ML12067A180
| ML12067A180 | |
| Person / Time | |
|---|---|
| Site: | Harris  |
| Issue date: | 02/29/2012 |
| From: | AREVA NP |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| HNP-12-023, TAC ME6999 ANP-3011Q1(NP), Rev 000 | |
| Download: ML12067A180 (147) | |
Text
HNP-12-023 Shearon Harris Nuclear Power Plant, Unit No. 1 Docket No. 50-400 / Renewed License No. NPF-63 License Amendment Request for Revision to Technical Specification Core Operating Limits Report (COLR) References for Realistic Large Break LOCA Analysis (TAC NO. ME6999)Response to Request for Additional Information Enclosure 4 Response to Request for Additional Information (Non-Proprietary)
ANP-301 1Q1 (NP)Revision 000 Harris Nuclear Plant Unit 1 Realistic Large Break LOCA Analysis February 2012 A AR EVA AREVA NP Inc.
A AR EVA Harris Nuclear Plant Unit 1 Realistic Large Break LOCA Analysis ANP-3011Q1(NP)
Revision 000 Page ii Copyright
© 2012 AREVA NP Inc.All Rights Reserved AREVA NP Inc.
A AREVA Harris Nuclear Plant Unit 1 Realistic Large Break LOCA Analysis ANP-3011Q1(NP)
Revision 000 Page iii Nature of Changes Description and Justification Item Page 1. All This is a new document.AREVA NP Inc.
A AREVA Harris Nuclear Plant Unit 1 Realistic Large Break LOCA Analysis ANP-3011Q1(NP)
Revision 000 Page iv Table of Contents Page NATU RE O F CHANGES .........................................................................................................................
III LIST O F TABLES ....................................................................................................................................
V LIST O F FIG URES .................................................................................................................................
VI 1.0 INTRO DUCTIO N ...........................................................................................................................
1 2.0 NRC REVIEW COMMENTS AND AREVA NP'S RESPONSES
..............................................
2 2.1 NRC Q1 ............................................................................................................................................
2 2.2 NRC Q2 ..........................................................................................................................................
10 2.3 NRC Q3 ..........................................................................................................................................
11 2.4 NRC Q4 ..........................................................................................................................................
13 2.5 NRC Q5 ..........................................................................................................................................
15 2.6 NRC Q6 ..........................................................................................................................................
20 2.7 NRC Q7 ..........................................................................................................................................
25 2.8 NRC Q8 ..........................................................................................................................................
31 2.9 NRC Q9 ..........................................................................................................................................
33 2.10 NRC Q10 ........................................................................................................................................
34 2.11 NRC Qll ........................................................................................................................................
37 2.12 Sleicher-Rouse Error Adjustment
..............................................................................................
42 3.0 HARRIS PCT SUM MARY ...........................................................................................................
43
4.0 REFERENCES
............................................................................................................................
44 APPENDIX A : EMF 2100 SECTIONS 3AND 4 ..................................................
4.........................................
A-1 APPENDIX B : FLECHT-SEASET DATA .....................................................................................................
B-1 APPENDIX C: M5 HIGH TEMPERATURE SWELLING AND RUPTURE MODEL .....................................
C-1 This document contains a total of 146 pages.AREVA NP Inc.
A AR EVA ANP-3011Q1(NP)
Harris Nuclear Plant Unit 1 Revision 000 Realistic Large Break LOCA Analysis Page v List of Tables Page Table 2-1: Results of Swelling and Rupture Sensitivities, No Relocation
...........................................
4 Table 2-2: 0.5 Packing Fraction Cases with Hot Assembly Rupture ..................................................
4 Table 2-3: 0.6 Packing Fraction Cases with Hot Assembly Rupture ..................................................
5 Table 2-4: 0.7 Packing Fraction Cases with Hot Assembly Rupture ..................................................
5 Table 2-5: 0.8 Packing Fraction Cases with Hot Assembly Rupture ..................................................
6 Table 2-6: Halden Test IFA 650.4 Post Irradiation Exam Data ..........................................................
36 Table 3-1: HN P PCT Rackup ...........................................................................................................
43 AREVA NP Inc.
A AREVA ANP-3011Q1(NP)
Harris Nuclear Plant Unit 1 Revision 000 Realistic Large Break LOCA Analysis Page vi List of Figures Page Figure 2-1: Sensitivity Study Results for Swell, Rupture, and Relocation Simulation
.........................
6 Figure 2-2: Histogram of PCT with Droplet Shattering Activated
-0.7 Packing Fraction ...................
7 Figure 2-3: Rupture Node Cladding Temperature Response for Limiting Case -57 ..........................
8 Figure 2-4: Rupture Node Cladding Temperature Response for Limiting Case -57, To Quench ..........
9 Figure 2-5: FLECHT-SEASET Test 61509 ........................................................................................
14 Figure 2-6: FLECHT-SEASET Test 61607 ........................................................................................
14 Figure 2-7: Rupture Phenom ena ......................................................................................................
16 Figure 2-8: REBEKA-6 Test Results ..................................................................................................
22 Figure 2-9: REBEKA-7 Test Results Maximum Blockage -70% .....................................................
23 Figure 2-10: FEBA Test Results ......................................................................................................
24 Figure 2-11: Test 61607 Benchmark
-Droplet Diameter and Void Fraction at Peak Power Location ..26 Figure 2-12: FLECHT-SEASET Test 61607, Void Fraction Comparison, -4 ft (below blockage)
.........
27 Figure 2-13: FLECHT-SEASET Test 61607, Void Fraction Comparison, -5 ft (below blockage)
.........
27 Figure 2-14: FLECHT-SEASET Test 61607, Void Fraction Comparison, -6 ft (approx. blockage lo ca tio n ) ............................................................................................................................................
2 8 Figure 2-15: FLECHT-SEASET Test 61607, Vapor Temperature, -5 ft (below blockage)
...............
28 Figure 2-16: FLECHT-SEASET Test 61607, Vapor Temperature, -6.5 ft (above blockage)
...........
29 Figure 2-17: FLECHT-SEASET Test 61607, Vapor Temperature, -8 ft (above blockage)
..............
29 Figure 2-18: FLECHT-SEASET Test 61607, Total Heat Transfer Coefficient, -6 ft .........................
30 Figure 2-19: Test Packing Factor vs. Rupture Strain ........................................................................
35 Figure 2-20: Exam ple of Power History .............................................................................................
39 Figure 2-21: Decay Heat Curve, Burnup = 1 GWd/MTU ..................................................................
41 AREVA NP Inc.
A AR EVA ANP-3011Q1(NP)
Harris Nuclear Plant Unit 1 Revision 000 Realistic Large Break LOCA Analysis Page 1
1.0 INTRODUCTION
AREVA NP Inc. performed an RLBLOCA analysis for the Harris Nuclear Plant (HNP) Unit 1.The plant is a Westinghouse 3-loop design with a rated thermal power of 2900 MWt and a pending uprate to a reated thermal power of 2948 MWt based on measurement uncertainty recapture (MUR). HNP has a dry atmospheric containment.
Implementation of a full core of AREVA NP 17x17 HTP fuel design at HNP is simulated in the analysis.HNP operation at a safety analysis power level of 2958 MWt (rated thermal power plus uncertainty), a steam generator tube plugging level of up to 3 percent in all steam generators, a total peaking factor (FQ) of 2.52 (including uncertainty) and a nuclear enthalpy rise factor (FAH) of 1.73 (including uncertainty) with no axial or burnup dependent power peaking limit is supported by the analysis.
The analysis demonstrates compliance with regulatory requirements for large break via application of an NRC-approved evaluation methodology
[1], predicting acceptable peak cladding temperature, oxidation thickness, and hydrogen generation (Summary Report, [2]).Progress Energy submitted the RLBLOCA Summary Report to the NRC for review. Section 2.0 contains AREVA NP Inc.'s responses to NRC comments and Section 3.0 is the Summary.AREVA NP Inc.
A AREVA ANP-3011Q1(NP)
Harris Nuclear Plant Unit 1 Revision 000 Realistic Large Break LOCA Analysis Page 2 2.0 NRC REVIEW COMMENTS AND AREVA NP'S RESPONSES 2.1 NRC Q1.The submitted ECCS evaluation is not clear as it does not appear to consider the droplet shattering model separate from the fuel relocation study. See, for instance, Figure 6-17.a. Please characterize the impact of enabling the droplet shattering model without modeling the fuel relocation.
- b. Similarly, please show the sensitivity of the fuel relocation model study to fuel relocation packing factor. Consideration should be made to include a range of packing factors that is more inclusive of the available data, i.e., 30-80 percent. It would be beneficial if this sensitivity is shown using the Shearon Harris Nuclear Plant (SHP) limiting-peak clad temperature (PCT)case, and using an evaluation model variant that does not include the droplet shattering enhancement.
Response: The base case results, of ANP-301 1, are compared with sensitivity studies characterizing swelling and rupture with no relocation of fuel fragments in Table 2-1. The limiting case for each set of 59 cases is compared and indicates a reduction in PCT with nominal, associated reductions in oxidation results. Sensitivity studies were also performed simulating swelling and rupture with relocation.
Table 2-2, Table 2-3, Table 2-4, and Table 2-5 report the results of cases resulting in hot assembly rupture for homogeneous fuel fragment packing fractions of 0.5, 0.6, 0.7, and 0.8 respectively.
Comparisons are made to cases simulating heat transfer enhancement associated with droplet shattering on the cladding rupture blockage.Figure 2-1 illustrates the results of the expanded sensitivity studies in a range of packing fractions from 0.0 to 0.8.It is recognized that there is some credit for steam de-superheating via droplet breakup and related heat transfer enhancement.
It is also recognized that the packing fraction of 0.8 will not occur in the Harris RLBLOCA cases due to the M5 cladding and strains observed in the rupture cases. Table 3-1 contains the resulting PCT including penalty that results from the position taken by the NRC.Figure 2-2 is a histogram of cases examined as part of the sensitivity studies performed to simulate swelling, rupture, and relocation.
Specifically, this figure reports the results of a relocation study, packing factor of 0.7, with simulation of droplet shattering.
Only two of the 59 cases were related to PCT results in excess of 1800 F. The histogram is characteristic of all the sensitivity studies performed.
AREVA NP Inc.
A AREVA Harris Nuclear Plant Unit 1 Realistic Large Break LOCA Analysis ANP-3011Q1(NP)
Revision 000 Page 3 Figure 2-3 illustrates the rupture node cladding temperature response for the limiting case associated with each of the sensitivity studies. Hot pin pressure is also plotted in this figure to indicate the time of rupture. Note that the slope of the temperature response during adiabatic heatup -15 to 30 seconds increases significantly with increasing packing factors [I Figure 2-4 is included to demonstrate successful clad quenches for the limiting case results.AREVA NP Inc.
A AR EVA Harris Nuclear Plant Unit 1 Realistic Large Break LOCA Analysis ANP-3011Q1(NP)
Revision 000 Page 4 Table 2-1: Results of Swelling and Rupture Sensitivities, No Relocation Table 2-2: 0.5 Packing Fraction Cases with Hot Assembly Rupture AREVA NP Inc.
A AREVA Harris Nuclear Plant Unit 1 Realistic Large Break LOCA Analysis ANP-3011Q1(NP)
Revision 000 Page 5 Table 2-3: 0.6 Packing Fraction Cases with Hot Assembly Rupture Table 2-4: 0.7 Packing Fraction Cases with Hot Assembly Rupture AREVA NP Inc.
A AREVA Harris Nuclear Plant Unit 1 Realistic Larae Break LOCA Analvsis ANP-3011Q1 (NP)Revision 000 Paae 6 Table 2-5: 0.8 Packing Fraction Cases with Hot Assembly Rupture Figure 2-1: Sensitivity Study Results for Swell, Rupture, and Relocation Simulation AREVA NP Inc.
A AREVA Harris Nuclear Plant Unit 1 Realistic Large Break LOCA Analysis ANP-3011Q1(NP)
Revision 000 Page 7 Figure 2-2: Histogram of PCT with Droplet Shattering Activated
-0.7 Packing Fraction AREVA NP Inc.
A AREVA Harris Nuclear Plant Unit 1 Realistic Larae Break LOCA Analysis ANP-3011Q1(NP)
Revision 000 Paae 8 Figure 2-3: Rupture Node Cladding Temperature Response for Limiting Case -57 AREVA NP Inc.
A AREVA Harris Nuclear Plant Unit 1 Realistic Larme Break LOCA Analysis ANP-3011Q1(NP)
Revision 000 Paae 9 Figure 2-4: Rupture Node Cladding Temperature Response for Limiting Case -57, To Quench AREVA NP Inc.
A AREVA Harris Nuclear Plant Unit 1 Realistic Large Break LOCA Analysis ANP-3011Q1(NP)
Revision 000 Paae 10 2.2 NRC Q2.Address whether droplet shattering is calculated on all flow blockage (non-vertical) surfaces in the S-RELAP5 calculation.
If not, provide the flow blockage surfaces which are assumed to cause droplet shattering.
Response: Response is addressed in Section 2.8 (NRC Q8).AREVA NP Inc.
A AREVA Harris Nuclear Plant Unit 1 Realistic Larqe Break LOCA Analysis ANP-3011Q1(NP)
Revision 000 Pace 11 2.3 NRC Q3.Page 122 of ANP-301 1(P), "Harris Nuclear Plant Unit 1 Realistic Large Break LOCA Analysis" states, In the present model, the rupture blockage ratio [which is correlated to the number of droplets to yield a maximum atomization factor], C, is taken from the swelling and rupture correlation.
- a. Address whether droplet shattering is calculated only against the hot pin rupture, or the additional flow blockage areas (i.e., balloon/burst regions, spacer grids, etc.) assumed to be present upstream of the hot pin rupture location.b. If the additional flow blockage areas are not based on pre-transient core geometry, discuss how the locations and sizes of flow blockages are distributed.
Response] Appendix C of BAW-1 0227PA has been added as Appendix C of this report. []The blockage area is taken from the NRC approved deterministic swelling and rupture model for M5 cladding presented in BAW-1 0227P-A Appendix C (Reference
[9], see Appendix C).The modeling follows the approach utilized in NUREG-0630:
- Pin rupture temperature is correlated to pin engineering hoop stress," Pin strain is then correlated to pin rupture temperatures, AREVA NP Inc.
A AR EVA Harris Nuclear Plant Unit 1 Realistic Larae Break LOCA Analysis ANP-3011Q1 (NP)Revision 000 Paae 12* Pin strain remote from the rupture is correlated to the proximity of pin temperature to pin rupture temperature.
This provides sufficient characterization for the hot pin.I AREVA NP Inc.
A A R EVA ANP-3011Q1 (NP)Harris Nuclear Plant Unit 1 Revision 000 Realistic Large Break LOCA Analysis Page 13 2.4 NRC Q4.On page 123 of ANP-3011(P), it is stated, It can be seen that the code predicted the peak cladding temperature variation well.The data is so tightly clustered that the degree of agreement is difficult to ascertain.
Please tabulate the data to provide a more quantitative indication.
Address how well the S-RELAP5 modification predicted the data.Response* Table B-1 and Table B-2 contain FLECHT-SEASET Test 61607 blockage data.* Figures 6-14 and 6-15 in ANP-301 1 P, Reference
[2], show S-RELAP5 conservatively predicted PCT compared to blockage test data for the FLECHT-SEASET and REBEKA-6 tests." Additional FLECHT-SEASET test benchmark results for Tests 61509 and 61607 are presented in Figure 2-5 and Figure 2-6.* COBRA-TF (NUREG/CR-4166, Reference
[5]) modeled all expected multi-dimensional flow phenomena at rupture location, including flow diversion." The figures show that S-RELAP5 conservatively predicts cladding response compared to the test data as well as the COBRA-TF results.AREVA NP Inc.
A AREVA Harris Nuclear Plant Unit 1 Realistic Large Break LOCA Analysis ANP-3011Q1(NP)
Revision 000 Page 14 Figure 2-5: FLECHT-SEASET Test 61509 Figure 2-6: FLECHT-SEASET Test 61607 AREVA NP Inc.
A AREVA Harris Nuclear Plant Unit 1 Realistic Large Break LOCA Analysis ANP-3011Q1(NP)
Revision 000 Page 15 2.5 NRC Q5.Comparative data demonstrate the global effects of the droplet shattering phenomena; however, the correlation as implemented discriminates between large and small droplets and the behavioral differences between the two. Validate droplet size distribution as implemented in model. Explain how the Sugimoto/Murao correlation applies to the scenario in which it is applied.Response Note that, for the Harris Cycle 18 RLBLOCA application, the theory regarding the mechanics and heat transfer aspects of droplet shattering in mist flow is based on Sections 3 and 4 of Reference
[3]. For ease of access, these sections have been excised and included in this document as Appendix A.The text below is taken from ANP-3011 (Reference
[2]) and includes clarifications regarding the use of the Sugimoto-Murao correlation (Reference
[8]) and how the droplet breakup model is incorporated in the current droplet heat transfer studies. The dispersed flow heat transfer sections from EMF-2100 (Reference
[3]) is included in Appendix A for ease of access.] The following drawing in Figure 2-7 illustrates the phenomena.
AREVA NP Inc.
A A REVA ANP-30 11 Q1 (NP)Harris Nuclear Plant Unit 1 Revision 000 Realistic Large Break LOCA Analysis Page 16 GRID STRAP 0 o 00. MICROOROPS 0*oi Figure 2-7: Rupture Phenomena Paik, Hochreiter, Kelly, and Kohrt examined the FLECHT-SEASET and FEBA tests in detail, Section 4, Reference
[5] , and concluded that the flow blockage heat transfer effects are essentially a combination of the two key thermal-hydraulic phenomena identified in the REBEKA summary report. They separated the blockage heat transfer effects into three phenomena for the purpose of the COBRA-TF model development:
- Single phase convective enhancement
- Droplet break-up" Droplet Impact COBRA-TF includes all three of these heat transfer enhancement phenomena either within the grid or within the rupture cooling models, Sections 3 and 4 of Reference
[5]. Convective enhancement and droplet break-up are included in the grid modeling.
All three are present in the model for rupture induced cooling. The droplet break-up and convective enhancement models implemented in COBRA-TF for rupture-induced cooling are similar to those employed for grid spacers. Droplet impact heat transfer accounts for the droplet-wall interaction with the pre-rupture-strained area of the cladding, where the channels are converging with a slight angle and is not appropriate for application to grid cooling models. From the sensitivity studies using FLECHT-SEASET blockage tests, the COBRA-TF developers concluded that the droplet impact heat transfer effect is very small compared to convective enhancement and droplet break-up.The modeling of droplet break-up in COBRA-TF, is based on the droplet impact Weber number formulation derived from the droplet impact tests on heated flat plates by Watchers, et al., References
[6] and [7]. The analogy, flat plate to balloon, comes about because the momentum of liquid droplets at high velocity prevents diversion around the balloon leaving the droplets to impinge on the balloon and shatter into smaller diameter droplets and micro-droplets.AREVA NP Inc.
A AR EVA Harris Nuclear Plant Unit 1 Realistic Large Break LOCA Analysis ANP-3011Q1(NP)
Revision 000 Page 17 In COBRA-TF the micro-droplets are treated as a separate population and are evaporated within the rupture zone, resulting in cooler steam at, and downstream from, the rupture zone.The quantity of micro-drops generated is expressed as a function of the entrained drop flow (me) and the blockage area.mDB ='(ADB/Ac)
- me ADB is the projected area of the blockage, Ac is the flow area of the channel, and me is entrained droplet flow. The blockage efficiency factor, -d, represents the portion of droplets within the projected area that are shattered into micro-drops.
The value of 'Y was determined using FLECHT-SEASET test benchmarks.
The ratio, ADB/Ac can be considered as proportional to the bundle blockage factor, C.The BEACH rupture cooling model incorporated in S-RELAP5 is based on the COBRA-TF modeling approach.
[During the majority of a LOCA event, the flow through the core is comprised of steam with entrained droplets of water or of single phase vapor. Most of the time the flow field is comprised of steam with dispersed droplets.
When such a flow interacts with a solid structure protruding into the flow field, the droplets impinge on the structure and shatter into a population of smaller droplets.
The degree of shattering is dependent on the blockage offered by the AREVA NP Inc.
A AREVA Harris Nuclear Plant Unit 1 Realistic Larce Break LOCA Analysis ANP-3011Q1(NP)
Revision 000 Paae 18 structure.
[] This approach is used in BEACH applications and is consistent with the approach used in COBRA-TF applications.
AREVA NP Inc.
A AREVA Harris Nuclear Plant Unit 1 Realistic Larae Break LOCA Analysis ANP-3011Q1(NP)
Revision 000 Paae 19 Summarizing this section, the BEACH and S-RELAP5 rupture cooling models are consistent with those incorporated in COBRA-TF.
In the Technical Evaluation Report of BEACH, Revision 2 of Reference
[4], the INEL reviewer evaluated and accepted the adequacy of the hydrodynamic blockage effects of cladding rupture including their adaption from a grid cooling model. In addition, several technical papers, References
[11], [12], [13], [14], and [15] have been presented indicating that the rupture cooling model added in BEACH and now S-RELAP5, has been accepted by the thermal-hydraulics community.
AREVA NP Inc.
A AR EVA Harris Nuclear Plant Unit 1 Realistic Large Break LOCA Analysis ANP-3011Q1(NP)
Revision 000 Paae 20 2.6 NRC Q6.Explain how droplet shattering model incorporates the following droplet-dependent heat transfer effects: a. Inter-phase heat transfer b. Fluid-structural interactions including cladding, balloon, and spacer heat transfer to coolant c. Validate heat transfer modeling for these separate effects.Response a. Inter-phase heat transfer The implementation of droplet shattering model is described in detail as part of the response to NRC question, Q5.b. Fluid-structural interactions including cladding, balloon, and spacer heat transfer to coolant.AREVA NP Inc.
A AREVA ANP-3011Q1 (NP)Harris Nuclear Plant Unit 1 Revision 000 Realistic Large Break LOCA Analysis Page 21 c. Validate heat transfer modeling for these separate effects The FLECHT-SEASET Blockage Tests 61607 and 61509, and REBEKA-6 test assessment used to demonstrate the droplet shatter effects.* Substantial liquid-cladding interaction has been observed in REBEKA, FEBA, and CEGB tests." REBEKA tests showed liquid accumulation at rupture location, resulting in quenching of cladding near rupture and this quench front propagates up-stream and downstream from the rupture location (see Figure 2-8 and Figure 2-9). This phenomenon is described in detail in Reference
[16] and Reference
[17]." CEGB Berkeley Lab concluded that (using high speed cine and T.V. recording) for tests with blockages up to 90%, majority of water droplets passed through the blockages.
For severe blockages, significant liquid coalescence and entrainment in the blockage region." FEBA tests showed substantial heat transfer improvements in the blockage regions, with blockages up to 90% (see Figure 2-10 extracted from NUREG/CR-4166, Reference[5]).* All these tests clearly show that liquid droplets penetrate the thermal boundary layer near the cladding surface." Dr. Hochreiter reviewed all the available reflood tests with blockages, and implemented models for single phase turbulent enhancement, droplet break-up and droplet impact heat transfer on the blockage entrance region in COBRA-TF (NUREG/CR-4166, Reference
[5]).AREVA NP Inc.
A AREVA Harris Nuclear Plant Unit 1 Realistic Larae Break LOCA Analysis ANP-3011Q1(NP)
Revision 000 PacQe 22 4-C 1000 900o O 800-S700 600.500 400 300 200-100 0 4.4.I..a.0 4..4.K 100 90*80 70 60 50 40 30 20 10 quench burst start of flooding sequence cad no. 29 bursLt (3)I (2)11)93.5s rod no. 54_1_5 s TC-position (mm)170-grid 4 S1800 1850 ,(1910) burst r1950 S2050 210 ri nI-30 0 30) 60' 90 12'0 15'0 18'0 21'0 24'0 27 .0 300 time (s)Figure 2-8: REBEKA-6 Test Results AREVA NP Inc.
A AR EVA Harris Nuclear Plant Unit 1 Realistic Large Break LOCA Analysis ANP-3011Q1(NP)
Revision 000 Page 23 I axial level: 2 1 00mm I Li 0+10 900 000 700 600 Soo 400 300 200 100 0 90 8o -70 I LAýO a-o.50 40 30-20 10 Near rupture V axial level: 1810 mm max. flow blockage:
-70 %I 0 0 30 60 90 120 150 100 210 2~0 0 0 60 9,0 10 iS'0 1,80 2110 2,40 time, s Figure 2-9: REBEKA-7 Test Results Maximum Blockage -70%AREVA NP Inc.
A AR EVA Harris Nuclear Plant Unit 1 Realistic Larme Break LOCA Analysis ANP-3011Q1(NP)
Revision 000 Page 24 orn~2 I.2 I I-I'I F LOOOMM KATE -Ga ~I &) ýJw V*~kf ew fcm w ap ManIU M: m oM TEST 139* 4Wb Wacsa mac"o £L1K9A LEVEL SA'1 1~ SY UPAW UISG~~ 0 *LOCKED Clues Upstream of blockag*m 16 1.I QQQOO I E~CIOU Figure 2-10: FEBA Test Results AREVA NP Inc.
A AREVA ANP-3011Q1(NP)
Harris Nuclear Plant Unit 1 Revision 000 Realistic Large Break LOCA Analysis Page 25 2.7 NRC Q7.(Question 1 -Received 1/10/2012)
For droplet break up model, show the drop sizes produced by the model for several low reflood rate data. Present the clad, vapor temperature and total heat transfer coefficient versus time at the measured axial locations.
Show the heat transfer coefficient for all of the components comprising the dispersed flow film boiling (DFFB) heat transfer, including the interfacial heat transfer coefficient.
Response Figure 2-11 is a representation of droplet size predicted by the S-RELAP5 model. The basis for the figure are FLECHT-Seaset Test 61607 benchmarks.
Void fractions are included.
The data has been screened to illustrate data only for mist flow. Cladding temperature comparisons are included in response to NRC Question 4 (see Section 2.4, Figure 2-5 and Figure 2-6). Figure 2-12 through Figure 2-18 shows S-RELAP5 benchmark comparisons to FLECHT-SEASET Test 61607. Parameters plotted in these figures are offered in response to NRC verbal requests in the NRC-Progress Energy-AREVA meeting on 1/11/12.AREVA NP Inc.
A AR EVA Harris Nuclear Plant Unit 1 Realistic Large Break LOCA Analysis ANP-3011Q1(NP)
Revision 000 Page 26 Figure 2-11: Test 61607 Benchmark
-Droplet Diameter and Void Fraction at Peak Power Location AREVA NP Inc.
A AR EVA Harris Nuclear Plant Unit 1 Realistic Larme Break LOCA Analysis ANP-3011Q1(NP)
Revision 000 Paqe 27 Figure 2-12: FLECHT-SEASET Test 61607, Void Fraction Comparison, -4 ft (below blockage)Figure 2-13: FLECHT-SEASET Test 61607, Void Fraction Comparison, blockage)~5 ft (below AREVA NP Inc.
A AREVA Harris Nuclear Plant Unit 1 Realistic Larqe Break LOCA Analysis ANP-3011Q1(NP)
Revision 000 Pace 28 Figure 2-14: FLECHT-SEASET Test 61607, Vold Fraction Comparison, -6 ft (approx.blockage location)Figure 2-15: FLECHT-SEASET Test 61607, Vapor Temperature, -5 ft (below blockage)AREVA NP Inc.
A AR EVA Harris Nuclear Plant Unit 1 Realistic Large Break LOCA Analysis ANP-3011Q1(NP)
Revision 000 Page 29 Figure 2-16: FLECHT-SEASET Test 61607, Vapor Temperature, ~6.5 ft (above blockage)Figure 2-17: FLECHT-SEASET Test 61607, Vapor Temperature, -8 ft (above blockage)AREVA NP Inc.
A AR EVA Harris Nuclear Plant Unit 1 Realistic Large Break LOCA Analysis ANP-3011Q1(NP)
Revision 000 Page 30 Figure 2-18: FLECHT-SEASET Test 61607, Total Heat Transfer Coefficient, -6 ft AREVA NP Inc.
A AR EVA ANP-3011 Q1 (NP)Harris Nuclear Plant Unit 1 Revision 000 Realistic Large Break LOCA Analysis Page 31 2.8 NRC Q8.(Question 2 -Received 1/10/2012)
Since RELAP5 is one-dimensional the vapor temperature and droplets are distributed evenly across the hot channel. The code computed cross-section average quantities appears to fail to properly capture the very high temperature gradient in the vapor phase boundary layer near the wall so that the distribution of the evaporating water droplets play a fundamental role in the heat transfer process. In particular, interfacial heat transfer is over predicted.
This appears to be a major limitation for all one-dimensional codes. Test data shows that the channel is three-dimensional with accumulation of drops in the central region and a highly superheated region near the walls. Modeling this multi-dimensional behavior leads to a substantial reduction in the interfacial heat transfer and limiting of the droplet de-superheating to the central core and not the highly superheated layer near the walls.Explain what adjustments are made to the DFFB model components to overcome this major discrepancy.
That is, the sink temperature is not the average channel temperature for computing single phase heat transfer, an interfacial heat transfer between the drops and the vapor is control by the lower vapor temperature in the central core where the drops reside.Response: AREVA NP Inc.
A AR EVA Harris Nuclear Plant Unit 1 Realistic Large Break LOCA Analysis ANP-3011Q1(NP)
Revision 000 Page 32 AREVA NP Inc.
A AR EVA ANP-3011Q1(NP)
Harris Nuclear Plant Unit 1 Revision 000 Realistic Large Break LOCA Analysis Page 33 2.9 NRC Q9.(Question 3 -Received 1/10/2012)
Due to the simplified one-dimensional averaging of thermodynamic quantities in RELAP5 and the limited data, it is difficult to quantify all of the component contributions to DFFB.a. Address how is the magnitude of the droplet contribution verified in the RELAP5 model.b. Without detailed knowledge of the magnitude of all of the components to DFFB, validation of this model against reflood data may result in including other phenomena/effects that are not pertinent to the heat transfer benefits from the droplet break up model. Explain and justify the magnitude of the impact on DFFB heat transfer with this new model.c. Describe the interfacial heat transfer model and the impact on interfacial heat transfer coefficient with the new droplet model. In comparing the DFFB against data with the new droplet model, show all of the contributions to the total heat transfer coefficient versus time at the peak clad temperature (PCT) location.Response: Based on the NRC position, a response to this question is not required.AREVA NP Inc.
A AREVA ANP-3011Q1(NP)
Harris Nuclear Plant Unit 1 Revision 000 Realistic Large Break LOCA Analysis Page 34 2.10 NRC Q10.(Question 4 -Received 1/10/2012)
The packing fraction of 50 percent does not appear to capture all of the test data. Packing fraction as a function of burst strain varies in the range 52 to 80 percent based on data from Broughton, J. M, 1981, "[Power Burst Facility]
PBF [loss-of-coolant accident]
LOCA Test Series, Test LOC-3 and LOC-5 Fuel Behavior Report," NUREG/CR-2073.
The Nuclear Energy Agency (NEA) Organization for Economic and Co-operation and Development (OECD)Nuclear Fuel Behaviour in Loss-of-coolant Accident (LOCA) Conditions State-of-the-art Report identifies 55.5 and 61.5 percent fill fraction for the FR2 reactor test E2. Values for the high burnup fuel in IFA-650.4 are expected to be higher than 70 percent, consistent with the bounds for PBF/LOCA gamma scanning and micrographies and FR-2. (See Grandjean, C "IRSN Calculation of the IFA-650.4 and .5 LOCA Tests ISRN, Cadadache, Fr. EHPG Meeting, Storefiell, Mar 12-15, 2007 meeting).Response: The PBF results were discussed in the January 11 NRC-Progress Energy-AREVA meeting (see Figure 2-19). All low strain rupture rods were fresh rods. FR2 tests with fresh rods indicated very little fuel relocation.
Some observations on NUREG-2073 Reference
[20]: AREVA NP Inc.
A AR EVA Harris Nuclear Plant Unit 1 Realistic Large Break LOCA Analysis ANP-3011Q1(NP)
Revision 000 Page 35 o[I Figure 2-19: Test Packing Factor vs. Rupture Strain AREVA NP Inc.
A AREVA Harris Nuclear Plant Unit 1 Realistic Large Break LOCA Analysis ANP-3011Q1(NP)
Revision 000 Page 36 Table 2-6: Halden Test IFA 650.4 Post Irradiation Exam Data AREVA NP Inc.
A AREVA ANP-3011Q1(NP)
Harris Nuclear Plant Unit 1 Revision 000 Realistic Large Break LOCA Analysis Page 37 2.11 NRC Q11.(Question 5 -Received 1/10/2012)
Address whether the use of a nominal decay heat curve has ever been applied to decay heat test data over the range of applicability to show that this approach captures all decay heat conditions.
The discussion should also address the uncertainty in generating this nominal curve and demonstrate that use of the nominal curve does not capture the decay heat for the first two seconds. Provide a multiplier which appropriate captures the decay heat behavior during this first two seconds of the curve.Response AREVA NP Inc.
A AR EVA Harris Nuclear Plant Unit 1 Realistic Large Break LOCA Analysis ANP-3011Q1(NP)
Revision 000 Paqe 38 AREVA NP Inc.
A AREVA Harris Nuclear Plant Unit 1 Realistic Lare Break LOCA Analysis ANP-3011Q1(NP)
Revision 000 M-- M~PH 4 p'1 Pas.-4 4 4 L to 92 ti -Pis Wb-T 4-41-TaI-'4.-T-4t-TIA-m t Time I Figure 2-20: Example of Power History AREVA NP Inc.
A AREVA Harris Nuclear Plant Unit 1 Realistic Larqe Break LOCA Analysis ANP-3011Q1(NP)
Revision 000 Paae 40 AREVA NP Inc.
A AR EVA Harris Nuclear Plant Unit 1 Realistic Lame Break LOCA Analysis ANP-3011Q1(NP)
Revision 000 Paae 41 Figure 2-21: Decay Heat Curve, Burnup = 1 GWd/MTU AREVA NP Inc.
A AREVA Harris Nuclear Plant Unit 1 Realistic Lame Break LOCA Analysis ANP-3011Q1(NP)
Revision 000 Pane 42 2.12 Sleicher-Rouse Error Adjustment AREVA NP Inc.
A AREVA Harris Nuclear Plant Unit 1 Realistic Larqe Break LOCA Analysis ANP-3011Q1(NP)
Revision 000 Paae 43 3.0 HARRIS PCT
SUMMARY
Table 3-1 reports the RLBLOCA PCT rackup for HNP, beginning fuel cycle 18. The basis of the rackup is the limiting case from Reference
[2]. The maximum PCT assessment possible for the range of sensitivities examined for swell, rupture, and relocation (see Section 2.1) is applied as required by the NRC. Also included in the rackup is the AREVA PCT assessment for an error in the S-RELAP5 application of the Sleicher-Rouse correlation for heat transfer from the cladding surface to vapor in the coolant channel (see Section 2.12). [I* ,Table 3-1: HNP PCT Rackup AREVA NP Inc.
A AR EVA ANP-3011Q1(NP)
Harris Nuclear Plant Unit 1 Revision 000 Realistic Large Break LOCA Analysis Page 44
4.0 REFERENCES
- 1. EMF-2103(P)(A)
Revision 0, "Realistic Large Break LOCA Methodology," Framatome ANP, Inc.2. ANP-301 1(P) Revision 1, "Harris Nuclear Plant Unit 1 Realistic Large Break LOCA Analysis." 3. EMF-2100(P)
Revision 13, "S-RELAP5 Models and Correlations Code Manual." 4. BAW-1 01 66(P)(A) Revision 5, "BEACH -Best Estimate Analysis Core Heat Transfer." 5. M.J. Loftus et. al., "Analysis of FLECHT-SEASET 163-Rod Blocked Bundle Data Using COBRA-TF," NUREG/CR-41166, NRC/EPRI/Westinghouse Report No. 15, October 1985.6. L.H.J. Watchers, et. al., "The Heat Transfer from a Hot Wall to Impinging Water Drops in the Spheroidal State," Chem. Eng. Sci., 21, pp. 1047-1056, 1966.7. L.H. J. Watchers, et. al., "The Heat Transfer from a Hot Wall to Impinging Mist Drops in the Spheroidal State," Chem. Eng. Sci., 21, pp. 1231-1238, 1966.8. J. Sugimoto and Y. Murao, "Effect of Grid Spacers on Reflood Heat Transfer in PWR LOCA," J. of Nucl. Sci. Technol., pp. 102-114, February 1984.9. BAW-10227(P)(A), Revision 001, "Evaluation Of Advanced Cladding and Structural Material (M5) in PWR Reactor Fuel," AREVA.10. J. Chiou, et. al., "Spacer Grid Heat Transfer Effects during Reflood," NUREG/CR-0043, U.S. NRC 1982.11. C.K. Nithianandan, M.J. Lepcio, and J.R. Biller, "Reflood Heat Transfer During PWR-LOCA," paper # NHTC01 1777, presented at the 3 5 th National Heat Transfer Conference held in Anaheim, California on June 10-12, 2001.12. C.K. Nithianandan, J.A. Klingenfus, and S.S. Reilly, "RELAP5 Model to Simulate the Thermal-Hydraulic Effects of Grid Spacers and Cladding Rupture during Reflood," paper presented at the NURETH-7 conference held in Saratoga Springs, New York on September 10-15, 1995.13. C.K. Nithianandan, R.J. Lowe, and J.R. Biller, "Thermal-Hydraulic Effects of Grid Spacers and Cladding Rupture During Reflood," Eighth Proceedings of Nuclear Thermal Hydraulics, pp. 83-92, 1992 ANS Winter Meeting, November 15-20, 1992, Chicago, Illinois.14. C.K. Nithianandan, "RELAP5 Model to Simulate the Thermal-Hydraulic Effects of Clad Swelling and Rupture during Reflood," pp 341-349, Proceedings of the 5 th Int. Meeting of Reactor Thermal Hydraulics, NURETH-5, held in Salt Lake City, September 21-24, 1992.15. C.K. Nithianandan and J.R. Biller, "Thermal-Hydraulic Effects of Clad Swelling and Rupture During Reflood," Sixth Proceedings of Nuclear Thermal Hydraulics, pp 381-392, ANS Winter Meeting, November 11-16, 1990, Washington, DC.16. F.J. Erbacher, et.al. "Temperature and Quenching Behavior of Undeformed, Balooned and Burst Fuel Rods in a LOCA," Fifth International Meeting on Thermal Nuclear Reactor Safety September 9-13, 1984 Karlsruhe, FRG.17. F.J. Erbacher, et.al. "Cladding Deformation and Emergency Core Cooling of a Pressurized Water Reactor in a LOCA," KfK 4781, August 1990.AREVA NP Inc.
A AREVA ANP-3011QI(NP)
Harris Nuclear Plant Unit 1 Revision 000 Realistic Large Break LOCA Analysis Page 45 18. C. A. Sleicher and M. W. Rouse, "A Convenient Correlation for Heat Transfer to Constant and Variable Property Fluids in Turbulent Pipe Flow," International Journal of Heat and Mass Transfer, Volume 18, pp. 677-683, 1975.19. K. H. Sun, J. M. Gonzales Santalo, and C. L. Tien, "Calculations of Combined Radiation and Convection Heat Transfer in Rod Bundles Under Emergency Cooling Conditions," Journal of Heat Transfer, pp. 414 420, 1976.20. NUREG/CR-2073 "PBF LOCA Test Series, Tests LOC-3 and LOC-5 Fuel Behavior Report," June 1981.21. T.R. Yackle, et. al. "Evaluation of the Thermal-Hydraulic Response and Fuel Rod Thermal and Mechanical Deformation Behavior during the power burst facility test LOCA-3," ANS Thermal Reactor Safety Meeting, Knoxville, TN, April 8,1980.22. B.C. Oberlander, M. et. al. "LOCA IFA650.4:
Fuel Relocation Study," LOCA Workshop /HPG-Meeting, Prague, September 2007.23. ANSI/ANS-5.1-1979 American National Standard for Decay Heat Power in Light Water Reactors, American National Standards Institute, Inc., August 29, 1979.24. Regulatory Guide 1.157, "Best Estimate Calculations of Emergency Core Cooling System Performance", May 1989.AREVA NP Inc.
A AR EVA Harris Nuclear Plant Unit 1 Realistic Large Break LOCA Analysis ANP-3011Q1(NP)
Revision 000 Page A-1 APPENDIX A: EMF 2100 SECTIONS 3 AND 4 A.1 Droplet Shattering in Mist Flow Text and equations in all sections of this attachment have been transcribed and/or paraphrased from the S-RELAP5 Models and Correlations Code Manual, EMF-2100 (P) Revision 13, Sections 3 and 4.A.1.1 Dispersed Flow AREVA NP Inc.
A AR EVA Harris Nuclear Plant Unit 1-.R.Jealistic Large Break LOCA Analysis ANP-3011QI(NP)
Revision 000 Page A-2 _AREVA NP Inc.
A AR EVA Harris Nuclear Plant Unit 1 Realistic Larme Break LOCA Analysis ANP-301 1QI(NP)Revision 000 PaQe A-3 AREVA NP Inc.
A AREVA Harris Nuclear Plant Unit 1 Realistic Large Break LOCA Analysis ANP-3011Q1(NP)
Revision 000 Page A-4 AREVA NP Inc.
A AREVA Harris Nuclear Plant Unit 1 Realistic Large Break LOCA Analysis ANP-3011Q1 (NP)Revision 000 Page A-5 AREVA NP Inc.
A AREVA Harris Nuclear Plant Unit 1 Realistic Large Break LOCA Analysis ANP-3011Q1(NP)
Revision 000 Paqe A-6 AREVA NP Inc.
A AREVA Harris Nuclear Plant Unit 1 Realistic Large Break LOCA Analysis ANP-3011Q1(NP)
Revision 000 Page A-7 AREVA NP Inc.
A AR EVA ANP-3011QI(NP)
Harris Nuclear Plant Unit 1 Revision 000 Realistic Large Break LOCA Analysis Page A-8 A.1.4 EMF-2100(P)
R13 -Section 3 -References A.1.1 The RELAP5 Development Team of INEL, RELAP5/MOD3 Code Manual Volume 1: Code Structure, System Models, and Solution Methods, NUREG/CR-5535, INEL-95/0174, August, 1995.A.1.2 B. Chexal and G. Lellouche, A Full-Range Drift-Flux Correlation for Vertical Flows (Revision 1), EPRI NP-3989-SR, September 1986; Void Fraction Technology for Design and Analysis, TR-106326, March 1997.A.1.3 EMF-2102(P)
Revision 0, S-RELAP5:
Code Verification and Validation, Sections 2.1 and 2.6, Framatome ANP, Inc., August 2001.A.1.4 EMF-21 00(P) Revision 3, S-RELAP5 Models and Correlations Code Manual, Siemens Power Corporation, January 2001.A.1.5 Y. Taitel and A. E. Dukler, "A Model for Predicting Flow Regime Transitions in Horizontal and Near Horizontal Gas-Liquid Flow," American Institute of Chemical Engineering Journal Volume 22, pp. 47-55, 1976.A.1.6 Y. Taitel, D. Bornea and A. E. Dukler, "Modeling Flow Pattern Transitions for Steady Upward Gas-Liquid Flow in Vertical Tubes," American Institute of Chemical Engineering Joumal Volume 26, pp. 345-354, 1980.A.1.7 G. B. Wallis, One Dimensional Two Phase Flow, McGraw-Hill, New York, 1969.A.1.8 M. Ishii and G. DeJarlais, "Inverted Annular Flow Modeling," presented at EG&G, July 27, 1982.A.1.9 G. DeJarlais and M. Ishii, Inverted Annular Flow Experimental Study, Argonne National Laboratory Report ANL-85-31, NUREG/CR-4277, 1985.A.1.10 M. Ishii and K. Mishima, Study of Two-Fluid Model and Interfacial Area, Argonne National Laboratory Report ANL-80-1 11, NUREG/CR-1 873, 1980.A.1.11 S. Z. Rouhani and M. S. Sohal, "Two-Phase Flow Patterns:
A Review of Research Results," Progress in Nuclear Energy Volume 11, pp.219-259, 1983.A.1.12K. Mishima and M. Ishii, "Flow Regime Transition Criteria for Upward Two-Phase Flow in Vertical Tubes," International Journal Heat and Mass Transfer Volume 27, pp.723-737, 1984.A.1.13 N. T. Obot and M. Ishii, "Two-Phase Flow Regime Transition Criteria in Post-Dryout Region Based on Flow Visualization Experiments," International Journal Heat and Mass Transfer Volume 31, pp. 2559-2570, 1988.A.1.140. C. Jones, Jr. and N. Zuber, "The Interrelation Between Void Fraction Fluctuations and Flow Patterns in Two-Phase Flow," International Journal of Multiphase Flow Volume 2, pp. 273-306, 1975.AREVA NP Inc.
A A R EVA ANP-301 1Q1 (NP)Harris Nuclear Plant Unit 1 Revision 000 Realistic Large Break LOCA Analysis Page A-9 A.1.15M. A. Vince and R. T. Lahey, Jr., "On the Development of An Objective Flow Regime Indicator," International Journal of Multiphase Flow, Volume 8, pp.93-124, 1982.A. 1.16 N. A. Radovcich and R. Moissis, The Transition from Two-Phase Bubbly Flow to Slug Flow, MIT Report 7-7673-22, June 1962.A.1.17 P. Griffith and G. A. Snyder, The Bubbly-Slug Transition in a High Velocity Two-Phase Flow, MIT Report 5003-29, TID20947, July 1964.A.1.18D. R. Liles et al., TRAC-PF1/MODI Correlations and Models, NUREG/CR-5069, LA-11208-MS, December 1988.A.1.19V. H. Ransom et al., RELAP5/MOD1 Code Manual Volume 1: System Models and Numerical Methods, NUREG/CR-1826, EGG-2070, March 1982.A.1.20M. Ishii and G. DeJarlais, "Flow Regime Transition and Interfacial Characteristics of Inverted Annular Flow," Nuclear Engineering and Design, Volume 95, pp. 171-184, 1986.A.1.21 D. D. Taylor et al., TRAC-BDI/MODI:
An Advanced Best Estimate Computer Program for Boiling Water Reactor Transient Analysis, Volume 1: Model Description, NUREG/CR-3633, EGG-2294, April 1984.A.1.22V. H. Ransom et al., RELAP5/MOD2 Code Manual, Volume 1: Code Structure, Systems Models, and Solution Methods, NUREG/CR-4312, EGG-2396, Revision 1, March 1987.A.1.23Y. Kukita, Y. Anoda, H. Nakamura and K. Tasaka, "Assessment and Improvement of RELAP5/MOD2 Code's Interphase Drag Models," 24th ASME/AIChE National Heat Transfer Conference, Pittsburgh, PA, August 9-12, 1987.A.1.24 EMF-92-139(P), Volume 2, Revision 1, S-RELAP5:
Realistic LOCA ECCS Evaluation Model for PWR Large Break LOCA Analysis, Siemens Nuclear Power Corporation, October 1994.A.1.25A.H.
Shapiro and A. J. Erickson, Transactions of ASME, Volume 79, p. 775, 1957.A.1.26M. Ishii and T. C. Chawla, Local Drag Laws in Dispersed Two-Phase Flow, Argonne National Laboratory Report ANL79-105, NUREG/CR-1230, December 1979.A.1.27S. M. Nujhawan, "Experimental Investigation of Thermal Non-Equilibrium in Post-Dryout Steam-Water Flow," Ph.D. Thesis, Lehigh University, May 1980.A.1.28 FLECHT SEASET Program, PWR FLECHT SEASET Unblocked Bundle, Forced and Gravity Reflood Task Data Evaluation and Analysis Report, NUREG/CR-2256, EPRI NP-2013, WCAP-8991, November 1981.A.1.29 1. Kataoka and M. Ishii, "Drift Flux Model for Large Diameter Pipe and New Correlation for Pool Void Fraction," International Journal Heat and Mass Transfer Volume 30, pp.1927-1939, 1987.A.1.30T. M. Anklam, R. J. Miller and M. D. White, Experimental Investigations of Uncovered-Bundle Heat Transfer and Two-Phase Mixture-Level Swell Under High-Pressure Low Heat-Flux Conditions, NUREG/CR-2456, ORNL-5848, March 1982.AREVA NP Inc.
A AREVA ANP-30111 Q(NP)Harris Nuclear Plant Unit 1 Revision 000 Realistic Large Break LOCA Analysis Page A-10 A.1.31 XN-NF-82-49 (P) (A), Exxon Nuclear Company Evaluation Model EXEM PWR Small Break Model, Revision 1, Exxon Nuclear Company, April 1989.A. 1.321. Kataoka and M. Ishii, Mechanism and Correlation of Droplet Entrainment and Deposition in Annular Two-Phase Flow, Argonne National Laboratory Report ANL 82-44, NUREG/CR-2885, 1982.A.1.332D/3D Program Upper Plenum Test Facility, Test No. 11 Countercurrent Flow in PWR Hot Leg Test, Quick Look Report, R515/87/08, Siemens KWU, March 1987.A.1.34 U9 316/89/2, UPTF Test No .6 Downcomer Countercurrent Flow Test, Quick Look Report, 2D/3D Program Upper Plenum Test Facility, Siemens KWU March 1989;E314/90/003, UPTF Test No. 7 Downcomer Countercurrent Flow Test, Quick Look Report, Siemens KWU, March 1990.A. 1.35 E314/90/19, UPTF Test No. 29 Entrainment/De-entrainment Test Quick Look Report, 2D/3D Program Upper Plenum Test Facility, November, 1990; U9 316/88/3, UPTF Test No. 10 Tie Plate Countercurrent Flow Test, March 1988.A.1.36A. Ohnuki, H. Adachi, and Y. Murao, "Scale Effects on Countercurrent Gas-Liquid Flow in Horizontal Tube Connected to Inclined Riser," Nuclear Engineering and Design, Volume 107, pp. 283-294, 1988.A. 1.37J. A. Findlay, BWR Refill-Reflood Program Task 4.8 -Model Qualification Task Plan, NUREG/CR-1899, EPRI NP-1527, GEAP-24898, August 1981.A.1.38V. H. Ransom et al., RELAP5/MOD2 Code Manual, Volume 3: Developmental Assessment Problems, EGG-TFM-7952, December 1987.A.1.39R. A. Riemke, H. Chow and V. H. Ransom, RELAP5/MODI Code Manual Volume 3: Checkout Problems Summary, Interim Report, EGG-NSMD-6182, February 1983.A.1.40H. R. Saedi and P. Griffith, "The Pressure Response of a PWR Pressurizer During an Insurge Transient," Transactions of ANS, Volume 44, pp. 606-607, 1983.A.1.41 D. Bharathan, H. T. Richter and G. B. Wallis, Air-Water Counter-Current Annular Flow in Vertical Tubes, EPRI NP-786, 1978.A. 1.42 EMF-2102(P)
Revision 0, S-RELAP5:
Code Verification and Validation, Framatome ANP, Inc., August 2001.A.1.43 U. S. Rohatgi, L. Y. Neymotin, J. Jo, and W. Wulff, Bias in Peak Clad Temperature Predictions Due to Uncertainties in Modeling of ECC Bypass and Dissolved Non-Condensible Gas Phenomena, NUREG/CR-5254, BNL-NUREG-52168, September 1990.A.1.44T. Iguchi, et al., Data Report on Large Scale Reflood Test-43 -CCTF Core Shakedown Test C2-SH2 (Run 54), JAERI-memo 58-155, Japan Atomic Energy Research Institute, May 1983.A. 1.45 V. H. Ransom, Course A -- Numerical Modeling of Two-Phase Flows for Presentation at Ecole d'Ete d'Analyse Numerique, EGG-EAST-8546, May 1989.AREVA NP Inc.
A AREVA ANP-3011QI(NP)
Harris Nuclear Plant Unit 1 Revision 000 Realistic Large Break LOCA Analysis Page A-11 A.1.46J. M. Putney, Development of A New Bubbly-Slug Interfacial Friction Model for RELAP5-Final Report, CEGB Draft in Confidence, PWRIHTWG/P(89)722, June 1989.A.1.47D. A. Drew, L. Y. Cheng, and R. T. Lahey, Jr., "The Analysis of Virtual Mass Effects in Two-Phase Flow," International Journal of Multiphase Flow, Volume 5, pp. 233-242, 1979.A.1.48C. G. Richards and I. Stopher, "Stratified Flow Benchmark Calculations Using RELAP5/MOD2," presented at 1990 RELAP51TRAC-BWR International Users Seminar, Chicago, September 1990.A. 1.49M. J. Thurgood et al., COBRA/TRAC-A Thermal-Hydraulics Code for Transient Analysis of Nuclear Reactor Vessels and Primary Coolant Systems, Volume 1, Equations and Constitutive Models, NUREG/CR-3046 PNL-4385, March 1983.A.1.50 M. S. Plesset and S. A. Zwick, "Growth of Vapor Bubbles in Superheated Liquids," Journal of Applied Physics, Volume 25, pp. 493-500, 1954.A.1.51 R. A. Dimenna et al., RELAP5/MOD2 Models and Correlations, NUREG/CR-5194, EGG-2531, August, 1988.A.1.52 K. Lee and D. J. Ryley, "The Evaporation of Water Droplets in Superheated Steam," Journal of Heat Transfer, pp. 445-456, 1968.A.1.53H. C. Unal, "Maximum Bubble Diameter, Maximum Bubble-Growth Time and Bubble-Growth Rate During the Subcooled Nucleate Flow Boiling of Water up to 17.7 MN/m 2 ," International Journal of Heat and Mass Transfer, Volume 19, pp. 643-649, 1976.A.1.54 R. T. Lahey, Jr., "A Mechanistic Subcooled Boiling Model," Proceedings of 6 th International Heat Transfer Conference, Volume 1, pp. 293-297, 1978.A.1.55Y. M. Chen and F. Mayinger, "Measurement of Heat Transfer at the Phase Interface of Condensing Bubbles," in ANS Proceedings 1989 National Heat Transfer Conference (Philadelphia, PA, August 6-9, 1989) HTC-Volume 4, pp. 147-152.A.1.56D. Moalem and S. Sideman, "The Effect of Motion on Bubble Collapse," International Journal of Heat and Mass Transfer Volume 16, pp. 2321-2329, 1973.A.1.57LA-12031-M, TRAC-PF1/MOD2 Code Manual: Volume 3 Programmer's Guide, NUREG/CR-5673, Los Alamos National Laboratory, July 1992; J. W. Spore et al., TRAC-PF1/MOD2 Volume I. Theory Manual, (Draft), LA-12031-M, NUREG/CR-5673, July 1993.A.1.58 S. G. Bankoff, "Some Condensation Studies Pertinent to Light Water Safety," International Journal of Multiphase Flow Volume 6, pp. 51-67, 1980.A.1.59L. K. Brumfield, R. N. Houze and T. G. Theofanous, "Turbulent Mass Transfer at Free, Gas-Liquid Interfaces, with Applications to Film Flows," International Journal of Heat and Mass Transfer, Volume 18, pp. 1077-1081, 1975.A.1.60Y. A. Hassan, "Dispersed-Flow Heat Transfer During Reflood in a Pressurized Water Reactor After a Large-Break Loss-of-Coolant Accident," Transactions of ANS, Volume 53, pp. 326-328, 1986.AREVA NP Inc.
A AREVA ANP-3011Q1(NP)
Harris Nuclear Plant Unit I Revision 000 Realistic Large Break LOCA Analysis Page A-12 A.1.61 M. C. Yuen and L. W. Chen, "Heat-Transfer Measurements of Evaporating Liquid Droplets," International Journal of Heat and Mass Transfer, Volume 21, pp. 537-542, 1978.A.1.62Y. A. Hassan, "Predictions of Vapor Superheat in Rod Bundle Geometry with Modified RELAP5/MOD2," Transactions of ANS, Volume 54, pp. 212-214, 1987.A.1.63G. Th. Analytis, "Developmental Assessment of TRAC-BF1 with Separate-Effect and Integral Reflooding Experiments," Proceedings of the 5 th International Topical Meeting on Reactor Thermal Hydraulics, Volume I, pp. 287-291, September 1992; also presentation given at 4th ICAP Management Meeting, October 18-20, Bethesda, 1989.A.1.64NUREG/CR-4356, EGG-2626, TRAC-BF1/MOD1:
An Advanced Best-Estimate Computer Program for BWR Accident Analysis, Idaho National Engineering Laboratory, August 1992.A. 1.65 G. Brown, "Heat Transmission of Condensation of Steam on a Spray of Water Drops," Proceedings of the General Discussion on Heat Transfer, pp. 49-52, 1951.A.1.66J. G. M. Andersen, REMI/HEA T COOL, A Model for Evaluation of Core Heat-Up and Emergency Core Spray Cooling System Performance for Light-Water-Cooled Nuclear Power Reactors, Riso Report No. 296, Danish Atomic Energy Commission, September 1973.A.1.67T. G. Theofanous, "Modeling of Basic Condensation Processes," The Water Reactor Safety Research Workshop on Condensation, Silver Springs, MD. May 24-25, 1979.A.1.68J. H. Linehan, M. Petrick, and M. M. EI-Wakil, "The Condensation of Saturated Vapor on a Subcooled Film During Stratified Flow," Chemical Engineering Symposium Series 66 (102), pp. 11-20, 1972.A.1.69ANCR-NUREG-1 335, RELAP4/MOD5 A Computer Program for Transient Thermal-Hydraulic Analysis of Nuclear Reactors and Related Systems, User's Manual Volume I RELAP4/MOD5 Description, Idaho National Engineering Laboratory, September 1976.A.1.70G. G. Sklover and M. D. Rodivilin, "Condensation on Water Jets with a Cross Flow of Steam," Teploenergetika Volume 23, 48-51, 1976.A.1.71ANF-913(P)(A)
Vollume 1 Revision 1 and Volume 1 Supplements 2, 3, and4, COTRANSA2:
A Computer Program for Boiling Water Reactor Transient Analyses, Advanced Nuclear Fuels Corporation, August 1990.A.1.72 EMF-92-139(P), Volume 3, Supplement 3, Realistic LOCA ECCS Evaluation Model Assessment for PWR Large Break Analysis Assessment for LOFT Test L2-5, Siemens Power Corporation, June 1993.A.1.73 EMF-92-139(P), Volume 3, Supplement 4, Realistic LOCA ECCS Evaluation Model Assessment for PWR Large Break Analysis Assessment for LOFT Test L2-6, Siemens Power Corporation, June 1993.A. 1.74 U9 316/87/17, U PTF Test No. 5 Downcomer Separate Effect Test, Quick Look Report, 2D/3D Program Upper Plenum Test Facility, Siemens KWU, October 1987; Exploratory assessment was made with ANF-RELAP and no report was written.AREVA NP Inc.
A AR EVA ANP-3011Q1(NP)
Harris Nuclear Plant Unit I Revision 000 Realistic Large Break LOCA Analysis Page A-13 A.1.75 EMF-92-139(P), Volume 3, Supplement 2, Realistic LOCA ECCS Evaluation Model Assessment for PWR Large Break Analysis Assessment for CCTF Test 54, Siemens Power Corporation, April 1993.A.1.76 C. F. Colebrook, "Turbulent Flow in Pipes with Particular Reference to the Transition Region Between Smooth and Rough Pipe Laws," Journal of the Institution of Civil Engineers, Volume 11, pp. 133-156, 1939.A.1.77A. K. Jain, "Accurate Explicit Equation for Friction Factor," ASCE J. Hydraulics Division, Volume 102, pp. 674-677, 1976.A.1.78K. T. Chaxton, J. G. Collier and J. A. Ward, H.T.F.S. Correlation for Two-Phase Pressure Drop and Void Fraction in Tubes, AERE-R7162, 1972.A.1.79 D. Chisholm, "A Theoretical Basis for the Lockhart-Martinelli Correlation for Two-Phase Flow," Journal of Heat Transfer, Volume 10, pp. 1767-1778, 1967.A.1.80 R. W. Lockhart and R. C. Martinelli, "Proposed Correlation of Data for Isothermal Two-Phase, Two-Component Flow in Pipes," Chemical Engineering Progress, Volume 45, pp. 39-48, 1949.A.1.81 R. A. Dimenna and D. L. Caraher, "Air-Water Hydraulics Modeling for a Mark-22 Fuel Assembly with RELAP5(U)," presented at 1990 RELAP5/TRAC-BWR International Users Seminar, Chicago, September 1990.A.1.82 D. J. Zigrang and N. D. Sylvester, "A Review of Explicit Friction Factor Equations," Transactions of ASME, Journal of Energy Resources Technology, Volume 107, pp. 280-283, 1985.A.1.83NEDE-30914P, ATLAS03 DAS, Hardware and software, Description and user's guide, GE, May 1985.A.1.84 NT34/2000/e068, Multifunction Karlstein thermal hydraulic test loop (KATHY) general description of test loop, Framatome ANP GmbH, January 2001.A.1.85 E6080-908-1, Revision 1, ATRIUM I0A pressure drop and critical power evaluation, Siemens Power Corporation, December 1996.A.1.86XN-NF-79-59(P)(A), Methodology for calculations of pressure drop in BWR fuel assemblies, Exxon nuclear company, Inc., October 1983.A.1.87NEDE-31023, Rod bundle pressure drop correlations for advanced fuel designs, GE, February 1985.A.1.88THRP-VO:
Thermal Hydraulic Core Analysis Code Description and User's Manual, Siemens Technical Report, E548, 1991.A.1.89L. F. Moody, "An Approximate Formula for Pipe Friction Factors," Trans. ASME, P.1005, Vol.69, 1947.A.1.902A4-RELAP5_MOD2 Users Manual, Volume 1, Revision AA, (see document number"2A4.26-2A4-RELAP5_MOD2_CompleteUserManual-AA" in Documentum) 2007.AREVA NP Inc.
A AREVA Harris Nuclear Plant Unit 1 Realistic Large Break LOCA Analysis ANP-3011Q1(NP)
Revision 000 Pale A-14 AREVA NP Inc.
A AREVA Harris Nuclear Plant Unit 1 Realistic Larae Break LOCA Analvsis ANP-3011Q1(NP)
Revision 000 Paae A-15 AREVA NP Inc.
A AR EVA Harris Nuclear Plant Unit 1 Realistic Large Break LOCA Analysis ANP-3011Q1(NP)
Revision 000 Page A-16 AREVA NP Inc.
A AREVA Harris Nuclear Plant Unit 1 Realistic Large Break LOCA Analysis ANP-3011Q1(NP)
Revision 000 Page A-17 AREVA NP Inc.
A AREVA Harris Nuclear Plant Unit 1 Realistic Large Break LOCA Analysis ANP-3011Q1(NP)
Revision 000 Paqe A-18 AREVA NP Inc.
A AR EVA Harris Nuclear Plant Unit 1 Realistic Large Break LOCA Analysis ANP-3011QI(NP)
Revision 000 Page A-19 AREVA NP Inc.
A AR EVA ANP-3011Q1(NP)
Harris Nuclear Plant Unit 1 Revision 000 Realistic Large Break LOCA Analysis Page A-20 A.2.4 EMF-2100(P)
R13 -Section 4- References A.2.1 V. H. Ransom et al., RELAP5/MOD2 Code Manual, Volume 1: Code Structure, Systems Models, and Solution Methods, NUREG/CR-4312, EGG-2396, Revision 1, March 1987.A.2.2 F. W. Dittus and L. M. K. Boelter, "Heat Transfer in Automobile Radiators of the Tubular Type," Publications in Engineering, Volume 2, pp. 443-461, University of California, Berkeley, 1930.A.2.3 W. H. McAdams, Heat Transmission, 3 rd edition, McGraw-Hill, New York, 1954.A.2.4 M. J. Thurgood et al., COBRA/TRA C-A Thermal-Hydraulics Code for Transient Analysis of Nuclear Reactor Vessels and Primary Coolant Systems, Volume 1, Equations and Constitutive Models, NUREG/CR-3046 PNL-4385, March 1983.A.2.5 E. M. Sparrow, A. L. Loeffler and H. A. Hubbard, "Heat Transfer to Longitudinal Laminar Flow Between Cylinders," Journal of Heat Transfer Volume 83, 415, 1961.A.2.6 J. P. Holman, Heat Transfer, 5 th edition, McGraw-Hill, New York, 1981.A.2.7 XN-NF-82-49 (P) (A), Revision 1, Exxon Nuclear Company Evaluation Model EXEM PWR Small Break Model, Exxon Nuclear Company, April 1989.A.2.8 D. R. Liles et al., TRAC-PF1/MOD1 Correlations and Models, NUREG/CR-5069, LA-11208-MS, December 1988.A.2.9 A. E. Bergles and W. M. Rohsenow, "The Determination of Forced-Convection Surface-Boiling Heat Transfer," Journal of Heat Transfer, Transactions of ASME, Volume 86, p. 365, 1964.A.2.10 P. Saha and N. Zuber, "Point of Net Vapor Generation and Vapor Void Fraction in Subcooled Boiling," Proceedings of 5 th International Heat Transfer Conference, Volume IV, pp.175-179, 1974.A.2.11 J. C. Chen, "A Correlation for Boiling Heat Transfer to Saturated Fluids in Convective Flow," Process Design and Development, Volume 5, pp. 322-327, 1966.A.2.12H. K. Forster and N. Zuber, "Bubble Dynamics and Boiling Heat Transfer," American Institute of Chemical Engineering Journal, Volume 1, pp. 532-535, 1955.A.2.13 R. Viskanta and A. K. Mohanty, TMI-2 Accident:
Postulated Heat Transfer Mechanisms and Available Data Base, NUREG/CR-2121, ANL-81-26, April 1981.A.2.14F. D. Moles and J. R. G. Shaw, "Boiling Heat Transfer to Subcooled Liquids Under Conditions of Forced Convection," Transactions Institution of Chemical Engineers, Volume 50, pp. 76-84, 1972.A.2.15 R. T. Lahey, Jr., "A Mechanistic Subcooled Boiling Model," Proceedings of 6/h International Heat Transfer Conference, Volume 1, pp. 293-297, 1978.AREVA NP Inc.
A AR EVA ANP-3011Q1(NP)
Harris Nuclear Plant Unit 1 Revision 000 Realistic Large Break LOCA Analysis Page A-21 A.2.16Z. Rouhani, Void Measurements in the Region of Subcooled and Low Quality Boiling, Part 2, AE-239 (1966).A.2.17Biasi et. al., "Studies on Burnout Part 3 -A New Correlation for Round Ducts and Uniform Heating and Its Comparison with World Data," Energia Nucleare, Volume 14, pp. 530-536, 1967.A.2.18P. Griffith, J. F. Pearson, and R. J. Lepkowski, "Critical Heat Flux During a Loss-of Coolant Accident," Nuclear Safety Volume 18, pp. 298-309, 1977.A.2.19J. G. Collier, Convective Boiling and Condensation, 2 nd edition, McGraw-Hill, New York, 1981.A.2.20 N. Zuber, M. Tribus and J. W. Westwater, "Hydrodynamic Crisis in Pool Boiling of Saturated and Subcooled Liquid," 2nd International Heat Transfer Conference, Denver, Colorado, 1961.A.2.21 R. A. Dimenna et al., RELAP5/MOD2 Models and Correlations, NUREG/CR-5194, EGG-2531, August, 1988.A.2.22 EMF-2209(P)(A), SPCB Critical Power Correlation, Revision 2, September 2003.A.2.23L. S. Tong, Boiling Heat Transfer and Two-Phase Flow, P.144, John Wiley & Sons, 1965.A.2.24 C. A. Sleicher and M. W. Rouse, "A Convenient Correlation for Heat Transfer to Constant and Variable Property Fluids in Turbulent Pipe Flow," International Journal of Heat and Mass Transfer, Volume 18, pp. 677-683, 1975.A.2.25J. C. Chen, R. K. Sundaram, F. T. Ozkaynak, A Phenomenological Correlation for Post-CHF Heat Transfer, NUREG-0237, June 1977.A.2.26The RELAP5 Development Team, RELAP5/MOD3 Code Manual Volume 4: Models and Correlations, NUREG/CR-5535, INEL-95/0174, Idaho National Engineering Laboratory, August, 1995, see footnote on Page 4-80.A.2.27V. H. Sanchez-Espinoza, E. Elias, C. Homann, W. Hering and D. Struwe, Development and Validation of a Transition Boiling Model for the RELAP5/MOD3 Reflood Simulation, FZKA 5954, Forschungszentrum Karlsruhe, September 1997.A.2.28R. P. Forslund and W. M. Rohsenow, "Dispersed Flow Film Boiling," Journal of Heat Transfer Volume 90 (6), pp. 399-407, 1968.A.2.29 L. A. Bromley, "Heat Transfer in Stable Film Boiling," Chemical Engineering Progress Volume 46, pp. 221-227, 1950.A.2.30 P. J. Berenson, "Film Boiling Heat Transfer from a Horizontal Surface," Journal of Heat Transfer, pp. 351-358, 1961.A.2.31 Y. Y. Hsu and R. W. Graham, Transport Processes in Boiling and Two-Phase Systems, McGraw-Hill, New York, 1976.A.2.32 R. W. Shumway, TRAC-PF1 Post-CHF Heat Transfer Coefficient Variation for CSAU, Appendix 0 to Quantifying Reactor Safety Margins: Application of Code Scaling, AREVA NP Inc.
A A R EVA ANP-3011Q1(NP)
Harris Nuclear Plant Unit 1 Revision 000 Realistic Large Break LOCA Analysis Page A-22 Applicability, and Uncertainty Evaluation Methodology to a Large-Break, Loss-of-Coolant Accident, NUREG/CR-5249, EGG-2552, December 1989.A.2.33 G. Th. Analytis, M. Richner and S. N. Aksan, Assessment of Interfacial Shear and Wall Heat Transfer of RELAP5/MOD2/36.02 During Reflooding, EIR-Bericht Nr. 624, May 1987.A.2.34Y. Murao and J. Sugimoto, "Correlation of Heat Transfer Coefficient for Saturated Film Boiling During Reflood Phase Prior to Quenching," Journal of Nuclear Science and Technology, Volume 18[4], pp. 275-284, 1981.A.2.35 M. S. Dougall and W. M. Rohsenow, Film Boiling on the Inside of Vertical Tubes with Upward Flow of a Fluid at Low Qualities, MIT-ME 9079-26, 1963.A.2.36Wong, S. and Hochreiter, L. E., "Analysis of the FLECHT-SEASET Unblocked Bundle Steam Cooling and Boil-off Tests", January 1981, NUREG/CR-1533.
A.2.37 Drucker, M., Dhir, V. K., "Studies of Single and Two Phase Heat Transfer in a Blocked Four Rod Bundle," EPRI-NP 3485, Electric Power Research Institute (1984).A.2.38S. C. Yao, L. E. Hochreiter and W. J. Leech, Heat Transfer Augmentation in Rod Bundles Near Grid Spacers, J. Heat Transfer, 104, pp. 76-81, February 1982.A.2.39M. J. Meholic, L. E. Hochreiter, J. H. Mahaffy, J. Spring, "Increased Convective Heat Transfer Caused by Spacer Grids in Laminar High Void Fraction Flows," 2008 ANS Winter Meeting, Reno.A.2.40K. H. Sun, J. M. Gonzales-Santalo, and C. L. Tien, "Calculations of Combined Radiation and Convection Heat Transfer in Rod Bundles Under Emergency Cooling Conditions," Journal of Heat Transfer, pp. 414-420, 1976.A.2.41 FLECHT SEASET Program, PWR FLECHT SEASET Unblocked Bundle, Forced and Gravity Reflood Task Data Evaluation and Analysis Report, NUREG/CR-2256, EPRI NP-2013, WCAP-8991, November 1981.A.2.42 D. D. Taylor et al., TRAC-BD1/MDDI:
An Advanced Best Estimate Computer Program for Boiling Water Reactor Transient Analysis, Volume 1: Model Description, NUREG/CR-3633, EGG-2294, April 1984.A.2.43 E. F. Carpenter and A. P. Colburn, "The Effect of Vapor Velocity on Condensation Inside Tubes," Proceedings of General Discussion on Heat Transfer, Institute Mechanical Engineering/American Society of Mechanical Engineers, pp. 20-26, 1951.A.2.44RELAP5/MOD3.3 Code Manual Volume 1: Code Structure, System Models, and Solution Methods, NUREG/CR-5535/Rev 1-Vol 1, December, 2001.A.2.45 L. Arrieta and G. Yadigaroglu, Analytic Model for Bottom Reflooding and Heat Transfer in Light Water Reactors, EPRI Report NP-756, 1978.A.2.46 D. C. Groeneveld and J. C. Stewart, "The Minimum Film Boiling Temperature for Water During Film Boiling Collapse," Proceedings of the 7 th International Heat Transfer Conference (Munich, FRG, 1982). Volume 4, pp. 393-398, 1982.AREVA NP Inc.
A AREVA ANP-3011Q1(NP)
Harris Nuclear Plant Unit 1 Revision 000 Realistic Large Break LOCA Analysis Page A-23 A.2.47 U. S. Nuclear Regulatory Commission, Compendium of ECCS Research for Realistic LOCA Analysis, (Draft Report for Comment), NUREG-1230, April 1987.A.2.48 M. Sencar and N. Aksan, "Preliminary Results on the Assessment of RELAP5/MOD3 Code Reflooding Models Using PSI-NEPTUN and Lehigh University Bundle Reflooding Tests," Presentation at the 7th International Code Assessment and Application Program (ICAP) Management Meeting, Idaho Falls, ID, October 1991.A.2.49D. C. Groeneveld, S. C. Cheng, and T. Doan, "1986 AECL-UO Critical Heat Flux Lookup Table," Heat Transfer Engineering, Volume 7, pp. 46-62, 1986.A.2.50A. W. Bennett, G. F. Hewitt, H. A. Kearsey and R. K. F. Keeys, Heat Transfer to Steam-Water Mixtures Flowing in Uniformly Heated Tubes in Which the Critical Heat Flux Has Been Exceeded, UKAEA Research Group Report, AERE-R 5373, October 1967.A.2.51 EMF-2102(P)
Revision 0, S-RELAP5:
Code Verification and Validation, Framatome ANP, Inc., August 2001.A.2.52 EMF-92-139(P), Volume 3, Supplement 3, Realistic LOCA ECCS Evaluation Model Assessment for PWR Large Break Analysis Assessment for LOFT Test L2-5, Siemens Power Corporation, June 1993.A.2.53 EMF-92-139(P), Volume 3, Supplement 4, Realistic LOCA ECCS Evaluation Model Assessment for PWR Large Break Analysis Assessment for LOFT Test L2-6, Siemens Power Corporation, June 1993.A.2.54 EMF-92-139(P), Volume 3, Supplement 6, Realistic LOCA ECCS Evaluation Model Assessment for PWR Large Break Analysis Westinghouse 3-Loop PWR Sample Problem, Siemens Power Corporation, May 1993.A.2.55 EMF-92-139(P), Volume 3, Supplement 7, Realistic LOCA ECCS Evaluation Model Assessment for PWR Large Break Analysis Westinghouse 4-Loop PWR Example Problem, Siemens Power Corporation, June 1993.A.2.56J. C. M. Leung, "Transient Critical Heat Flux and Blowdown Heat Transfer Studies," Ph.D. dissertation, Northwestern University, June 1980.A.2.57S. Wang, Y. K. Kao, and J. Weisman "Studies of Transition Boiling Heat Transfer with Saturated Water at 1-4 Bar," Nuclear Engineering Design, Volume 70, pp.223-243, 1982.A.2.58J. Zhang, S. M. Bajorek, R. M. Kemper, and L. E. Hochreiter, "WCOBRA/TRAC Analysis of ORNL High Flow Rod Bundle Film Boiling Tests," 1997 National Heat Transfer Conference, August 10-12, Baltimore, MD.A.2.59A. Y. Inayatov, "Correlation of Data on Heat Transfer Flow Parallel to Tube Bundles at Relative Pitches of 1.1 < s/b < 1.6," Heat Transfer-Soviet Research, 7.3, May-June 1975.A.2.60 FLECHT SEASET Program, Analysis of the FLECHT SEASET Unblocked Bundle Steam Cooling and Boiloff Tests, NUREG/CR-1533, EPRI NP-1460, WCAP-9729, January 1981.AREVA NP Inc.
A AR EVA Harris Nuclear Plant Unit 1 Realistic Large Break LOCA Analysis ANP-3011Q1(NP)
Revision 000 Page B-1 APPENDIX B: FLECHT-SEASET DATA Table B-I: FLECHT-SEASET Test 61607 0.81 inlsec -Data Data-Blocked Data-Bypass (in) °F in -°FI 12 707.81 12 707.81 24 954.44 12 709.90 39 1368.00 24 920.41 48 1582.48 24 925.56 48 1611.91 39 1334.23 60 1885.16 39 1337.41 60 1928.61 39 1340.56 60 1890.83 48 1614.11 60 1911.38 48 1575.98 60 1948.16 48 1615.20 67 1975.73 48 1610.81 67 1984.94 48 1603.11 67 1957.34 48 1598.73 67 1956.20 60 1881.77 67 2006.89 60 1885.16 67 1936.66 60 1894.22 67 2036.13 60 1836.48 67 1994.13 60 1798.03 67 1947.00 60 1864.78 67 1950.45 60 1889.69 69 1983.78 67 1936.66 69 2008.06 67 1927.45 69 2053.66 67 1932.06 71 2009.22 67 1984.94 71 1974.59 67 1987.23 72 1995.28 67 1979.19 72 1927.45 67 1956.20 73 1997.58 67 1957.34 73 1994.13 67 1915.97 74 2013.91 67 1978.03 74 2048.97 67 1926.31 74 1998.72 69 1959.64 75 2044.30 69 1969.98 75 2037.28 69 1960.80 75 2059.50 69 1971.14 75 2052.47 70 1991.83 75 2044.30 70 2031.44 75 2038.45 70 1942.41 76 2088.72 70 2012.73 AREVA NP Inc.
A AREVA Harris Nuclear Plant Unit 1 Realistic Large Break LOCA Analysis ANP-3011Q1(NP)
Revision 000 Page B-2 Data-Blocked Data-Bypass (in) (°I in) T°F 76 2113.56 71 1949.30 76 2079.38 71 2011.56 76 2075.88 72 2017.41 77 2072.34 72 2034.95 77 2068.84 72 2019.75 77 2091.06 72 2050.16 77 2099.25 72 1989.53 77 2043.14 72 2003.38 77 2121.91 72 2037.28 77 2098.09 72 2016.23 77 2124.28 72 2043.14 77 2120.72 72 2027.94 77 2089.91 72 1984.94 77 2096.91 72 1989.53 77 2107.59 72 2023.25 77 2096.91 73 2045.47 78 2105.19 73 2033.78 78 2109.97 73 1995.28 78 2099.25 73 1963.09 78 2096.91 73 1982.64 78 2119.53 73 2031.44 78 2112.38 73 1996.42 78 2080.53 73 2015.08 78 2101.63 73 2025.59 78 2117.13 73 1991.83 78 2142.19 73 2018.58 78 2135.03 73 1951.59 78 2119.53 74 2044.30 78 2112.38 74 2009.22 78 2121.91 74 2070.03 78 2051.31 74 2051.31 78 2121.91 74 2047.81 79 2136.22 74 2059.50 79 2115.94 74 2026.77 79 2150.53 74 2056.00 79 2155.31 74 2059.50 79 2124.28 74 2064.19 79 2142.19 74 2040.80 79 2115.94 74 2046.64 79 2127.88 75 2045.47 79 2188.72 75 2056.00 79 2132.66 75 1984.94 79 2115.94 75 2059.50 79 2096.91 75 1991.83 79 2149.34 75 2024.42 79 2127.88 75 2003.38 AREVA NP Inc.
A AREVA Harris Nuclear Plant Unit 1 Realistic Lame Break LOCA Analvsis ANP-3011Q1(NP)
Revision 000 Paae B-3 Data-Blocked Data-Bypass (in) (OF) (in) O°F)80 2139.81 75 2068.84 80 2164.88 75 2045.47 80 2144.59 75 2071.19 80 2151.75 75 2059.50 80 2125.50 75 2045.47 80 2138.63 75 2045.47 80 2173.22 76 2061.84 80 2143.38 76 2105.19 80 2138.63 76 2068.84 80 2148.16 76 2040.80 80 2167.25 76 2068.84 80 2109.97 76 2089.91 81 2107.59 76 2018.58 81 2123.09 76 2089.91 81 2162.47 76 2082.88 81 2168.44 76 2088.72 81 2152.94 76 2078.22 81 2058.31 77 2078.22 81 2143.38 77 2080.53 81 2114.75 77 2081.72 84 2126.69 77 2099.25 84 2051.31 77 1998.72 84 2030.27 77 2091.06 84 1975.73 77 2106.41 86 2063.00 77 2038.45 86 2082.88 77 2077.03 86 2046.64 78 2098.09 86 2044.30 78 2102.81 86 2054.81 78 2108.78 86 2066.50 78 2106.41 86 2038.45 78 2075.88 86 2067.69 78 2060.66 86 2112.38 78 1997.58 86 2084.06 78 2108.78 86 2084.06 78 2137.41 86 2060.66 78 2088.72 86 2104.00 78 2100.44 86 2081.72 78 2120.72 86 2086.38 78 2101.63 86 2070.03 79 2099.25 86 2050.16 79 2052.47 90 2081.72 79 2102.81 90 2068.84 79 2012.73 90 2104.00 79 2102.81 90 2117.13 79 2025.59 90 2095.75 79 2102.81 AREVA NP Inc.
A AR EVA Harris Nuclear Plant Unit 1 Realistic Large Break LOCA Analysis ANP-3011QI (NP)Revision 000 Page B-4 Data-Blocked Data-Bypass (in) (°I in) -°F 90 2079.38 79 2093.41 90 2138.63 80 2163.69 90 2142.19 80 2166.06 90 2127.88 80 2112.38 90 2152.94 80 2121.91 90 2137.41 80 2119.53 90 2094.56 80 2149.34 90 2115.94 80 2024.42 90 2100.44 80 2082.88 90 2138.63 80 2078.22 90 2113.56 80 2105.19 90 2098.09 80 2060.66 90 2121.91 81 2111.16 90 2088.72 81 2114.75 90 2074.69 81 2101.63 90 2088.72 84 1979.19 90 2088.72 84 1991.83 90 2018.58 84 2005.72 96 2071.19 84 1975.73 96 2046.64 84 2022.09 96 2107.59 84 1982.64 96 2092.25 84 1896.48 96 2145.78 86 1971.14 96 2138.63 86 2050.16 96 2135.03 86 1951.59 96 2108.78 86 2034.95 96 2118.34 86 2048.97 96 2142.19 86 2052.47 96 2104.00 86 2065.34 96 2101.63 86 1990.67 96 2087.56 86 1975.73 96 2089.91 90 2004.55 96 2112.38 90 2064.19 96 2093.41 90 2056.00 96 2068.84 90 2074.69 96 2011.56 90 2059.50 102 1967.69 90 2066.50 102 1899.88 90 2082.88 102 1922.86 90 2040.80 102 1943.55 90 1997.58 102 1856.86 90 2079.38 102 1885.16 90 2046.64 111 1789.11 90 1997.58 111 1810.45 96 1975.73 111 1834.23 96 2040.80 111 1890.83 96 2039.63 AREVA NP Inc.
A AREVA Harris Nuclear Plant Unit 1 Realistic Lame Break LOCA Analysis ANP-3011Q1(NP)
Revision 000 Pa-qe B-5 Data-Blocked Data-Bypass (in) (°I in) -°F 111 1874.97 96 1975.73 111 1874.97 96 2057.16 111 1846.67 96 2023.25 111 1843.28 96 2010.39 111 1833.09 96 1966.55 111 1769.03 96 2020.92 111 1844.42 96 2057.16 111 1839.89 96 2056.00 111 1820.64 96 1992.98 111 1808.19 96 2023.25 111 1744.48 96 2002.22 111 1728.86 102 1848.94 120 1746.70 102 1861.39 120 1770.14 102 1886.30 120 1805.94 102 1817.25 120 1742.25 111 1800.27 120 1766.80 111 1718.81 120 1736.67 111 1696.55 120 1747.83 111 1670.17 120 1777.95 111 1791.34 120 1725.52 111 1758.98 120 1706.55 111 1675.66 120 1708.77 111 1689.95 120 1761.22 120 1621.80 120 1775.72 120 1632.80 120 1670.17 120 1696.55 120 1698.75 120 1611.91 120 1667.97 120 1535.89 132 1368.00 120 1595.48 132 1461.64 120 1525.05 132 1416.75 120 1583.56 132 1361.67 120 1681.16 138 1306.80 120 1652.58 138 1375.39 120 1598.73 120 1644.89 132 1327.91 132 1384.88 138 1243.08 138 1314.19 138 1299.42 AREVA NP Inc.
A AREVA Harris Nuclear Plant Unit 1 Realistic Lame Break LOCA Analysis ANP-3011Q1(NP)
Revision 000 Pane B-6 Table B-2: FLECHT-SEASET Test 61607 0.81 inlsec -S-RELAP5 Viiiii AREVA NP Inc.
A AR EVA Harris Nuclear Plant Unit 1 Realistic Large Break LOCA Analysis ANP-3011QI(NP)
Revision 000 Page C-1 APPENDIX C: M5 HIGH TEMPERATURE SWELLING AND RUPTURE MODEL For ease of reference, this Appendix contains a copy of BAW-1 0227P-A (Reference
[9]), Rev.1, Appendix C "M5 High Temperature Swelling and Rupture Model which is the basis for the M5 swelling and rupture model utilized to support responses to RAIs in this document.AREVA NP Inc.
A AREVA Harris Nuclear Plant Unit 1 Realistic Large Break LOCA Analysis ANP-3011Q1(NP)
Revision 000 Paqe C-2 This page intentionally left blank AREVA NP Inc.
MS Allov ToDical FCF M5 HIGH TEMPERATURE SWELLING and RUPTURE MODEL C. I NTRODUCTION The Framatome Technologies Incorporated (FTI) LOCA evaluation models, References C-1 and C-2, require the simulation and calculation of fuel cladding swelling and rupture.For Zircaloy-4, the LOCA modeling is based on NUREG-0630, Reference C-3. The models include provision for the determination of the occurrence of rupture as a function of the cladding surface temperature, the degree of swelling at the rupture location, and the amount of assembly average flow area reduction.
The flow area reduction is also used to determine the amount of entrained water droplet mechanical inter-action within the fuel assembly at the location of rupture. This interaction causes droplet shattering which reduces vapor temperatures and increases cooling efficiency.
The NUREG-0630 model is based on Zircaloy-4 testing. M5 is a new material with differing creep characteristics and will behave differently during the high temperature strain and rupture of a LOCA. Because NUREG-0630 can not be applied without recognition of the differing behavior of MS, rupture testing at the CEA EDGAR facility in Saclay, France was conducted to determine M5 characteristics, Reference C-4. This Appendix presents the fuel cladding swelling and rupture model and materials data that have been developed byE lil. approach.The FTI swelling and rupture model'is divided into the determination of individual pin characteristics, bundle flow blockage effects, and bundle fluid droplet interaction effects.Each of these, as related to the M5 material, is discussed in sequence within this Appendix.
Also presented is a brief description of the EDGAR test facility and the results obtained for the M5 cladding.C.2 INDIVIDUAL PIN CHARACTERISTICS C-3 MS Alloy Topical FCF_C.2.1 EDGAR Test Apparatus and Data C-4 WAC All Tnn;o-af FCF W-L FOR C.2.2 Fuel Pin Cladding Rupture Temperature Versus Stress C-5 ILIJ A-Il^. *-,I"IwM]C.2.3 Fuel Pin Cladding Rupture Strain versus Rupture Temperature C-6 Mr%' Allnu Tnnni--n1 Fr-F j NA~ Allnv Tnriir~m1 FCF I C.2.4 Fuel Pin Cladding Strain Prior to Rupture and Remote from the Rupture Location C.2.4.1 Fuel Pin Cladding Strain Prior to Rupture and Gap Heat Transfer I C-7 M5 Alloy Tonic~al FCF?vf5 AlInv Tani&~aI 1~CF C.2.4.2 Fuel Pin Cladding Strain Prior to Rupture and Flow Blockage i C-9 Ikfrt Allr% Td% r"l FC1l7 C-9 U'q Allnv "lnnirmlM~ AlnvTnnr~dFCF 7 7 C.3 BUNDLE FLOW BLOCKAGE CHARACTERISTICS C.3.1 Assembly Simulation Effects C-10 U.31. A xle To% stribtinoAlptr T:rr x 19 , I -F C.3. 1.1 Axial Distribution of Rupture C-II.
M5 Alloy Topical FCF'C-12 M5 Alloy Tonical FCF C.3.1.2 Number of Fuel Pins in Simulation or Assembly C.3.1.3 Unheated Surfaces and Pellet Simulation C-13 MS AllayTon~ical F-cF: C.3.2 The FTI Blockage Simulation Model C.3.2.I Blockage Limitation C-14 Mj.AIIM TopiW 145 lla TopcLF-CF C-15 MW Allnv Tnnir.ql C-16 a XAC Allm Tgrv%;p-A Fri: -C.3.2.2 Bundle Simulation Benchmarks C.3.2.2.1 Benchmarks of the Chapman Bundle Tests C- 17 M5 Alloy Topical FCF C-18 Aflnl TriAr-a W~rip C.3.2.2.1 Benchmarks of the NUREG-0630 Blockage Calculation C- 19 Wkq All 'r ;-1 Y-L wig rur C.3.2.3 Blockage Model Co'nservatisms C.4 DROPLET INTERACTION MODELING C-20 I MS Aloy Topica F C.5
SUMMARY
OF FTI MS CLADDING SWELLING AND RUPTURE MODEL I C-21 L4 Allnu Temir-l FiwF C.6 REFERENCES C-1. BAW-10168-A Rev. 3, RSG LOCA, BWNT Loss-of-Coolant Accident Evaluation Model for Recirculating Steam Generator Plants, B&W Nuclear Technologies, Lynchburg, Virginia, 1996.C-2. BAW-10192-P, BWNT LOCA, BWNT Loss-of-Coolant Accident Evaluation Model for Once-Through Steam Generator Plants, B&W Nuclear Technologies, Lynchburg, Virginia, 1994.C-3. D. A. Powers and R.O. Meyer. NUREG-0630, Cladding Swelling and Rupture Models for LOCA Analysis, U.S. Nuclear Regulatory Commission, Washington DC, 1980.C-4. Letter J. Hivroz (CEA) to A. Le Bourhis (Framatome
/ TFP), Alliage MS.Transmission des resultats portant sur les essais de fluage et de rampe de temperature effectues sur le dispositif d'essai EDGAR, SRMA/97.391, July 3, 1997, Saclay, France. also included herein as Appendix A.C-5. Note Technique SRMA 84-1346, EDGAR FROID -Gaines Framatome, Modelisation de la Deformation des Gaines en Ziracloy dans des Conditions d'un Accident de Penre de Refrigerant Primaire Critere de Rupture, Service de Recherches Metallurgiques Appliquees.
Centre d'Etudes Nucleaires de Saclay, France, on file at.Framatome Technologies Inc., Lynchburg, Virginia, as 38-1247171-00.
C-6. F.J. Erbacher and S. Leistikow, " Zircaloy Fuel Cladding Behavior in a Loss-of-Coolant Accident:
A Review," Zircaloy in the Nuclear Industry, Seventh International Symposium, American Society for Testing and Materials, Philadelphia, Pennsylvania, 1987.C-7. F. J. Erbacher et al, "Burst Criterion of Zircaloy Fuel Claddings in a Loss-of-Coolant Accident," Zirconium in the Nuclear Industry, Fifth Conference, ASTM STP 754, D. G. Franklin, Ed. American Society for Testing Materials, 1982, pp.271-283.C-8. D.G. Hardy, " igh-Temperature Rupture Behavior of Zircaloy Tubing," CONF-730304, USAEC/TIC, Water-Reactor Safety, 1973.C-9. BAW-10164-A Revision 3, RELAPS/MOD2-B&W, An Advanced Computer Code for Light Water Reactor LOCA and Non-LOCA Transient Analysis, B&W Nuclear Technologies, Lynchburg, Virginia, 1996.C-22 MS Alie Tonkal W,(IP C-i0.NUREG/CR-0103, ORNL/NUREGIrM-200, Multirod Burst Test Program Progress Report for July-December 1997, Us Nuclear Regulatory Commission, Washington DC.C-I LNUREGICR-0655, OR.NLYUREGITM-297, Multirod Burst Test Program Progress Report for July-December 1998, Us Nuclear Regulatory Commission, Washington DC.C-12. NUREGICR-1023, ORNI/NUREG/TM-351, Multirod Burst Test Program Progress Report for April-June 1999, Us Nuclear Regulatory Commission, Washington DC.C-13. WA Fiveland and AR Barber, BW-4702, Rupture Characteristics of Zircaloy-4 Fuel Cladding Supplemental Report -Ruptured Clad Geometry, Babcock and Wilcox Company, Alliance, Ohio, February 1978.C-14. D.A. Powers and R.O. Meyer, NUREG-0536, Evaluation of Simulated-LOCA Tests that Produced Large Fuel Cladding Ballooning, U.S. NRC, March 1979.C-15. R.E. Williford and C.R. Hann, uEffects of Fill Gas Composition and Pellet Eccentricity," Battelle Northwest Laboratories Report, BNWL-2285, July 1977.C-16. A.L. Lowe, jr. (Babcock and Wilcox Company).
letter to D.A. Powers (NRC),.October 10j 1978, Available in file for USNRC Report, NUREG-0536, U.S. NRC.C-23 MS Alfnv Tnnic~al FCF I?vfS Alinv Tonical FCF]-Table C-1. EDGAR Swellina and RuDture. Slow Ramo Tests C-24 MS Alloy Tonical FCF -MS Ahoy Tonical FCF I C-25 C A All~t 'r- ,: Io r im LVIC &I,. k!: ke --Table C-2-EDWAR Swelling and Rupture, Fast Ramp Tests a C-26 M5 Alloy Topical FCF CL A.0i C-2 Figure C-2. EDGAR Test Results Rupture Temperature versus Stress 0 C)
CV)co)Figure C-3. Chapman Rupture Temperature Correlation versus M5 Data Slow Ramp Rate 0-n 0 Figure C-4. Chapman Rupture Temperature Correlation versus M5 Data Fast Ramp Rate Figure C-5. NUREG-0630 Rupture Strain versus Rupture Temperature Curves 0 6 ,C)
Figure C-6. M5 Rupture Strain versus Rupture Temperature Slow Ramn Rate U'5.I.,0"11 Figure C-7. M5 Rupture Strain versus Rupture Temperature Fast Ramp Rate 5-0 9n CD0 Figure C-8. M5 Rupture and Pre-Rupture Strain ersus Rupture Temperature Slow Ramp Rate Figure C-9. M5 Rupture and Pre-Rupture Strain versus Rupture Temperature Fast Ramp Rate cli Figure C-10.a Pin Strains in MBTP Bundle B-1 for Ruptures Above Lower Grid 0 0 9"C1 Figure C-1O.b Pin Strains In MBTP Bundle B-1 for Ruptures Just Below Upper Grid ,I 0"-4 Figure C-IO.c Pin Strains In MBTP.Bundle.B-1 for Ruptures Above Upper Grid Figure C-11.a Pin Strains In MBTP Bundle B-2 for Ruptures Just Above Lower ,- Grid 0 00 0o Figure C-11.b Pin Strains in MBTP Bundle B-2 for Ruptures Just Below Upper Grid"n Figure c-I I .c Pin Strains In MBTP Ruindle B-2 for Ruptures Above Upper Grid 00 00 n1 Figure C-12.a Pin Strains in MBTP Bundle B-3 for Ruptures Just Above Lower Grid"n 00 C,0 AT Figure C-12.b Pin Strains in MBTP Bundle B-3 for Ruptures Just Below Upper Grid CA'11 Figure C-13. Charterization of Axial Distribution of Strain for Ruptured M5 Cladding ,
Figure C-14. Cladding Temperature Profiles for the B&W Designed NSS Mid-Core Peak Grid Grid aJ-2 Inlet or Outlet Peak S 'with edge of core shifting~/U /Axial Dimension Figure C-15. Cladding Temperature Profile for NSSs with Recirculating Steam Generators E h.a 0 0.Axial Dimension Figure C-16. Probability Density Function for The Axial Position of Rupture within A Grid Span 2- > 1 -I I a1 I -------1.8 --- -------1.4 -- -, I --a I a a a I " a I II I I a 1. I -r 4 -, --I I a a a I I I i iI I " a a I a-----.2----------------
.----------
a---------------
----a-----------
OI~ a a a , I I I ---a- a- -IL* a a a a I a a I I I a a , a aI a I aI a a a a a 0 .... .T ----- -------- ----------
r ...------
, --------.
.--- ----------a a I a a -I II a a a a I I a a a a a a ' a a I 00 a I I a a a I -a a a a a I a aI a a I a a a a a- a a a I a a a I a a I I I 0 .2OA I ..0 a 0 0.9 N r m ai. a arom T a G I a a -* a a a a a .a a a a 0I- e I I--I---I-.
0I. .0. 0. 0.5 0.6 0. 0. 0.t~orahzd Ditane frm Tp ofGri A All^; 10ftml crc ILI!& =~IU I -Wb! I~Figure C-17. Azimuthal Temperature Gradient Effect on Cladding Strain Reprinted from Figure 22.of Reference 4.,Original by FJ Erbacher, 0 o C 0 L..LI)4-U, a-.0 0 4-C 0)a-05Q-E U a-U 7/1 I I I I.REBEKA -burst I A I I I criterion model (curvei)I I I I I I z_.t c Constant internal overpressure
- 65 bar* time of burst : 482 -512s-burst temperature 827 1 845*C 42 REBEKA -data ,. single rod tests Obundte tests I I I I 0 I-4 I I I I 0------ J I --I I 50. .100 cladding temperature difference.
ATaz[K]azimuthal C-48 US; AIlnv "Tnnin--l Figure C-18. Pin Rupture Strains for MBTP Tests B-1 and B-2 100 90 80 70 CE CL 12 60 50 40 30 20 10 0 O-600 700 800 900 1000 1100 Rupture Temperature, C 1200 C-49 MS Allov Tooleal FCF Figure C-19. Pin Rupture Strains for MBTP Test B-3 100 90 80 70*0 B.C.60 50 40 30 20 10 0 4-600 700 800 900 1000 1100 Rupture Temperature, C 1200 1-- ,NUREG-0630 Slow Ramp r3 B-3 Test 10 C/s ---Factored 06301 C-50 a Figure C-20. MBTP Test B-1 Probability Density Function for Rupture Location 5 ,o)U, 02 0 2 4 a a 10 12 14 1 18 -20 22 24 26 Position, in 28 30 32 34 Figure C-21. MBTP Test B-2 Probability Density Function for Rupture Location M 0;4 -------_ -- _3~.z-n 00"1 "1 l II'1 0o 0 2 4 6 8 10 12 14 16 16 20 22 24 26 28 30 32 34 Position, in
, U'06 E z.Figure C-22. MBTP Test B-3 Probability Density Function for Rupture Location 5 -------------.-- --.2 [Hll L I_0 2 4 6 8 10 12 14 1B 1I Position, in 20 22 24 26 28 30 32 34 Figure C-23. Benchmark of MBTP Test B-I 100.90 I I- -I " I -I --I -I I 90 -----------
-- -- -- -- -- ---- -- -- -- -----------
--- -- -- ---- --- ---------
J .. .. .00 .---------------
--.--- ---------------------
III I I I I I 30- --- --- --- --, --- --, -.--20 I ----------------
r---------
r---------
- I I I--0 --. .....*1 ., I I I 0. 5 10 is 20 25 30 35 40 Height, in -1, MBTP Minimum Blockage Assumption Sm MBTP Maximum. Blockage AssumptionI Figure C-24., Benchmark of MBTP Test B-1 with All Rods Allowed to Strain*1!C 100 90 80 70 60 50 40 30 20 10 0 0 CA Cn 010 15 20 25 30 35 4 Height, In"-- MBTP Maximum Blockage Assumption1 l0-MBTP Minimum Blockage Assumption Figure C-25. Benchmark of MBTP Test B-2 0 V O ft 100 90 80 70 s0 50 40 30 20 10 0 0 5 10 15 20 25 30 35 Height, In 40 I ý M13TP Minimum Blockage Assumption own= anew MBTP Ma)dmum. Blockage Assum-MBTP Minimum Blockage Assumption
---MBTP Maximum, Blockage Assumptlo 9]
Figure C-26. Benchmark of MBTP Test B-2 with All Rods Allowed to Strain 0 100 90 80 70 60 50 40 30 20 10 0 0 0 5 10 15 20 25 30 35 Height, in 40 I'.. I-MBTP Minimum Blockage Assumption
-, MBTP Maximum Blockage Assumptionf Figure C-27. Benchmark.
of MBTP Test B-3 100 90 80 70 0 60 so o 30 20 10 0 0 0.0 5 10 15 20 25 30 35 Height, In 40 1n'MBTP Minimum BlockageAssumption
-MBTP Maximuni Blockage Assumption Figure C-28. Benchmark of MBTP Test B-3 with All Rods Allowed to Strain.2!C)0 100 9o 80 70 60 5o.40 30 20 10"0 0 0 5 10 15 20 25 30 35 Height, in 40 I -MBTP MInImum Blockage Assumption MBTP Ma)dmum. Blockage Assumption I MBW Minimum Blockage Assumption
--MBTP Maximum Blockage Assumption Figure C-29. Blockage Technique Comparison to NuReg-0630
-Slow Ramp Rate 00 Figure C-30. Blockage Technique Comparison to NuReg-0630
-Fast Ramp Rate 0._)120 .
Figure C-31. Flow Blockage Between the Mark-B and Mark-BW Assemblies Slow Ramp Rate 5.-I 0 In 0 In Figure C-32. Flow Blockage Between the Mark-B and Mark-BW Assemblies Fast Ramp Rate 5-'1 Figure C-33. Chapman Rupture Temperature Correlation for 0, 10, and 28 Cls Ramp Rates-yl>"11 0)'A Figure C-34. M5 Strain and Blockage Curves for Slow Ramp Rates 0 I C["
Figure C-35. M5 Strain and Blockage Curves for Fast Ramp Rates 5.i-