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Category:Report
MONTHYEAR3F0324-03, Enclosures 10 - 17: Crystal River, Unit 3, Response to Request for Additional Information2024-03-31031 March 2024 Enclosures 10 - 17: Crystal River, Unit 3, Response to Request for Additional Information ML24090A0042024-03-29029 March 2024 Enclosure 6, Part 1: Crystal River, Unit 3 - Appendix E – CHAR-05 Site Characterization of Buildings ML24030A7472024-02-13013 February 2024 Audit Report Attachment - Crystal River Unit 3 Nuclear Generating Plant LTP ML24030A7482024-02-12012 February 2024 Audit Report Cover Letter and Report - Crystal River Unit 3 Nuclear Generating Plant LTP ML23345A1882023-12-0606 December 2023 Fws to NRC Crystal River Species List of Threatened and Endangered Species That May Occur in Your Proposed Project Location or May Be Affected by Your Proposed Project ML24090A0052023-10-10010 October 2023 Enclosure 6, Part 2: Crystal River, Unit 3, DD Survey 23-10-0096 ML23160A2962023-06-0909 June 2023 Response to Crystal River, Unit 3 Supplemental Information Needed for Acceptance on the Application for a License Amendment Regarding Approval of the License Termination Plan ML23160A2972023-06-0909 June 2023 CR3 Site Characterization Survey Report (CHAR-01) Impacted Open Land Survey Areas 3F0623-02, Maintenance Support Building2023-06-0909 June 2023 Maintenance Support Building ML23160A2982023-06-0909 June 2023 Site Characterization Surveys ML23107A2732023-06-0707 June 2023 Orise Independent Survey Report Dcn 5366-SR-01-0 3F0522-01, Safety Analysis Report, Quality Assurance Program and 10 CFR 50.59 - 10 CFR 72.48 Report - May 20222022-05-17017 May 2022 Safety Analysis Report, Quality Assurance Program and 10 CFR 50.59 - 10 CFR 72.48 Report - May 2022 3F0520-01, Safety Analysis Report, Quality Assurance Program and 10 CFR 50.59 - 10 CFR 72.48 Report - May 20202020-05-18018 May 2020 Safety Analysis Report, Quality Assurance Program and 10 CFR 50.59 - 10 CFR 72.48 Report - May 2020 3F0320-01, NRC Commitment Change Report - March 20202020-03-17017 March 2020 NRC Commitment Change Report - March 2020 ML19343A8252019-12-0606 December 2019 Letter from Erika Bailey, Oak Ridge Institute for Science and Education, to John Hickman, NRC, Forwarding Independent Confirmatory Survey Summary and Results for the 3,854-Acre Area Partial Site Release at the Crystal River Energy Complex ML19022A0762019-01-22022 January 2019 Partial Site Release Request ML19029A0092018-11-0707 November 2018 Reference 16 - Defueled Safety Analysis Report DSAR-R002 ML18303A2942018-06-21021 June 2018 Golder Associates, Inc. - Citrus Combined Cycle Project - CFR 122.21(r) Report 3F0518-03, Safety Analysis Report and 10 CFR 50.59 - 10 CFR 72.48 Report - May 20182018-05-24024 May 2018 Safety Analysis Report and 10 CFR 50.59 - 10 CFR 72.48 Report - May 2018 ML16176A3392016-10-28028 October 2016 Decommissioning Lessons Learned Report and Transmittal Memorandum ML19029A0102016-06-28028 June 2016 Reference 3 - Crystal River, Unit 3, Historical Site Assessment Rev. 00 3F0616-02, Nrg Commitment Change Report - June 20162016-06-14014 June 2016 Nrg Commitment Change Report - June 2016 ML13343A1782013-12-31031 December 2013 Report P23-1680-001, Rev. 0, Site-Specific Decommissioning Cost Estimate for Crystal River Unit 3 Nuclear Generating Plant. 3F0113-08, Attachment D: ANP-3195(NP), Revision 0, Response for Crystal River Unit 3, EPU Licensing Amendment Report NRC Reactor Systems Branch Requests for Additional Information (Non-Proprietary) and Attachment E: Location of Reactor Systems RAI Re2013-01-31031 January 2013 Attachment D: ANP-3195(NP), Revision 0, Response for Crystal River Unit 3, EPU Licensing Amendment Report NRC Reactor Systems Branch Requests for Additional Information (Non-Proprietary) and Attachment E: Location of Reactor Systems RAI Res 3F1112-01, Alion Technical Report ALION-PLN-ENER-8706-02, Rev. 0, Crystal River 3: Bypass Fiber Quantity Test Plan2012-11-0707 November 2012 Alion Technical Report ALION-PLN-ENER-8706-02, Rev. 0, Crystal River 3: Bypass Fiber Quantity Test Plan 3F1112-02, 17877-0001-100, Rev. 1, CR-3 Inadequate Core Cooling Mitigation System Reliability Assessment, Task 1, Page 174 of 250 Through Page 250 of 2502012-10-31031 October 2012 17877-0001-100, Rev. 1, CR-3 Inadequate Core Cooling Mitigation System Reliability Assessment, Task 1, Page 174 of 250 Through Page 250 of 250 ML12314A3932012-10-31031 October 2012 17877-0001-100, Rev. 1, CR-3 Inadequate Core Cooling Mitigation System Reliability Assessment, Task 1, Page 81 of 250 Through Page 173 of 250 ML12314A3922012-10-31031 October 2012 17877-0001-100, Rev. 1, CR-3 Inadequate Core Cooling Mitigation System Reliability Assessment, Task 1, Page 1 of 250 Through Page 80 of 250 3F0912-01, ANP-3156 Np, Crystal River 3 EPU Boric Acid Precipitation RAI Responses, Attachment C2012-09-30030 September 2012 ANP-3156 Np, Crystal River 3 EPU Boric Acid Precipitation RAI Responses, Attachment C 3F0712-03, Attachment E to 3F0712-03, Technical Report, ANP-3052, Rev. 2, CR-3 EPU Feedwater Line Break Analysis with Failure of First Safety Grade Trip.2012-06-30030 June 2012 Attachment E to 3F0712-03, Technical Report, ANP-3052, Rev. 2, CR-3 EPU Feedwater Line Break Analysis with Failure of First Safety Grade Trip. ML12314A3912012-05-31031 May 2012 17877-0002-100, CR-3 Inadequate Core Cooling Mitigation System Reliability Assessment 2 ML12205A3582012-05-31031 May 2012 Attachment D to 3F0712-03, Technical Report, ANP-3114(NP), Rev. 0, CR-3 EPU - Feedwater Line Break Analysis Sensitivity Studies. ML12284A1382012-05-25025 May 2012 Report EGS-TR-HC589-01, Seismic Qualification Test Report for Structural Verification Testing of Iccms Cabinet Assembly. 3F0512-01, NRC Commitment Change Report - May 20122012-05-14014 May 2012 NRC Commitment Change Report - May 2012 3F0112-04, Response to Request for Additional Information to Support NRC Steam Generator Tube Integrity and Chemical Engineering Branch Technical Review of the CR-3 Extended Power Uprate LAR2012-01-0505 January 2012 Response to Request for Additional Information to Support NRC Steam Generator Tube Integrity and Chemical Engineering Branch Technical Review of the CR-3 Extended Power Uprate LAR ML12205A3572011-12-15015 December 2011 Attachment a to 3F0712-03, CR-3 LOCA Summary Report - EPU/ROTSG/Mark-B-HTP, Revision 4 3F1211-14, Summary of Changes to Evaluation Models and Peak Cladding Temperature for Large Break Loss of Coolant Analysis and Small Break Loss of Coolant Analysis2011-12-14014 December 2011 Summary of Changes to Evaluation Models and Peak Cladding Temperature for Large Break Loss of Coolant Analysis and Small Break Loss of Coolant Analysis 3F1011-08, ANP-3052, Rev. 0, CR-3 EPU Feedwater Line Break Analysis with Failure of First Safety Grade Trip2011-10-25025 October 2011 ANP-3052, Rev. 0, CR-3 EPU Feedwater Line Break Analysis with Failure of First Safety Grade Trip ML11237A0682011-08-0505 August 2011 Siemens Technical Report CT-27438, Missile Probability Analysis Report Progress Energy Crystal River 3, Revision 1A ML11207A4442011-06-15015 June 2011 Attachment 7- Crystal River Unit 3 Extended Power Uprate Technical Report 3F0511-02, Response to Request for Additional Information Required for the Development of the Confirmatory LOCA and Non-LOCA Models2011-05-0606 May 2011 Response to Request for Additional Information Required for the Development of the Confirmatory LOCA and Non-LOCA Models ML1101906672010-10-0404 October 2010 Levy, Units 1 and 2, Cola (Sensitive Material), Rev. 2 - Levy County Emergency Plan Part 02 - Draft (Redacted) 3F0910-01, CFR 50.46 Notification of Change in Peak Cladding Temperature for Small Break Loss of Coolant Accident Analysis2010-09-0808 September 2010 CFR 50.46 Notification of Change in Peak Cladding Temperature for Small Break Loss of Coolant Accident Analysis ML1019304172010-05-0606 May 2010 Tritium Database Report ML1010603472010-04-0909 April 2010 5.2.2.4.4. Quality Control and Nondestructive Testing ML1028710882010-03-12012 March 2010 7.6 Vibration Due to Cutting Tendons ML1028711112010-02-25025 February 2010 7.11 Added Stress from Pulling Tendons ML1028711102010-02-23023 February 2010 6.3 Thermal Effects of Greasing ML1028804682010-02-19019 February 2010 7.9 Inadequate Hydro Blasting Nozzles Rate Part 3 ML1028711212010-02-19019 February 2010 7.10 Hydrodemolition Induced Cracking 2024-03-31
[Table view] Category:Miscellaneous
MONTHYEARML16176A3392016-10-28028 October 2016 Decommissioning Lessons Learned Report and Transmittal Memorandum 3F0616-02, Nrg Commitment Change Report - June 20162016-06-14014 June 2016 Nrg Commitment Change Report - June 2016 3F0113-08, Attachment D: ANP-3195(NP), Revision 0, Response for Crystal River Unit 3, EPU Licensing Amendment Report NRC Reactor Systems Branch Requests for Additional Information (Non-Proprietary) and Attachment E: Location of Reactor Systems RAI Re2013-01-31031 January 2013 Attachment D: ANP-3195(NP), Revision 0, Response for Crystal River Unit 3, EPU Licensing Amendment Report NRC Reactor Systems Branch Requests for Additional Information (Non-Proprietary) and Attachment E: Location of Reactor Systems RAI Res 3F0512-01, NRC Commitment Change Report - May 20122012-05-14014 May 2012 NRC Commitment Change Report - May 2012 3F0112-04, Response to Request for Additional Information to Support NRC Steam Generator Tube Integrity and Chemical Engineering Branch Technical Review of the CR-3 Extended Power Uprate LAR2012-01-0505 January 2012 Response to Request for Additional Information to Support NRC Steam Generator Tube Integrity and Chemical Engineering Branch Technical Review of the CR-3 Extended Power Uprate LAR 3F1211-14, Summary of Changes to Evaluation Models and Peak Cladding Temperature for Large Break Loss of Coolant Analysis and Small Break Loss of Coolant Analysis2011-12-14014 December 2011 Summary of Changes to Evaluation Models and Peak Cladding Temperature for Large Break Loss of Coolant Analysis and Small Break Loss of Coolant Analysis 3F0511-02, Response to Request for Additional Information Required for the Development of the Confirmatory LOCA and Non-LOCA Models2011-05-0606 May 2011 Response to Request for Additional Information Required for the Development of the Confirmatory LOCA and Non-LOCA Models 3F0910-01, CFR 50.46 Notification of Change in Peak Cladding Temperature for Small Break Loss of Coolant Accident Analysis2010-09-0808 September 2010 CFR 50.46 Notification of Change in Peak Cladding Temperature for Small Break Loss of Coolant Accident Analysis ML1019304172010-05-0606 May 2010 Tritium Database Report ML1028710882010-03-12012 March 2010 7.6 Vibration Due to Cutting Tendons ML1028711112010-02-25025 February 2010 7.11 Added Stress from Pulling Tendons ML1028711102010-02-23023 February 2010 6.3 Thermal Effects of Greasing ML1028702742010-02-0606 February 2010 6.4 Rigid and Flexible Sleeves ML1028804952010-01-20020 January 2010 5.3 Exhibit 1 Interviews Regarding Maintenance ML1028708922010-01-16016 January 2010 Email - from: Miller, Craig L (Craig.Miller@Pgnmail.Com) to: Lake, Louis; Thomas, George; Carrion, Robert; 'Trowe@Wje.Com' Cc: Williams, Charles R. Dated Saturday, January 16, 2010 1:10 PM Subject: Failure Mode 5.1 for Review and Comment ML1028709832010-01-12012 January 2010 from: Thomas, George to: Naus, Dan J. CC: Lake, Louis Dated Tuesday, January 12, 2010 3:57 PM Subject: FW: Petrographic Report ML1028709862010-01-12012 January 2010 Email - from Thomas, George to Naus, Dan J. Cc: Lake, Louis Dated Tuesday, January 12, 2010 3:56 PM Subject: FW: Petrographic Report Attachments: S&Me Modulus Core 16, 40, 63, 65, 66.pdf; S&Me Density Core 16, 40, 60, 63, 65, 66.pdf... ML1028709792010-01-12012 January 2010 Email - from: Thomas, George to: Naus, Dan J. CC: Lake, Louis Dated Tuesday, January 12, 2010 3:59 PM Subject: FW: Petrographic Report ML1028710352010-01-11011 January 2010 Email - from: Williams, Charles R. to: Lake, Louis; Thomas, George; Carrion, Robert; 'Nausdj@Ornl.Gov'; 'Daniel.Fiorello@Exeloncorp.Com'; Lese, Joseph A. Cc: Miller, Craig L (Charles.Williams@Pgnmail.Com) Dated Monday, January 11, 2010 5:38 ML1028710342010-01-11011 January 2010 Email - from: Williams, Charles R. (Charles.Williams@Pgnmail.Com) to Lake, Louis; Thomas, George; 'Nausdj@Ornl.Gov'; Carrion, Robert; Souther, Martin; 'Trowe@Wje.Com' Dated Monday, January 11, 2010 5:42 PM Subject: Fm 5.4 Draft for Review ML1028709902010-01-11011 January 2010 Email - from: Williams, Charles R. (Charles.Williams@Pgnmail.Com) to Lake, Louis; Thomas, George; 'Nausdj@Ornl.Gov'; Carrion, Robert; Souther, Martin; 'Archer, John C. (Reading)'; 'Wells, Richard P. (Reading)' Cc: Miller, Craig L Subject: F ML1028710402010-01-0606 January 2010 Email - from: Williams, Charles R. (Charles.Williams@Pgnmail.Com) to: Lake, Louis; Thomas, George; 'Nausdj@Ornl.Gov'; Carrion, Robert; Souther, Martin; 'Trowe@Wje.Com' Cc: Miller, Craig L Dated Wednesday, January 06, 2010 4:51 PM, Subject: ML1028710462010-01-0101 January 2010 5.5 Exhibit 3a Petrographic Erlin Hime May 19 ML1028710492009-12-31031 December 2009 Email - from: Williams, Charles R. Charles.Williams@Pgnmail.Com) to: Lake, Louis; Thomas, George; 'Nausdj@Ornl.Gov'; Carrion, Robert; Miller, Craig L Dated Thursday, December 31, 2009 5:20 PM Subject: Fm 9.3 Draft and Exhibits for Review ML1029106012009-12-23023 December 2009 5.8 Exhibit 5a Petrographic ML1029106322009-12-18018 December 2009 from: Miller, Craig L (Craig.Miller@Pgnmail.Com) to: Lake, Louis;Thomas, George; Nausdj@Ornl.Gov; Carrion, Robert Cc: Williams, Charles R. Dated Friday, December 18, 2009 4:47 PM Subject: Draft Fm 3.5 for Review Attachments: Fm 3.5.pptx; Fm 3F1209-11, 10 CFR 50.46 Loss-of-Coolant Accident Evaluation Model Change and Peak Cladding Temperature Change Report2009-12-16016 December 2009 10 CFR 50.46 Loss-of-Coolant Accident Evaluation Model Change and Peak Cladding Temperature Change Report ML1029107052009-12-11011 December 2009 5.2 Exhibit 2 Mactec Petrographic ML1029201382009-12-0808 December 2009 5.1 Exhibit 6A Petrographic Mactec ML1029201442009-12-0808 December 2009 5.8 Exhibit 5b Petrographic ML1029107272009-10-27027 October 2009 Summary of Results from Comparison.Pdf ML1029205002009-10-13013 October 2009 Condition Report Summary -NCRs Initiated Since Oct 12, 2009 ML1007504262009-10-12012 October 2009 Summary Report ML1028804902009-10-0505 October 2009 Condition Report Summary - Ncrs Initiated Since Sep 28, 2009 ML1029201092009-02-0404 February 2009 5.6 Exhibit 8 Cap Grease Leakage ML1029201072009-02-0404 February 2009 5.6 Exhibit 5 Surveillance Conclusion 3F0508-14, NRC Commitment Change Report - May 20082008-05-28028 May 2008 NRC Commitment Change Report - May 2008 3F1207-03, Engineering Report ER-608NP, Revision 2, LEFM + Meter Factor Calculation and Accuracy Assessment for Crystal River Unit 3 Nuclear Power Station.2007-12-13013 December 2007 Engineering Report ER-608NP, Revision 2, LEFM + Meter Factor Calculation and Accuracy Assessment for Crystal River Unit 3 Nuclear Power Station. ML0731103812007-05-21021 May 2007 Progress Energy Florida, Inc. Proposed Florida Nuclear Site Transmission Planning Study, Final Report, Attachments Dand F Included 3F1205-04, 10 CFR 50.46 Loss-Of-Coolant Accident Evaluation Model Change and Peak Cladding Temperature Change Report2005-12-12012 December 2005 10 CFR 50.46 Loss-Of-Coolant Accident Evaluation Model Change and Peak Cladding Temperature Change Report 3F1205-03, Special Report 05-01: Once-Through Steam Generator (OTSG) Notifications Required Prior to Mode 42005-12-0303 December 2005 Special Report 05-01: Once-Through Steam Generator (OTSG) Notifications Required Prior to Mode 4 ML0613806752005-12-0101 December 2005 Report of Independent Auditors ML0613806782005-09-30030 September 2005 Financial Highlights ML0524401902005-08-10010 August 2005 Attachment E, Crystal River Unit 3 - License Amendment Request #290, Revision 1 Probabilistic Methodology to Determine the Contribution to Main Steam Line Break Leakage Rates for the Once-Through Steam Generator from the Tube End Crack ML0505301682005-02-17017 February 2005 Background Discussion Material for February 24-25, 2005 NRC Meeting Relating to Replacement for BAW-2374, Revision 1, Evaluation of OTSG Thermal Loads During Hot Leg Loca. ML0318909362003-07-10010 July 2003 Relaxation of the Order, Exercising Enforcement Discretion, and Extension of the Time to Submit an Answer or Request a Hearing Regarding Order EA-03-038, Fitness-for-Duty Enhancements for Nuclear Security Force Personnel for Brunswick, Crys 3F1102-08, Sea Turtle Mortality Report Submitted to the U. S. National Marine Fisheries Service2002-11-19019 November 2002 Sea Turtle Mortality Report Submitted to the U. S. National Marine Fisheries Service ML0300701722002-10-25025 October 2002 Summary Report - Ncrs Initiated Since 10/18/2002 3F0902-09, Submittal of Core Operating Limits Report, Cycle 13, Revision 1, for Crystal River Unit 32002-09-24024 September 2002 Submittal of Core Operating Limits Report, Cycle 13, Revision 1, for Crystal River Unit 3 ML0300701742002-09-23023 September 2002 Summary Report - Ncrs Initiated Since 09/16/2002 2016-06-14
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FM 9.3 Exhibit 1 Page 3 of 11 30.The cavities are usually vertically oriented and generally contain a secondary infill of sand and silty sand.Considering the extreme variation in engineering properties and the difficulty of obtaining data from low yield zones, the differentially cemented Iimerock was characterized as a weakly cemented sand containing random discontinuities in the form of solution cavities and associated highly altered limerock zones. The load-settlement response taken as.representative of the weaker elements of the formation was assumed to be characterized by a modulus of deformation of 54 ksi as would be derived from the loading of a one foot diameter, rigid bearing plate. The in situ shear strength of the differentially cemented limerock although dependent on confining pressure and varying with the degree of cementation, was very conservatively assumed to have an in situ shear strength of 9 tsf, independent of confining pressure.1.2.11.4 Avon Park Limestone Beneath a thin discontinuous zone usually identified as a depositional discontinuity and termed the "Transition Zone", the Avon Park doloarenite member -although containing randomly distributed solution voids -was characterized as a rigid, relatively incompressible rock with a modulus of deformation of 140 ksi for the upper zone and 530 ksi for the FM 9.3 Exhibit 1 Page 4 of 11 31.underlying rock. Based on uniaxial compression tests on representative core specimens, the average unconfined strength of the formation was. assumed to be 700 tsf. On the basis of the depth, extent and character of the "Transition Zone" materials (usually classified as a stiff to hard dolomitized silt), these materials were not considered to influence the foundation analysis.1.2.11.5 Bearinq Capacity Analysis Analyses of bearing capacity were performed first assuming that foundations would be based on the differentially cemented limerock and that the limerock would react as a weakly cemented cohesive material.
Thus, the bearing capacity expression is given by: quit. = 6 c (I + 0.21/8) (01)where, c is the in situ strength, D is the depth of the mat base below final grade, and B is the diameter of the mat. To assess the deep crushing potential of the limerock, an analysis of the imposed vertical stresses (lA0z ) was made to determine the average unit pressure imposed at various elevations below the mat. By letting A =z = quit. and solving Eq. (01) for c, the influence of solution voids may also be qualitatively considered by assuming the average 6az to be increased on a given horizontal plain in accordance with the following expression:
= -A~z T--n(2)
FM 9.3 Exhibit 1 Page 5 of 11 32.where n is the ratio of the total area of voids to the total stressed area under consideration, assuming an idealized regular distribution of solution voids.For a conventional bearing capacity analysis assuming a mat width of 147 ft. and an average contact pressure of 7.8 ksf, the required shear strength for a safety factor of 3 is only in the order of 3.0 ksf.Extending the analysis to consider the failure potential of any extensive weak zones within the foundation rock, the most critical condition is postulated at elevation
+60 where the shear strength requirement for a safety factor of 3 is approximately 2.0 ksf. Should the void area ratio at elevation 60 be as much as 50 percent, the required shear strength for a safety factor of 3 would be doubled, indicating a shear strength requirement of 4.0 ksf. Comparison of these values with an average in situ shear strength of 9 tsf (assumed to characterize the differentially cemented limerock) indicates a wide margin of safety against a bearing capacity failure provided that massive unfilled solution voids are not present within a zone extending below the foundation down to about elevation
+30.Analysis of the 1.5 times accident pressure condition where imposed transient pressures at the center of the Reactor Building foundation mat are assumed to be about 35 ksf FM 9.3 Exhibit 1 Page 6 of 11 33.indicates that a bearing capacity failure will not occur although some localized overstressing and additional foundation settlement would be expected.
A similar analysis to the foregoing was conducted assuming a c = o condition and solving for required frictional strength parameters.
This analysis also demonstrated that the Reactor Building foundation mat was not subject to a bearing capacity failure under the most unfavorable condition which could be reasonably postulated.
1.2.11.6 Settlement Analysis Settlement analysis was predicated on removal of the Quaternary deposits and of the immediately underlying loose to medium dense decomposed limerock horizon. Thus, it was assumed that foundation elements would bear directly either on the cap rock or the differentially cemented limerock units of the Inglis Limestone.
It was also assumed that any load-bearing fill materials used beneath the foundation would consist of materials of a quality at least equivalent to the weakly cemented limerock materials.
A pseudo-elastic method of analysis was used by adapting a form of Equation (03) for a multi-layered foundation system as proposed by Vesic.(1)P = pIp2(J-y 2)ao d E (03)(1) Vesic, A.B. (1963) "The Validity of Layered Solid Theories", Proceedings, International Conference, Structural Design of Asphalt Pavements, University of Michigan.
FM 9.3 Exhibit 1 Page 7 of 11 34.where, I 1 , and 1/22 are embedment and shape factors, Y = Poissons Ratio, a0 = average contact pressure, d = diameter of mat and E = Modulus of Deformation.
The angular deformation of the Reactor Building mat under transient wind loading was also estimated in accordance with elastic theory using a pseudo-static method of analysis proposed by Weissman and White(1).It was concluded that the foundation deformation contributed by the Inglis and Avon Park formations would occur as a small, essentially immediate deformation, the major settlement contribution being derived from the differentially cemented limerock member of the Inglis Limestone.
Estimates of total operating load deformation of the Reactor Building foundation system considered the load superposition from the adjacent structures and from the exterior fills. Results of this analysis indicated the Upper limit of total settlement of the mat to be in the order of 7/8 of an inch at the center of the semi-rigid mat foundation.
It was noted that all but a very small fraction of this settlement would be expected to occur during construction
-before installation of equipment or instrumentation which may be sensitive to slight differential movement.Differential between the load center and edge of the mat was estimated to be in the order of 5/16 of an inch in 75 ft.(1) Weissman and White (1961) "Small Angular Deflexions of Rigid Foundations", Geotechnique, Vol. II, No. 3.
FM 9.3 Exhibit 1 Page 8 of 11 35.indicating a maximum angular distortion in the order of 3.5 x 10-4 radians for the most unfavorable supporting conditions.
Consider~ng an estimated additional angular distortion of 0.2 x 10-5, due to wind forces, the total angular distortion from the center to the edge of the Reactor Building mat was not expected to exceed an order of magnitude of three to four times 10-4 radians under the most unfavorable wind and static loading conditions which could be postulated.
It was also concluded that the settlement distortion of foundations supporting other components of the plant complex would be less than estimated for the Reactor Building foundation.
1.2.11.7 Foundation Treatment On the basis of bearing capacity and settlement analyses, it was concluded that the continuity and integrity of the solutioned limestone within a zone directly beneath all foundation units extending at least down to elevation
+30 in the Reactor Building area should be assured by cement grouting, primarily to fill all solution voids of significant extent and secondarily to provide some densification of loose discrete grained infill materials associated with solution voids. With respect to the optimum grout zone depth, consideration was given to extending consolidation grouting to the doloarenite in lieu of employing a quick-set additive or other procedures to minimize grout excape beyond the base of the consolidation zone. This latter FM 9.3 Exhibit 1 Page 9 of 11 36.alternative was adopted and consolidation grouting was accomplished using a procedure which employed a grout curtain to aid in groundwater control and to prevent lateral escape of grout during split-hole consolidation grouting.1.2.11.8 Excavation and Groundwater Control Considering the undesirable characteristics of the surficial materials, it was concluded that excavations should extend down to competent materials below the loose to medium dense decomposed limerock horizon. As it was expected that excavation of unsuitable materials would require excavations extending well below groundwater level, special groundwater control techniques were recommended to minimize detrimental ground loss by piping of foundation materials under excessive hydraulic gradients.
It was therefore concluded that dewatering should be primarily accomplished by pumping from shallow sumps and other subdrainage systems filtered to preclude excessive removal of fines.It was recognized that a piping potential would exist even with the most appropriate dewatering system and that piping may have localized detrimental influence on the stability of foundation materials.
The occurrence of extensive infill deposits not detected by the subsurface exploration and which would be unsuitable for foundation support was also recognized.
FM 9.3 Exhibit 1 Page 10 of 11.37, It was therefore concluded that should check borings, cement grout-take analysis or permeability tests made after grouting indicate an area of cc-nparatively high porosity or low density, chemical grouting would be required if the unsuitable materials were too extensive for removal and replacement and could be spanned by the foundation system.An alternative subaqueous excavation technique was recommended utilizing a confined or unconfined excavation, the latter recomrnended for conditions where the depth of excavation below water level is limited. A confined excavation (sheeted cofferdam) was recommended where the depth of excavation below water level would exceed about 10 ft. over an extensive area. Bottom clean-out procedures were specified including vacuum cleaning (air lifting) of any collected bottom sediments.
1.2.11.9 Load-BearinQ Materials As the depth to suitable bearing materials was expected to vary considerably in some areas, it was anticipated that it would be desirable to utilize load-bearing fills beneath foundation elements.
It was recommended that fill placed below groundwater level consist of a crushed limestone aggregate (Zone I), suitably graded for underwater placement and for in-place grouting.
For above water placement, the use of well graded, crushed limestone aggregates (Zone II, Zone A and Zone B) was recommended.
These FM 9.3 Exhibit 1 Page 11 of 11 38.materials are capable of being compacted to a high relative density and are graded (Zones A and B) to facilitate subdrainage.
Alternatively, a lean concrete fill was recommended.
A third material, friable crushed limestone (Zone Il), was recommended for placement outside of structure areas.The recommended material quality requirements and compaction criteria for the three load-bearing fill types are contained in Specification SP-5629, "Specifications for Excavation and Placement of Structural Fill". These criteria were developed from the results of compaction, uniaxial compression and triaxial compression tests on representative samples prepared in a manner to simulate anticipated field conditions.
The strength and compressibility of both the grouted and compacted materials were found to be acceptable for foundation support.1.2.12 Unit No. 2 Foundation Grouting To prepare for grouting of the foundation of the proposed Unit No. 3 Nuclear facility, Unit No. 2 foundation was used to develop the techniques and materials necessary to provide adequate support for the structures. (See Volume Il, Section 3.3.0 for detailed report.)In order to establish an acceptable grouting process and to document the effectiveness of such a procedure, the following were performed at various stages in the grouting process:
FM 9.3 Exhibit 3 From: Pugh, C-Glenn Sent: Thursday, October 22, 2009 8:41 AM To: Miller, Craig L
Subject:
FW: RB Settlement Craig, To help close the loop on RB settlement I talked to several plant personnel involved in Maintenance, Ops, I&C Engineering, Civil Engineering, etc and no one can ever remember any instruments or programs for monitoring RB settlement.
Our FSAR contains a statement that predicted settlement is essentially ignored and not a concern.Consider this issue closed.Glenn Pugh