Letter Sequence Supplement |
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MONTHYEARML20082D6391983-11-14014 November 1983 Forwards Safety Evaluation for McGuire Units 1 & 2 Transition to Westinghouse 17x17 Optimized Fuel Assemblies. License Amend Requests,Including Tech Spec Changes & cycle- Specific Reload Safety Evaluations,Will Be Submitted Project stage: Other ML20082D6571983-11-14014 November 1983 Safety Evaluation for McGuire Units 1 & 2 Transition to Westinghouse 17x17 Optimized Fuel Assemblies Project stage: Other ML20099G7481984-11-16016 November 1984 Application for Amends to Licenses NPF-9 & NPF-17,revising Tech Specs Re Plant Operating Limitations Affected by Use of Optimized Fuel Assembly Designs.Fee paid.Marked-up Tech Specs Encl Project stage: Request ML20101E3661984-12-14014 December 1984 Forwards Peaking Factor Limit Rept for Cycle 2 to Support 841116 Proposed Amends to Licenses NPF-9 & NPF-17 Re Plant Operating Limitations Affected by Use of Optimized Fuel Assembly Design Project stage: Other NRC-85-2995, Requests Review of WCAP-9561,Addendum 1, Addendum to Bart A1:Computer Code for Best Estimate Analysis of Reflood Transients, in Support of Anticipated Startup in Late Mar 19851985-01-0909 January 1985 Requests Review of WCAP-9561,Addendum 1, Addendum to Bart A1:Computer Code for Best Estimate Analysis of Reflood Transients, in Support of Anticipated Startup in Late Mar 1985 Project stage: Other ML20108B6681985-02-26026 February 1985 Supplementary Info to Addendum 1 to, Addendum to Bart A1: Computer Code for Best Estimate Analysis of Reflood Transients Project stage: Supplement NRC-85-3011, Forwards Supplementary Info to WCAP-9561,Addendum 1, Addendum to Bart A1:Computer Code for Best Estimate Analysis of Reflood Transients. Timely Completion of Review Needed to Support Anticipated Mar 1985 Startup1985-02-26026 February 1985 Forwards Supplementary Info to WCAP-9561,Addendum 1, Addendum to Bart A1:Computer Code for Best Estimate Analysis of Reflood Transients. Timely Completion of Review Needed to Support Anticipated Mar 1985 Startup Project stage: Supplement 1984-12-14
[Table View] |
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Category:REPORTS-TOPICAL (BY MANUFACTURERS-VENDORS ETC)
MONTHYEARML20211B1281999-08-31031 August 1999 Dynamic Rod Worth Measurement Using Casmo/Simulate ML20211F3441999-08-17017 August 1999 Updated non-proprietary Page 2-4 of TR DPC-NE-2009 ML20196L1881999-05-31031 May 1999 Non-proprietary Rev 1 to DPC-NE-3004, Mass & Energy Release & Containment Response Methodology ML20236U1601998-07-31031 July 1998 Non-proprietary DPC-NE-2009, DPC W Fuel Transition Rept ML20202D1921997-12-31031 December 1997 Rev 2 to UFSAR Chapter 15 Transient Analysis Methodology ML20197J3011997-11-30030 November 1997 Non-proprietary TR DPC-NE-3004-A, Mass & Energy Release & Containment Response Methodology ML20248L2631997-02-20020 February 1997 Non-proprietary Version of DPC-NE-2004-A, Duke Power Co McGuire & Catawba Nuclear Stations Core Thermal-Hydraulic Methodology Using VIPRE-01 ML20108D2091996-04-30030 April 1996 Nonproprietary App C, Mcguire/Catawba Plant Specific Data Mark-BW Fuel BWU-Z CHF Correlation to DPC Thermal- Hydraulic Statistical Core Design Methodology ML20107M8931995-10-31031 October 1995 Nonproprietary DPC Fuel Reconstitution Analysis Methodology ML20073J9151994-09-19019 September 1994 Replacement Steam Generator Topical Rept ML20070K0711994-06-30030 June 1994 Rev 1 to FSAR Chapter 15 Sys Transient Analysis Methodology ML20065D8061994-02-28028 February 1994 Analysis of Capsule V Specimens & Dosimeters & Analysis of Capsule Z Dosimeters from Duke Power Co McGuire Unit 1 Reactor Vessel Radiation Surveillance Program ML20064M4521994-02-28028 February 1994 WCAP-13948, Evaluation of Pressurized Thermal Shock for McGuire Unit 1 ML20116N0531992-11-11011 November 1992 Rev 1 to Mark-BW Reload LOCA Analysis for McGuire & Catawba ML20114A9771991-12-31031 December 1991 Nonproprietary McGuire & Catawba Nuclear Stations Core Thermal-Hydraulic Methodology Using VIPRE-01 ML20114A9911991-11-30030 November 1991 Nonproprietary Multidimensional Reactor Transients & Safety Analysis Physics Parameters Methodology ML20114A9711991-11-30030 November 1991 FSAR Chapter 15 Sys Transient Analysis Methodology ML20065M7381990-11-30030 November 1990 Mark-BW Reload LOCA Analysis for Catawba & McGuire Units ML20073H6281990-01-31031 January 1990 Nonproprietary Topical Rept DPC-NE-2001-A, Fuel Mechanical Reload Analysis Methodology for Mark-BW Fuel, Rev 1 ML20011E2541989-12-31031 December 1989 ARROTTA-HERMITE Code Comparison, Final Rept ML19351A4191989-09-30030 September 1989 Mark-BW Reload LOCA Analysis for Catawba & McGuire Units. ML20042F2321989-08-31031 August 1989 Nonproprietary DCHF-1 Correlation for Predicting Critical Heat Flux in Mixing Vane Grid Fuel Assemblies. ML19332C1531989-08-31031 August 1989 Analysis of Capsule X from Duke Power Co McGuire Unit 1 Reactor Vessel Radiation Surveillance Program. ML20235M9111989-02-20020 February 1989 Nonproprietary Catawba Unit 1 Evaluation for Tube Vibration Induced Fatigue ML20207L7781988-10-0606 October 1988 Nonproprietary WCAP 11936, McGuire Unit 2 Evaluation for Tube Vibration Induced Fatigue ML20207L9261988-09-30030 September 1988 Nonproprietary Addendum 2 to COBRA-NC,Analysis for Main Steamline Break in Catawba Unit 1 Ice Condenser Containment (Response to NRC Questions) ML20212B8841986-07-31031 July 1986 Nonproprietary Tubesheet Region Plugging Criterion for Duke Power Co McGuire Nuclear Station Units 1 & 2 Steam Generators ML20137L0591985-12-31031 December 1985 Nonproprietary Supplementary Info to WCAP-10585, Technical Bases for Eliminating Large Primary Loop Pipe Rupture on Structural Region Bases for McGuire Units 1 & 2 ML20127K1381985-06-30030 June 1985 Nuclear Physics Methodology for Reload Design ML20100H4581985-02-28028 February 1985 Analysis of Capsule U from Duke Power Co McGuire Unit 1 Reactor Vessel Radiation Surveillance Program ML20093N0811984-10-19019 October 1984 App a to Topical Rept DPC-NF-2010, Code Summary ML20135E2691984-06-30030 June 1984 Technical Bases for Eliminating Large Primary Loop Pipe Rupture as Structural Design Bases for McGuire Units 1 & 2 ML20092N5761984-06-30030 June 1984 Nonproprietary Technical Basis for Eliminating RHR Line Rupture as Structural Design Basis for Catawba Units 1 & 2 & McGuire Units 1 & 2 ML20078C2411983-07-31031 July 1983 Nonproprietary Preliminary Version of McGuire Post-Mod Steam Generator Tube Vibration Monitoring Program ML19331C6921980-07-31031 July 1980 Reliability Analysis of Auxiliary Feedwater Sys for McGuire Nuclear Station Unit 1. 1999-08-31
[Table view] Category:TEXT-SAFETY REPORT
MONTHYEARML20217G7951999-09-30030 September 1999 Monthly Operating Repts for Sept 1999 for McGuire Nuclear Station,Units 1 & 2 ML20217F3661999-09-22022 September 1999 Rev 18 to McGuire Unit 1 Cycle 14 Colr ML20212D1911999-09-20020 September 1999 SER Accepting Exemption from Certain Requirements of 10CFR50,App A,General Design Criterion 57 Closed System Isolation Valves for McGuire Nuclear Station,Units 1 & 2 ML20216E8851999-08-31031 August 1999 Monthly Operating Repts for Aug 1999 for McGuire Nuclear Station,Units 1 & 2 ML20211B1281999-08-31031 August 1999 Dynamic Rod Worth Measurement Using Casmo/Simulate ML20217G8101999-08-31031 August 1999 Revised Monthly Operating Repts for Aug 1999 for McGuire Nuclear Station,Unit 1 & 2 ML20211G5261999-08-24024 August 1999 SER Accepting Approval of Second 10-year Interval Inservice Insp Program Plan Request for Relief 98-004 for Plant,Unit 1 ML20211F3441999-08-17017 August 1999 Updated non-proprietary Page 2-4 of TR DPC-NE-2009 ML20210S2371999-07-31031 July 1999 Monthly Operating Repts for July 1999 for McGuire Nuclear Station,Units 1 & 2 ML20216E8951999-07-31031 July 1999 Revised Monthly Operating Repts for July 1999 for McGuire Nuclear Station,Units 1 & 2 ML20209E4361999-07-0909 July 1999 SER Agreeing with Licensee General Interpretation of TS LCO 3.0.6,but Finds No Technical Basis or Guidance That Snubbers Could Be Treated as Exception to General Interpretation ML20196K6631999-07-0707 July 1999 Safety Evaluation Supporting Licensee 990520 Position Re Inoperable Snubbers ML20209H1631999-06-30030 June 1999 Monthly Operating Repts for June 1999 for McGuire Nuclear Station,Units 1 & 2 ML20210S2491999-06-30030 June 1999 Revised Monthly Operating Rept for June 1999 for McGuire Nuclear Station,Units 1 & 2 ML20209H1731999-05-31031 May 1999 Revised Monthly Operating Rept for May 1999 for McGuire Nuclear Station,Units 1 & 2 ML20206T4771999-05-31031 May 1999 Rev 3 to UFSAR Chapter 15 Sys Transient Analysis Methodology ML20196L1881999-05-31031 May 1999 Non-proprietary Rev 1 to DPC-NE-3004, Mass & Energy Release & Containment Response Methodology ML20195K3691999-05-31031 May 1999 Monthly Operating Repts for May 1999 for McGuire Nuclear Station,Units 1 & 2 ML20206N3511999-05-11011 May 1999 Safety Evaluation Accepting Licensee Response to GL 96-05, Periodic Verification of Design-Basis Capability of Safety- Related Movs ML20195K3761999-04-30030 April 1999 Revised MORs for Apr 1999 for McGuire Nuclear Station,Units 1 & 2 ML20206R0891999-04-30030 April 1999 Monthly Operating Repts for Apr 1999 for McGuire Nuclear Station,Units 1 & 2 ML20205L2341999-04-0505 April 1999 SFP Criticality Analysis ML20206R0931999-03-31031 March 1999 Revised Monthly Repts for Mar 1999 for McGuire Nuclear Station,Units 1 & 2 ML20205P8991999-03-31031 March 1999 Monthly Operating Repts for Mar 1999 for McGuire Nuclear Station,Units 1 & 2 ML20205C4171999-03-25025 March 1999 Special Rept 99-02:on 801027,Commission Approved for publication,10CFR50.48 & 10CFR50 App R Delineating Certain Fire Protection Provisions for Nuclear Power Plants Licensed to Operate Prior to 790101.Team Draft Findings Reviewed ML20207K2051999-03-0505 March 1999 Special Rept 99-01:on 990128,DG Tripped After 2 H of Operation During Loaded Operation for Monthly Test.Caused by Several Components That Were Degraded or Had Intermittent Problems.Parts Were Replaced & Initial Run Was Performed ML20204C8911999-02-28028 February 1999 Monthly Operating Repts for Feb 1999 for McGuire Nuclear Station,Units 1 & 2 ML20205P9021999-02-28028 February 1999 Revised Monthly Operating Repts for Feb 1999 for McGuire Nuclear Station,Units 1 & 2 ML20204C8961999-01-31031 January 1999 Revised Monthly Operating Repts for Jan 1999 for McGuire Nuclear Station,Units 1 & 2 ML20199E0301998-12-31031 December 1998 Monthly Operating Repts for Dec 1998 for McGuire Nuclear Station,Units 1 & 2 ML20216F9931998-12-31031 December 1998 Piedmont Municipal Power Agency 1998 Annual Rept ML20198A4481998-12-11011 December 1998 Safety Evaluation Concluding That for Relief Request 97-004, Parts 1 & 2,ASME Code Exam Requirements Are Impractical. Request for Relief & Alternative Imposed,Granted ML20198D7561998-11-30030 November 1998 Monthly Operating Repts for Nov 1998 for McGuire Nuclear Station,Units 1 & 2 ML20199E0491998-11-30030 November 1998 Revised Monthly Operating Rept for Nov 1998 for McGuire Nuclear Station,Units 1 & 2 Re Personnel Exposure ML20199E9651998-11-24024 November 1998 Rev 1 to ATI-98-012-T005, DPC Evaluation of McGuire Unit 1 Surveillance Weld Data Credibility ML20196D4171998-11-24024 November 1998 Special Rept 98-02:on 981112,failure to Implement Fire Watches in Rooms Containing Inoperable Fire Barrier Penetrations,Was Determined.Repair of Affected Fire Barriers in Progress ML20196G0581998-11-0606 November 1998 Rev 17 to COLR Cycle 13 for McGuire Unit 1 ML20196G0761998-11-0606 November 1998 Rev 15 to COLR Cycle 12 for McGuire Unit 2 ML20198D7771998-10-31031 October 1998 Revised Monthly Operating Rept for Oct 1998 for McGuire Nuclear Station,Units 1 & 2 ML20195E5961998-10-31031 October 1998 Monthly Operating Repts for Oct 1998 for McGuire Nuclear Station,Units 1 & 2 ML20154L6251998-09-30030 September 1998 Monthly Operating Repts for Sept 1998 for McGuire Nuclear Station,Units 1 & 2 ML20195E6021998-09-30030 September 1998 Revised Monthly Operating Rept for Sept 1998 for McGuire Nuclear Station,Units 1 & 2 ML20154B4131998-09-22022 September 1998 Rev 0 to ISI Rept for McGuire Nuclear Unit 1 Twelfth Refueling Outage ML20151W3521998-09-0808 September 1998 Special Rept 98-01:on 980819,maint Could Not Be Performed on FPS Due to Isolation Boundary Leakage.Caused by Inadequate Info Provided in Fire Impairment Plan.Isolated Portion of FPS Was Returned to Svc ML20154L6321998-08-31031 August 1998 Rev 1 to MOR for Aug 1998 for McGuire Nuclear Station,Unit 1 ML20153B3741998-08-31031 August 1998 Monthly Operating Repts for Aug 1998 for McGuire Nuclear Station,Units 1 & 2 ML20236U1601998-07-31031 July 1998 Non-proprietary DPC-NE-2009, DPC W Fuel Transition Rept ML20237B2381998-07-31031 July 1998 Monthly Operating Repts for July 1998 for McGuire Nuclear Station,Units 1 & 2 ML20153B3931998-07-31031 July 1998 Revised Monthly Operating Repts for Jul 1998 for McGuire Nuclear Station,Units 1 & 2 ML20236P0451998-07-0808 July 1998 Part 21 Rept Re non-conformance & Potential Defect in Component of Nordberg Model FS1316HSC Standby Dg.Caused by Outer Spring Valves Mfg from Matl That Did Not Meet Specifications.Will Furnish Written Rept within 60 Days 1999-09-30
[Table view] |
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ATTACHMENT SUPPLEMENTARY INFORMATION l
TO WCAP-9561 ADDENDUM 1 8503070428 850226 PDR ADOCK 05000370 P PDR 8289Q:10/022185
- 1. PRESSURE During the reflood portion of the transient, the BART code assumes a constant pressure as shown in Figure 1. This pressure of 16 psia is below the lower limit of the approved pressure band which is 20 psia. Using a pressure outside the approved band makes it necessary to demonstrate that the BART calculations are still conservative.
In the FLECHT Skewed Test Series, tests 13404 and 13609 have identical conditions except for system pressure. The pressure in test 13404 was 40 psia and the pressure in test 13609 was 20 psia. The clad temperatures in the data and the BART calculations of the tests will be examined to show that lowering
. the pressure retains the conservatism in BART. Figure 2 shows the clad temperature trani.ients at the 6 foot, 8 foot, and 10 foot elevations for both the test data and the BART calculations. The peak clad temperatures are listed in the following table.
\
Data Data BART BART Test 13609 Test 13404 Test 13609 Test 13404
' ' Elevation 20 psia 40 psia 20 psia 40 psia (feet) PCT (*F) PCT (*F) PCT (*F) PCT (*F) 6 1591 1569 1614 1608 8 1731 1788 1829 1882 10 1715 1851 2000 2088 In these tests, BART conservatively overpredicts the clad temperatures.
Figure 3 compares the BART predicted cladding temperatures to the measured
' cladding temperatures. This figure shows that reducing the pressure does not adversely affect the conservatism in BART. Thus, the consequences of using 16 psia instead of 20 psia are that the BART calculations would still be conservative.
8289Q:10/022185
l l
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E do - APPROVED BAND E
to -__________.v - _ _ _ _ .
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Figure 1 BART Pressure Compared to Approved Band i
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Figure 2 Clad Temperature Transients in Tests 13404 and 13609 .
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TIME tstC ,,
TIME estC b ATA AT 4 F1 Ete. d Tsor4 BART 41 G. GT ELEJATiW aus. i s aass.
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TIME ISEC) TIME ESEC1 b AT A AT go FT EL2#7 ton BART A7 to FT tt EdM'4
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isce loco I?e teco 1900 2000 ruo 2 tee bAM C.L% b WMPELRA tuRE (,'Fh l
l Figure 3 BART vs. DATA Clad Temperature i -
_ _ _ _ _ r; l
- 2. INITIAL PEAK ROD TEMPERATURE t
The initial Riak rod temperature concerns only the maximum of the average rod temperature at 80C. Typical initial temperatures of the cladding a.re shown in Figure 4. The magnitude of the peak average rod temperature may vary
! depending on the break being analyzed. However, these variations may only slightly exceed or fall below the approved range.
A number of tests and BART predictions of the tests will be used to demonstrate that BART is conservative over a wide range of initial peak temperature. The sets of experimental data included:
A. FLECHT-SEASET 161 Rod Tests: 30817 31203 31805 32235 32333 B. G-2 Tests: 538 561 C. FLECHT Cosine Power Tests: 4831 5132 5342 6638 7934 Comparisons between BART and the data will be made using the cladding temperature rise ratio, A.
(TPEAK - INITIAL) BART (T PEAK ~ INITIAL}
^^
?
8289Q:10/022185 L
If this ratio is greater than 1, BART censervatively overpredicts the cladding temperature rise. Figure 5 plots this ratio as a function of initial temperature. The points on the figure can be correlated in the form of a straight line using a least squares fit. The resulting correlation is:
A = 1.323 - 0.0001348* T NTM The small coefficient of T indicates that the conservatism in BART is NTM not very sensitive to initial temperature. This correlation can be used to determine the upper limit of T f r which A is greater than 1. The INITIAL calculated upper limit of T is 2396*F. Thus, for initial INITIAL temperatures lower than 2396*F, BART conservatively overpredicts the cladding temperature rise.
8289Q:10/022185
{1eoo -
g1 o ._ _ _ _ _ _ _ ___ _ _I._ _ _ _ _ _
g ,,,, _ sREAn SPECIFIC VARIATIONS APPROVED SAND FOR PEAK lia0o -
7 .
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E 8 '"
4.o -
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I I I I I I I I O I I I O 1 '2 3 4 5 6 7 8 e io 11 12 AX1AL ELEVATION (m -
Figure 4 Initial Rod Tesiperatures Compared to Approved Band 2.2 ; ; q
- 2. l l l
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Figure 5 Cladding Temperature Rise Ratio as a Function of Initial j Temperature
- 3. REFLOOD RATE BART has been approved for transients in which the core reflood rate varies between 0.6 in./sec and 1.5 in./sec. The core reflood rate for a s, ample UHI transient is compared to the approved band in Figure 6. This comp ~arison shows that using BART for UHI applications remains within the NRC limits on core reflood rates except f ~ r a spike where the reflood rate drops to 0.5 in./sec.
l Heat transfer coefficients in BART are calculated using the mass flow rate through the core. The core mass flow rate through the core is approximately proportional to the product of the flooding rate, liquid density, and mass ,
. entrainment. When the reflood rate spiked down to 0.5 in./sec the mass I entrainment fraction concurrently increased. The mass flow rate through the core remained essentially the same even though the flooding rate spiked down.
This can be seen in Figures 7 and 8. Thus, the calculated heat transfer coefficient would not be significantly affected by the downward spike in flooding rate.
8289Q:lD/022185
1.so - _ _ _ - - _ _ ,, _ _ _ _ .
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o i 0 100 200 soo soo TIME (SEC)
Figure 6 Reflood Rate Compared to Approved Band l
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. 68 -
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o.o To 92 ft PG 98 do roz rov ros me no 7;ms 6dch Figure 7 Core Flow Rate g40 P
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to 91 9% 99 fB te 1o2. m so,, ta6 no Tond (sn')
i Figure 8 Core Heat Transfer Coefficier.ts
. . __. _ _ . - . - . . _ - - - . _ _ . . _ . _ _ _ . - - - - - _ - - - -