss of material due to corrosion or wear.
- b. Debris, dirt, or excessive wear that could prevent or restrict sliding of the sliding surfaces as intended in the design basis of the support.
- c. Cracked or sheared bolts, including high-strength bolts, and anchors.
- d. Cracks.
The above conditions may be accepted provided the technical basis for their acceptance is documented. {Renumbered - Supplement 4)
The 10 CFR 50, Appendix J program is an existing performance monitoring program that is credited.
Virgil C. Summer Nuclear Station Application for Subsequent License Renewal Appendix A-UFSAR Supplement PageA-68 AMP B2.1.32 Implementation Program enhancements for SLR will be implemented 6 months prior to the subsequent period of extended operation. The one-time inspections are to begin no earlier than 5 years prior to the subsequent period of extended operation and are to be completed 6 months prior to the subsequent period of extended operation or no later than the last refueling outage prior to the subsequent period of extended operation.
B2.1.33 Ongoing Supplement 4
Supplement 4 B2.1.32 Page 15 of 17 ASME Section XI, Subsection IWF Program Description Virgil C. Summer Nuclear Station Application for Subsequent License Renewal Appendix B - Aging Management Programs The ASME Section XI, Subsection IWF program is an existing condition monitoring program that manages cracking, loss of material, loss of mechanical function, reduction in fracture toughness, loss of intended function, and loss of preload for ASME Classes 1, 2, 3, and MC piping and components supports.
During the fourth inservice inspection interval (January 1, 2014 through December 31, 2023),
inspections of supports for Class 1, 2, 3, and MC piping and components are performed consistent with ASME Code,Section XI, 2007 Edition through 2008 Addenda, as incorporated in 10 CFR 50.55a. In conformance with 10 CFR 50.55a(g)(4)(ii), the lnservice Inspection program is updated during each successive 120-month inspection interval to comply with the requirements of the edition and addenda of the Code that is applicable twelve months before the start of the inspection interval.
ASME Code editions and addenda will be used consistent with the provisions of 10 CFR 50.55a during the subsequent period of extended operation.
Supports for Class 1, 2, 3, and MC piping and components are selected for examination per the requirements of ASME Code,Section XI, Subsection IWF. Acceptance standards are specified in ASME Code,Section XI, Subsection IWF, Subarticle IWF-3400. The scope of the inspection for supports is based on class and total population as defined in Table IWF-2500-1. When a component support requires corrective measures in accordance with the provisions of Article IWF-3000, that support is re-examined during the next inspection period. When the reexaminations do not require additional corrective measures during the next inspection period, the inspection schedule reverts to the requirements of the original inspection program.
Component support examinations that detect flaws or relevant conditions exceeding the acceptance standards of ASME Code,Section XI, Subsection IWF, Subarticle IWF-3400 are extended to include additional examinations in accordance with ASME Code,Section XI, Subsection IWF, Subarticle IWF-2430. The ASME Section XI, Subsection IWF program provides a systematic method for periodic examination of supports for Class 1, 2, 3, and MC piping and component supports. The primary inspection method is visual examination. Procedures include instructions and acceptance standards for the visual examinations (VT-3).
If a component support does not exceed the acceptance standards of ASME Code,Section XI, Subsection IWF, Subarticle IWF-3400, but is electively repaired to as-new condition, then the sample will be increased or modified to include another support that is representative of the remaining population of supports that were not repaired.
Procedures include preventive actions to ensure bolting integrity for replacement and maintenance activities by specifying proper selection of bolting material and lubricants, and appropriate installation torque or tension to prevent or minimize loss of bolting preload and cracking of PageB-166
Supplement 4 Enhancements Page 16of 17 Virgil C. Summer Nuclear Station Application for Subsequent License Renewal Appendix B - Aging Management Programs Prior to the subsequent period of extended operation, the following enhancements will be implemented in the following program element(s):
Scope of Program (Element 1)
- 1.
Procedure(s) will be revised to include class MC component supports in the scope of the program.
- 2.
Procedure(s) will be revised to evaluate the acceptability of inaccessible areas (e.g., portions of supports encased in concrete, buried underground, or encapsulated by guard pipe) when conditions exist in accessible areas that could indicate the presence of, or result in, degradation to such inaccessible areas.
Preventive Actions (Element 2)
- 3.
Procedure(s) will be revised to require ASTM A325 and ASTM A490 bolts and associated nuts and washers to be stored in closed containers to protect them from dirt and corrosion.
Additionally, the closed containers will be required to be stored in a protected shelter (Storage Level B or C) until use.
Detection of Aging Effects (Element 4)
- 4.
Procedure(s) will be revised to specify a one-time inspection within five years prior to entering the subsequent period of extended operation of an additional 5% of the sample populations for Class 1, 2, and 3 piping supports. The additional supports will be selected from the remaining population of IWF piping supports and will include components that are most susceptible to age-related degradation.
- 5.
Procedures will be revised to require that at least one reactor pressure vessel support will be inspected every 5 years during the subseguen t period of extended operation.
(Added - Supplement 4)
Monitoring and Trending (Element 5) e:-
§.,_Procedure(s) will be revised to require that if a component support does not exceed the acceptance standards of IWF-3400 but is repaired to as-new condition, the sample will be increased or modified to include another support that is representative of the remaining population of supports that were not repaired. (Renumbered - Supplement 4)
Acceptance Criteria (Element 6)
L_Procedure(s) will be revised to include the additional unacceptable conditions indicated below.
- a.
Loss of material due to corrosion or wear.
PageB-169
Supplement 4 Page 17 of 17 Virgil C. Summer Nuclear Station Application for Subsequent License Renewal Appendix B - Aging Management Programs
- b.
Debris, dirt, or excessive wear that could prevent or restrict sliding of the sliding surfaces as intended in the design basis of the support.
- c.
Cracked or sheared bolts, including high-strength bolts, and anchors.
- d.
Cracks.
The above conditions may be accepted provided the technical basis for their acceptance is documented. (Renumbered - Supplement 4)
Operating Experience Summary The following examples of operating experience provide objective evidence that the ASME Section XI, Subsection IWF program has been, and will be effective in managing the aging effects for SSCs within the scope of the program so that the intended functions will be maintained consistent with the current licensing basis during the subsequent period of extended operation.
- 1.
In 2012, during an outage closeout walkdown, a nut and bolt on the control rod drive mechanism (CROM) cable bridge connection to the support beam were discovered missing. A bolt attaching the missile shield wall bracket to the top of the missile shield was protruding and did not appear fully tightened, and a gap was identified between the missile shield wall bracket and the wall.
The nut and bolt missing from the CROM cable bridge connection to the support beam were re-installed. The bolt attached to the missile shield wall bracket was reworked. The gap between the missile shield wall bracket and wall were repaired the following outage.
- 2.
In 2013, it was identified that the reactor vessel supports were not included in the ASME Code,Section XI, lnservice Inspection Program. The reactor vessel steel support structures were previously considered to be inaccessible since they are mostly embedded or encased in concrete; and there is no access for a direct general visual inspection. Additionally, technology at the time of construction and early operation was considered inadequate for an acceptable remote visual inspection of these supports. In 2015, a remote visual examination of the accessible portion of the steel supports was performed with satisfactory results. The examinations were performed through air vent channels located on the left and right side of each support and were limited to the bottom section of the support assembly due to accessibility. These supports were also added to the list of ASME Code,Section XI, lnservice Inspection Program inspections.
- 3.
In 2018, a visual examination of a residual heat removal piping spring hanger support found two jam nuts loose. Engineering determined the piping spring hanger support was fully able to perform its design function, and the loose jam nuts were tightened. The piping spring hanger support satisfied the other acceptance standards of the ASME Code,Section XI, Subsection IWF.
PageB-170 Serial No.: 24-309 Docket No.: 50-395 REQUESTED INFORMATION IN RESPONSE TO LIMITED AGING MANAGEMENT AUDIT Dominion Energy South Carolina, Inc.
(Dominion Energy South Carolina, or DESC)
Virgil C. Summer Nuclear Station Unit 1
Serial No.: 24-309 Docket No.: 50-395, Page 2 of 4 During the Limited Aging Management Audit, the NRG staff indicated that additional information related to three topics was needed to support writing the Safety Evaluation Report for Virgil C. Summer Nuclear Station (VCSNS). Dominion Energy South Carolina (DESC) agreed to provide the required information as part of this supplement letter. The three topics and information requested are provided in this enclosure.
- 1. Primary shield wall (PSW) and reactor pressure vessel (RPV) support structure exposure uncertainty estimates Estimates of the uncertainties associated with the PSW and RPV support structure exposures were established using the RPV extended beltline uncertainty analysis described in WCAP-18124-NP-A, Revision 0, Supplement 1-NP-A. Note that the level of detail in the model used for the extended beltline uncertainty analysis described in WCAP-18124-NP-A, Revision 0, Supplement 1-NP-A is commensurate with the plant-specific models for VCSNS. For example, the mesh sizes, treatment of anisotropic scattering, angular quadrature, modeling of internals structures, etc., are similar.
The existing RPV extended beltline analysis quantified the analytical uncertainty associated with calculated fast neutron (E > 1.0 MeV) fluence rates at the RPV inner and outer surfaces at various elevations above and below the active fuel. As part of this analysis, numerous parameters that were identified as having a potentially significant contribution to the core neutron source, reactor geometry, coolant temperature, discretization, and modeling approximation uncertainties at the RPV inner and outer surfaces were evaluated. More specifically, each parameter identified was evaluated on an individual basis by determining the maximum relative change in the base-case fluence rate that occurred as the magnitude of that parameter was varied over a bounding range of values. The net analytical uncertainty associated with a given RPV location was then determined by taking the root sum of squares of the individual parameter uncertainty values determined at that location. Given the parameters considered, the magnitudes of the parameter variations evaluated, and the relative proximity of the RPV outer surface to the PSW and RPV support structure, the extended beltline uncertainty analysis results for the RPV outer surface were judged to provide a reasonable basis for estimating the analytical uncertainty associated with the PSW and RPV support structure exposures.
The maximum neutron fluence and gamma dose projections at the inner surface of the PSW occur at elevations that are near the core midplane. However, since the extended beltline uncertainty analysis was, by design, focused on the RPV extended beltline region only, it did not consider axial elevations near the core midplane; the elevations nearest the mid plane considered were 30 cm above the top and 30 cm below the bottom of the active fuel. Therefore, the extended beltline uncertainty analysis results determined at the RPV outer surface, 30 cm above the top of the active fuel were used as the starting point for estimating the uncertainty associated with the PSW neutron and gamma exposures. This is conservative because analytical uncertainties increase with axial distance above the top of the active fuel.
Serial No.: 24-309 Docket No.: 50-395, Page 3 of 4 In addition to using this bounding RPV location as a starting point, the concrete composition parameter uncertainty value determined at this location was increased by a factor of 2. This value was increased because it was associated with the one parameter evaluated in the RPV extended beltline analysis whose uncertainty was judged to be potentially impacted in a non-negligible manner if a detailed uncertainty analysis for the PSW were performed. Note that the standard concrete composition from the BUGLE-96 documentation (BUGLE-96, Coupled 47 Neutron, 20 Gamma-Ray Group Cross Section Library Derived from ENDF/B-VI for LWR Shielding and Pressure Vessel Dosimetry Applications) was used for the PSW in both the extended beltline analytical uncertainty analysis base-case calculations and the VCSNS radiation transport calculations.
Following this process, the analytical uncertainty associated with the fast neutron (E > 1.0 MeV) fluence and gamma dose results at the inner surface of the PSW was conservatively estimated to be 20%. It is worth noting that:
the estimated 20% value is based on the extended beltline uncertainty analysis results determined at the RPV outer surface, 30 cm above the top of the active fuel, and analytical uncertainties at the RPV outer surface increase with distance from the core midplane elevation.
Therefore, the estimated 20% uncertainty is:
representative for fast neutron (E > 1.0 MeV) fluence and gamma dose results determined at the PSW inner surface and axial elevations within a foot of the top and bottom of the active fuel, and bounding for fast neutron (E > 1.0 MeV) fluence and gamma dose results determined at the PSW inner surface and axial elevations near the core midplane.
The estimated 20% uncertainty does not explicitly account for neutrons with energies between 1.0 MeV and 0.1 MeV. However, the maximum PSW exposures determined for VCSNS occur at elevations near the core midplane, where the analytical uncertainty for fast neutron (E > 1.0 MeV) fluence at the PSW inner surface is significantly less than 20%. For example, Section 4.5 of WCAP-18124-NP-A-documents that the analytical uncertainty for fast neutron (E > 1.0 MeV) fluence in the reactor cavity (i.e., at the RPV outer surface) at the core midplane elevation is approximately 12%. While the uncertainty associated with fast neutron (E > 0.1 MeV) fluence at the PSW inner surface and elevations near the core midplane is greater than 12%, it would not be expected to be significantly different, or greater, than the estimated uncertainty of 20%
assigned to the PSW maximum exposures.
Westinghouse has not performed an analytical uncertainty analysis associated with neutron fluence exposures at the inner surface of the reactor cavity liner plate or at various depths within the PSW concrete. Since the maximum neutron exposures of
Serial No.: 24-309 Docket No.: 50-395, Page 4 of 4 the reactor cavity liner plate and RPV support structure anchorage embedded in the PSW concrete occur at elevations near the core midplane, the estimated 20%
uncertainty was applied to these exposures as well.
Finally, it is recognized that the analytical uncertainty at various depths within the PSW concrete may be greater than the uncertainty at the inside surface of the PSW concrete.
However, even if the 20% uncertainty applied to the maximum fast neutron (E > 1.0 MeV) fluence exposures of the RPV support structure anchorage were increased by, for example, a factor of five, the resulting exposures would still be significantly (i.e., more than 5x) less than the damage threshold of 1.00 E+19 n/cm2.
The maximum neutron fluence and iron atom displacement projections for the RPV support structure (i.e., the support box plate, support box, support shoe, and support box plate bolt) occur at elevations that are less than 2 ft above the top of the active fuel. Therefore, the extended beltline uncertainty analysis results determined at the RPVouter surface 90 cm above the top of the active fuel were used for the RPV support structure. This is conservative because analytical uncertainties increase with axial distance above the top of the active fuel.
Following the same process that was used for the PSW concrete, the analytical uncertainty associated with the neutron fluence and iron atom displacement results for the RPV support structure was estimated to be 25%.
- 2. Reactor cavity air flow velocity As documented in WCAP-18772-P/NP, gamma heating on the PSW has been evaluated. It was concluded that the maximum PSW concrete temperature would be less than 145°F. This temperature is below the long-term temperature limit imposed on concrete structures. In addition, this temperature is conservatively based on a reactor cavity air flow velocity of 5 ft/sec instead of the calculated value of 34.8 ft/sec that is based on the design reactor cavity flow rate of 30,000 CFM. Therefore, gamma heating is not an issue for the PSW concrete.
- 3. ASME Code,Section XI, lower bound Kie fracture toughness value for the plant-specific material of the plate components of the RPV supports The NRG staff requested that DESC confirm and docket the ASME Code,Section XI, lower bound Kie fracture toughness value of 33.2 ksiin is bounding for the plant-specific material of the plate components of the RPV supports. DESC confirms that the ASME Code,Section XI, lower bound Kie fracture toughness value of 33.2 ksiin is bounding for the plant-specific material of the plate components of the RPV supports.