ML20210K765
| ML20210K765 | |
| Person / Time | |
|---|---|
| Site: | Three Mile Island  |
| Issue date: | 07/09/1999 |
| From: | Pacheco K, Page D FRAMATOME |
| To: | |
| Shared Package | |
| ML20210K743 | List: |
| References | |
| 86-5002073-02, 86-5002073-02-R02, NUDOCS 9908060215 | |
| Download: ML20210K765 (111) | |
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SUMMARY
SHEET (CSS) j DOCUMENT IDENTIFIER 86-5002073-02 TITLE Summary Report for BWOG 20% TP LOCA PREPARED BY: REVIEWE0 BY: NAME D.R.Page NAME K.S.Pacheco SIGNATURE R00m f. b S<GNATuRE Y M 4 ) , ,,! d TITLE Engineer ll kATE Supe isory Engineer yA y TITLE DATE 9' g COST CENTER 41010 REF. PAGE(C) 109,110 TU STATEMENT: - REvi2WER INDEPENDENCE ..- A PURPOSE AND
SUMMARY
OF RESULTS: " l l The results of new ECCS analyses of record for LBLOCA and SBLOCA transients are describ3d in this document. Individual LOCA linear heat rate (LHR) limits are provided for the Mark-89 fuel design with 20 percent SG tube plugging and a minimum DNB flow of 102 percent of the design flow. Section 2 summarizes the results that consider both the pump data and two-phase degradation (PSC 1-99) and the corrected reflood grid data. Sections 4 and 5 discuss specific aspects of the large and small break LOCA analyses, respectively, including generic- and plant-specific sensitivity studies. Section 6 summarizes the compliance with the acceptance criteria of 10 CFR 50.46. Finally, Section 7 identifies the NRC Safety Evaluation Report (SER) restrictions on the BWNT LOCA evaluation model, BAW-10192-PA, and shows that they have been met for these I analyses. This revision was completed to provide a more complete representation of the analyses performed. The generic and plant specific sensitivity studies that support both the l LBLOCA and SBLOCA analyses are described. Updated LBLOCA results generated in response to the corrected grid data are presented. Detailed discussions of the energy deposition factor and the SER restriction on the analyzed axial peaking factor are l provided. Additionally, plots illustrating the LBLOCA and SBLOCA transient ! progressions are included. l l THE FoLLoWING COMPUTER CODES HAVE BEEN USED IN THIS DOCUMENT: l CODE I VERSION I REV CODE / VERSIONI REV THIS DOCUMENT CONTAINS ASSUMPTIONS THAT MUST BE D D 89 ON F -RE TE ORK P PDR YES ( ) NO ( X ) PAGE 1 oF 110 u
Framatome Technolooies. Inc. 86-5002073-02 RECORD OF REVISIONS Revision Chanoe/ Para. Descriotion 00 NA Originalissue 01 5-7,11,21,29,30, & 34 Modified text to reflect new analyses due to PSC 1-99. 3,21, & 23-25 Deleted Section 4.4. 34 Identified proprietary references. 02 Marked with change bars. Add additionalinformation on SBLOCA and LBLOCA 177-LL generic and plant- < specific sensitivity studies. Additional discussion on energy deposition factor and SER restriction relating to analyzed APF.
~ Updated LBLOCA results in response to reanalysis with corrected grid data.
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Framatome Technoloaies. Inc. 86-5002073-02 TABLE OF CONTENTS List of Ta b le s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ................................................5 Li s t o f F ig u re s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
- 1. I n tro d u ctio n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
- 2. S u mm a ry o f Re s uits . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
- 3. Pla nt Paramete rs a nd i np uts . . . .. . . ... ...... ... . . . . . . .. . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . 13 3 .1 L B LO C A Analy se s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . .
- 3. 2 S B LO CA A n alys e s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
- 4. LBLOCA Sensitivity Studies and Analyses....................................... . .... ...... ...... 25 4.1 L B LOCA Sensitivity Stu dies.... ... . .. .. .. . . . .... ... .... . . ....... .. .. . .... .. . . . .... .. .. . . ...... ... . . . . . . . . 2 5 4.1.1 LBLOCA Evaluation Model Generic Sensitivity Studies.... . . . . . . . . . . . 25 4.1.1.1 RELAP5/ MOD 2-B&WTime-Step Study. . . . . . . . . . . . . . . . . . . . . . . . . , . 25 4.1.1.2 RELAPS/ MOD 2-8&W Pressurizer Location Study.. . . . . .. .. .26 4.1.1.3 RELAP5/ MOD 2-B&W Break Noding Study.... ...... . .. . . . . . . . .. 26 4.1.1.4 RELAP5/ MOD 2 8&W Core Crossflow Study.. . . . . . . . . . . . . . . .. 26 4.1.1.5 RELAP5/ MOD 2-B&W Core Noding Study . .. . . . . . . . . . . . . . 27 1
4.1.1.6 RELAPS/ MOD 2-B&W ECCS Bypass Study... . . . . . .... . . .. . . .27 ( 4.1.1.7 REFLOD3B Loop Noding Study.. . . ........ . . . . . . . . . .. .. . . . . . . . . . . 27 4.1.1.8 REFLOD3B RCP Locked Versus Free-Spinning Rotor Study.. .. .. . . .27 4.1.1.9 BEACH Time Step Study.. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 27 4.1.1.10 BEACH Axial Fuel Segmentation Study.. . . . . . . . . . . . . . . .. .. 28 { 4.1.1.11 AxialVersus Radial Core Peaking Factor Study.. . . . . . . . . . . . . . . . .... . 28 I 4.1.2 LBLOCA Evaluation Model Plant-Specific Sensitivity Studies... . . . . . . . . . . .28 4.1.2.1 RELAPS/ MOD 2-B&W Pump Degradation Study... .. .. . ... . ... . ..... . . . 28 4.1.2.2 RELAP5/ MOD 2-B&W RC Pump Power Study.. .. . . . . . . . . . . . . . . . . . .29 4.1.2.3 Containment Pressure and ECCS Configuration Study.. . . . ... . . . . . . .29 4.1.3 LBLOCA Break Spectrum Study........ . . . . . . . . . . . .. . .. .. .. .30 4.1.4 ~ CFT Initial Conditions Study .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .32 l 4.2 LB LO CA An alyses at 2 772 MWt ... .. . ..... ........ .. ... ......... .. . ........... . . . ... . . . ...... . . . ....... 34 ! 4.2.1 Base Model .... . . .. . ... .. ..... . . . ... . . . . . . . . . . . . . . . .34 4.2.2 Transient Progression.... .. . . . ... . . . . . . . . . . . . . . . . . . . .... . . . .35 4.2.3 LI I R Limits....... ...... .. . . . . . . .. .... . .. . . . . . . . . . . . .35 4.2.3.1 BOL LOCA LHR Limits .. ..... . . .. . ...... . . . . . . . . . . . .. . . . . . . .. .36 4.2.3.2 MOL LOCA LHR Limits . .. . . .. .. . . . . . . . . . . . . ... . . . . .. . . 36 4.2.3.3 EOL LOCA LHR Limits.. . . . . . . . . .. . . . . . . . . .. .. .. . 37 4.2.3.4 Partial-Power LOLOCA LHR Limits.. .-. . . . . ........ .. . . . . . . . 37 4.2.3.5 Core Inlet and Exit LHR Limits .. . ... .. . .. .. . . . . . . . . . . . . . . . . . . .. .38 4.2.4 Recent LBLOCA issues.. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .39 , 4.2.4.1 Energy Deposition Factors....... . .. . . . . . . . . . . . . .. .. . ... .. .39 ; 4.2.4.2 Bumup FuelThermalConductivity. . . . .. . . . . . . . . . . . . . . . . . . . . 40 I 4.2.4.3 Increased Containment Spray Flow (1800 gpm vs.1500 gpm). . . .. . . . . . 41 l 4.2.4.4 Pump Type ...... .... . .. .. .. . . ..... . . . . . . . . . . . . . . . . . . . .41 4.2.4.5 Fuel Dimensions.. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . 42 4.2.4.6 Grid Type .. .. . . .. . . . . . . .. . . . . . . . . . . . . . . . . . . . ... . .. 42 4.2.4.7 Replacement 15% Tube Plugging LHR Limits for TMI-1.. .. . . . . . . . . . . .42 4.2.4.8 Extension of LBLOCA Results to Greater than 60,000 mwd /mtU.... .. .. . . .42 l 3 j 1
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Framatome Technoloaies. Inc. 86-5002073-02
- 5. SBLOCA Sensitivity Studies and Analyses .. ...... ...... . .. . ..... .. . .... . . . . . .68 5.1 S BLOCA Sensitivity Studies .. . . . . . . . . . . . . . . . . . .. . ... . . . . . . . . . . . . . . . . . . . . . . . . . . . 68 5.1.1 SBLOCA Evaluation Model Generic Studies. .68 5.1.1.1 SBLOCA Time-Step Study . . . . .68 5.1.1.2 SBLOCA Pressurizer Location Study.. .69 5.1.1.3 SBLOCA Core Crossflow Resistance Study . .69 5.1.1.4 SBLOCA Core Channel Modeling Study.. . . . .69 5.1.1.5 SBLOCA CFT Line Resistance Study. .. . . .70 5.1.1.6 SBLOCA Break Discharge Coefficient Study. . .70 5.1.2 SBLOCA Evaluation Model Plant-Specific Sensitivity Studies. . .71 4 5.1.2.1 CFT Level for CLPD Breaks. . . . . 71 l 5.1.2.2 CFT Line Break.. . . . . . . . . . . 71 l 5.2 S B LO C A Analyse s at 2772 MWt . .. . . . . . . ... .. . .. . . . . . . . . . .. . ..... . . . . .. . . . . . . . . . . . . .. . . . . .. . . 7 3 5.2.1 SBLOCA Base Model . . .73 5.2.2 SBLOCA Transient Progression. . .. . .75 5.2.3 Break Spectrum Analysis at RCP Discharge.. , .78 i 5.2.3.1 Small SBLOCAs at RCP Discharge. .. . . .78 5.2.3.2 Intermediate SBLOCAs at RCP Discharge.. .. . , . 79 5.2.3.3 Large SBLOCAs at RCP Discharge. . .79 5.2.4 Special Breaks.. . . . . . . , . .80 5.2.4.1 CFT Line Break.. . .... . .. . . . .60 5.2.4.2 HPl Line Break.. . .. .. . ... . .91 1 5.2.5 Recent SBLOCA issues.. . . . . . . . .. . . .. . 81 5.2.5.1 EDF., . . . . . . .. .. .. .. . 81 5.2.5.2 Assumption of Manual HPI Actuation.. .. .... .. .82 5.2.5.3 Pump Type . . . .. . .82 i 5.2.5.4 MSSV Lift Tolerance.. .. . . . . . . . . . .. .83 !
- 6. 10 C F R 5 0.4 6 C rite ria . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 9 6.1 Pea k C lad d ing Te m p e ratu re . . . .. . . . ... . . . . . . . . . . . . ... .. . .. .. . .. . .. . .. .. . .. .. . . . .... .... . . . . . .. . . . . . . 9 9 6.2 Local Cla d d in g Oxid atio n . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 9 6.3 Whole-Core Oxidation and Hydrogen Generation ............ ....... ....... ....... .. .. 100 6.4 C o re G e o m etry . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .100 6.5 Lo ng -Te rm Cooling .... . . . .. . . . ... .. .. . .. . . ... . . . .. . .. . . .. . ......................................101
- 7. RELAPS/ MOD 2-8&W EM SER Restrictions.......... .......... ............. ......... .. . 103
- 8. R e f e r e n ce s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 l
g l l
- framatome Technoloaies. Inc. 86-5002073-02 List of Tables 1
- Table 2-1: Summary of Mark-B9 'LHR LBLOCA Limits .................................................. 9 1
Table 2-2: . Summary of Calculated PCTs for Mark-B9 SBLOCA Analyses..................12 Table 3-1: LBLOCA inputs and Assumptions for Burnups of 0-60 Gwd/mtU.......... ....14 Ta ble 3-2: Ope rator Actio ns for LB LOCA ... ..... .......... ...... ...... . .... .... ... .. . ... ... . . ........... . .. 15 Table 3-3: L B L-O CA Dopple r Re a ctivity . . . . . .. . . . . . . . . .. . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . .. . . .. . .. . . . . . . . . .. 15 Table 3-4: L B LOC A LP I F lows . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . Table 3-5: . LBLOCA Moderator Density Reactivity inputs.............................................16 Table 34: Mark-89 Radial Power Profile History ........................................................17 Table 3-7: Containment Assumptions - LBLOCA Minimum Backpressure Analysis....18 Table 3-8: SB LOC A Inputs and Assumptions .. ....................... .. .. . .. .... . ........... .. .. .... .... . 19 Table 3-9: , Operator Actions for S BLOCA ....... . .. .... ................. ........... . ....... .... ... . . .. . ...... 20 Table 3-10: SBLOCA HPI Flow Rates - CLPD Break ................................................. 21
- Table 3-11: SBLOCA HPl Flow Rates - HPI Line Break ............................................. 21 Table 3-12: SBLOCA HPl Flow Rates - CFT Line Break ............................................ 22 Table 3-13 : S BLOC A LPI Flows . . .... . . ... . .. . . . . . . . . . . . . . .. ... . . . . .... .. .. .. . . . . .. . . . .. . .. . . ....... ..... . . . .
Table 3-14: S B LOCA S cra m C u rve . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Table 3-15: SBLOCA Moderator Density Reactivity inputs.......................................... 23 Table 4-1: t ANO-1/TMI-1 Hot Channel initial Conditions Used for the LBLOCA LHR i L im it A n a ly se s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ! Table'4-2: ANO-1/TMI-120% Tube Plugging 2772 MWt Mark-B9 BOL LBLOCA LHR
- Lim it s S u m m a ry . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
Table 4-3: ANO-1/TMI-120% Tube Plugging 2772 MWt Mark-89 MOL LBLOCA LHR j
-- Lim it s S u m m a ry , . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . Table 4-4: ANO-1/TM/-120% Tube Plugging 2772 MWt Mark-89 EOL LOCA LHR Lim it s S u m m a ry . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
Table 5-1: SBLOCA CFT Volume and Pressure inputs............................................... 72 Table 5-2, Summary of Small SBLOCA Results 0.01 fta to 0.08 ft2 ............................. 84 1 Table 5-3 Summary 'of Intermediate SBLOCA Results 0.9 ft2 to 0.15 ft2 .................... 85 i Table 5-4. Summary of Large SBLOCA Results 0.3 ft2 to 0.75 ft2......... ..................... 86 l
. Table 5-5. Summary of HPI and CFT Line Breaks ...................................................... 87 List of Figures !
Figure 2-1: TMI/ANO 20% Tube Plugging LBLOCA LHR Limits at 2772 MWt with I Burnup..................................................................................................................10 ' Figure 2-2: TMI/ANO 20% Tube Plugging LBLOCA LHR Limits at 2772 MWt with
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i Elevation................................................................................................................10 Figure 2-3: MTC Limit vers us Power Level ..................... ................ ... .......... .... .......... ..11 Figure 2-4: = SBLOCA Spectrum, PCT versus Break Size ................................ ...........12 i Figure 3-1: LBLOCA Minimum Containment Pressure Response....... ....................... 24 i Fig u re 4- 1 : ' Axial Power Shapes . .. .. .. . . .. ...... ... ... . .... . .. . .... .. .. .. . . .. ... . . . ... . . .. .... . .. . ... .. .. . . . .. . . 47 4 5 l 1 l 3
Framatome Technolpaies, Inc. 86 5002073-02
- 1. Introduction Entergy Operations and GPU Nuc! car operate the B&W-designed plants ANO-1 and TMI-1, respectively. Previously, the ANO-1 plant has been shown to be in compliance with the five criteria of 10CFR50.46 based on the CRAFT 2-based emergency core cooling system (ECCS) evaluation model (EM) described in BAW-10104 Rev. 5 (Reference 1) for large break loss-of-coolant accident (LBLOCA) analyses and in BAW-10154 (Reference 2) for small break loss-of-coolant accident (SBLOCA) analyses.
Framatome Technologies incorporated (FTI) has since developed and received NRC ' approval for use of a RELAPS-based EM, described in BAW-10192-PA (Reference 4), l to replace the CRAFT 2-based EM. Subsequently, FTl prepared RELAPS/ MOD 2-B&W, REFLOD38, and BEACH input models that represent the TMI-1 plant in accordance with this new EM and showed that TMI-1 was in compliance with the five criteria of 10CFR50.46 for both I.BLOCA and SBLOCA analyses (Reference 5). The LOCA analyses performed using either of these methods are conservative and represent valid licensing results based on the boundary conditions analyzed. The LOCA analyses of record for both ANO-1 and TMI-1 are now being revised with new analyses that represent improved modeling techniques and/or new plant conditions. Specifically, new RELAP5/ MOD 2-8&W LOCA analyses have been performed for the TMI-1 and ANO-1 plants which model steam generator tube plugging levels up to 20 percent (15 percent in the intact loop,25 percent in the broken loop), a minimum departure from nucleate boiling (DNB) flow of 102 percent of design flow, and a core power level of 102 percent of 2772 MWt. The new RELAPS/ MOD 2-8&W LOCA analyses are suitable for replacement of the ECCS analyses of record for the TMI-1 and ANO-1 plants. They were performed in compliance with the EM methods and the limitations and restrictions stated in the NRC Safety Evaluation Report (SER) on BAW-10192-PA. The large and small break LOCA generic noding and sensitivity studies documented in BAW-10192-PA have been shown l to apply to the 177 fuel assembly, lowered loop plant design of TMI-1 (Reference 5). By virtue of the plant similarities between TMI-1 and ANO-1, these studies also apply to the ANO-1 plant. FTl also perfom cd the necessary plant-specific sensitivity studies to confirm that the most llmiting set of plant boundary conditions were applied to the licensing analyses for both large and small LOCAs (References 3,6,8,13,17 and 32). l Finally, after completing these studies, the required spectrum of ECCS analyses of the limiting break sizes were performed at a variety of core operating parameters that envelope the entire range of core operation (References 6 and 32). l The results of the revised LBLOCA and SBLOCA analyses for the TMI-1 and ANO-1 plants are described in this document. Section 2 summarizes the results of the LBLOCA and SBLOCA analyses. Parameters and input assumptions used in the' calculations are provided in Section 3. Sections 4 and 5 discuss specific aspects of the large and small break LOCA analyses, respectively. Section 6 summarizes the compliance with the acceptance criteria of 10CFR50.46. Finally, Section 7 identifies EM SER restrictions that have been met or must be monitored for the revised analyses. 6
r Framatome Technologies. Inc. 86-5002073-02 4
- 2. Summary of Results 10CFR50.46 specifies that the emergency core cooling system for a commercial nuclear power plant must meet five criteria:
- 1. The calculated peak cladding temperature (PCT) is less than 2200 F.
- 2. The maximum calculated local cladding oxidation is less than 17.0 percent.
- 3. The maximum amount of core-wide oxidation does not exceed 1.0 percent of the fuei cladding.
- 4. The cladding remains amenable to cooling.
- 5. Long term cooling is established and maintained after the LOCA.
The large and small brea'K LOCA calculations documented in References 6 and 32 with the approved RELAPS/ MOD 2-B&W evaluation model (Reference 4) demonstrate compliance with these criteria for breaks up to and including the double-ended severance of the largest primary coolant pipe for the ANO-1 and TMI-1 plants. The 2 spectrum also included a 0.44-ft core flood line break and a 0.02463-ft2 high pressure injection line break. An initial core power level of 102 percent of 2772 MWt was modeled with an axial peaking factor of 1.7. l Large break sensitivity studies and break spectrum studies performed with the RELAPS-based EM for the 177-FA lowered loop plant show that the double-ended guillotine break at the cold leg pump discharge (CLPD) with a discharge coefficient of 1.0 and minimum ECCS injection is the limiting case (Reference 5). Table 2-1, Figure 2-1, and Figure 2-2 summarize the allowed linear heat rate (LHR) limits that were determined for five core axial power shapes and two times in life, plus one analysis at the maximum core burnup. Each analysis determined the maximum LHR limit that could be achieved while still meeting the criteria of 10 CFR 50.46. These limits are extended throughout the core by linear interpolation between elevations. The limits below 2.506 feet are reduced linearly to 0.95*LHR 2. sos at 0.0 feet. The limits above 9.536 feet are reduced linearly to 0.95*LHR9 338 at 12.0 feet. These extrapolations were determined to be appropriate for Mark-B9 assemblies in Reference 17. Steady-state and transient energy deposition factors specific to the time in life were used for the hot channel. The full power linear heat rats limits can be maintained for partial-power and three-pump operation as long as the allowable moderator temperature coefficient (MTC) as a function of full power shown in Figure 2-3 is preserved. These limits are valid for a plant average steam generator tube plugging of 20 percent or less with a maximum of 25 percent plugged in a single loop. The full sequence of events and analytical resuits for each LBLOCA case analyzed are provided in Table 4-2, Table 4-3 and Table 4-4 in Section 4.2. 7
Framptome Technolooies. inc. 86-5002073-02_ Small break sensitivity studies performed with the RELAP5-based EM for the 177-F lowered loop plant show that the most limiting break location is in the bottom of th , leg piping between the reactor vessel inlet nozzle and the HPl nozzle (Referen Two special break cases, an HPiline break and a CFT line break, were also an A peak linear heat rate of 16.8 kW/ft at the 9.536-ft elevation was chosen to maximize the heatup for those cases that experience core uncovering. Twelve fuel assemblies were the grouped average into the hot channel with the remaining 165 fuel assemblies grouped into channel. Steady-state and transient energy deposition fears of 0.973 were used. Table 2-2 end Figure 2-4 summarize the results of the Mark-89 small break i LOCA calculations performed for a spectrum of cold leg pump discharge breaks, an line break, and a CFT line break. The full sequence of events and analytical results fo each 5.2. SBLOCA case analyzed are provided in Table 5-2 through Table 5-5 in Section i I 8
Framatome Technoloaies. Inc. 86-5002073-02 Table 2-1: Summary of Mark-B9 LHR LBLOCA Limits LBLOCA 20% Tube Plugging (Reference 32) Limits Based on Nuclear Source Power LOCA LHR Limit, kW/ft LOCA LHR Limit, kW/ft LOCA LHR Limit, kW/ft (PCT, F) (PCT F) (PCT. F) Elevation BOL MOL (40 GWd/mtu) EOL (60 GWd/mtU) 0.0 ft 15.9 (<2059) 15.9 (<1855) 11.6 (<1535)
' 2.506 ft 16.8(2059) 16.8(1855) 11.6 ( 1535) 4.264 ft 16.8(2030) 16.8(1883) 11.6 (<1700) 6.021 ft 17.0 (2036) 17.0 (1864) 11.6 (<1700) 7.779 ft 16.8(1964) .
16.8(1864) 11.6 (<1700) 9.536 ft 16.8 (2083) 16.8 (1935) 11.6 (<1700) 12.0 ft 15.9 (<2083) 15.9 (<i935) 11.6 (<1700) Notes: 1. The LHR limits presented above represent the power generated by the pin (i.e. all sources of useable energy caused by the fission process).
- 2. Analyses at BOL and MOL used a steady state energy deposition factor (EDF) of 0.973 for initial core energy deposition and a transient EDF of 1.0. Analyses at EOL used a steady state EDF of 1.0 and a transient EDF of 1.1. For further information regarding the use of EDFs, see Section 4.2.4.1.
3, Linear interpolation for LHR limits is allowed between 40000 mwd /mtU and 60000 mwd /mtu. j
- 4. The LHR limits below 2.506 feet are reduced linearly to 0.95'LHR2 son at 0.0 feet. The LHR limits above 9.536 feet are reduced linearly to 0.95'LHRs s3s at 12.0 feet. The EOL LHRs are maintained constant at 11.6 kW/ft for all elevations.
- 5. LHRs are valid for fuel enrichments of 5.1 weight percent (maximum) and pin propressures of 355 psia.
- 6. The fCL and EOL PCTs were not explicitly calculated. The PCTs from Reference 17 were conservatively increased by the calculated BOL PCT penalty for the corrected hot channel reflood grid data (Reference 32).
l l l l l I 9
Framatome Technoloaies. Inc. 86-5002073-02 Figure 2-1: TMI/ANO 20% Tube Plugging LBLOCA LHR Limits at 2772 MWt wit.h Burnup ie
! e t
if . k.+-..-.'....-... , i . e i I f q 16 : .e . e >
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LHRs frenartled as nuclear source pows, ' 8 12
.+.0A 124 (15 9 hW/ft) i . , H e ten a scL 1
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! i i
. *-6 0214 (17 0 kW/ft) ,
11 0 10000 20000 30000 40000 50000 60000 70000 Burnup, mwd /mtU Figure 2-2: TMI/ANO 20% Tube Plugging LBLOCA LHR Limits at 2772 MWt with . Elevation ' 18 e 4 8 9 4 37 _ 16 , --
..-.....q.....
" 15 .
i I g i 6 4 4 % i i . p I \
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0 2 4 6 8 10 12 Dovetten (ft) 10 d
Framatome Technoloaies. Inc. 86-5002073-02 1. Figure 2-3: MTC Limit versus Power Level
} -+-Recommendek kNk i e RELAPSMowable 10 . _ - - - . _. ..a-.-x== -.
3< -. g - - . . - , 1 7 .. 6- - 4
;. -4 N
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l 0- --. - : ; O 20 40 '60 80 100 120' Percent Fun Power l
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Framatome Technoloaies. Inc. 86-5002073-02 Table 2 2: Summary of Calculated PCTs for Mark-B9 SBLOCA Analyses BWOG SBLOCA 20% Tube Plugging (Reference 6) Break Size PCT Time of PCT (ft') (F) (s) 0.75 860 98 l 0.50 845 150 0.30 790 270 0.15 922 688 l 0.10 1334 1045 J 0.09 1354 1267 0.08 1375 1480 0.07 1331 1577 0.06- 1357 1818 0.05 1412 2550 0.04 1361 3314 0.03 1287 5126 0.01 715 0.0 0.44 (CFT Line) 715 0.0 0.02463 (HPI Line) 1299' 5036 2 0.02463 (HPl Line) 1297 5570
- 1. Without manual HPl actuation.
- 2. With r:anual HPl initiation and letdown isolation at 10 min after LSCM.
Figure 2-4: SBLOCA SpectnJm, PCT versus Break Size
'5 i
i -+-CLPD Spectrum 1400 -j--~ -~ !
-e.CFT Une Break '---
-e-HPl Line Broek 1300 -4 ,
i l E 1200 - -~~L ! ! ! I l 1100 -
* , i if C --
o I i f . 900 800 - -. v i 700 l I i 600 0 0.1 0.2 0.3 0.4 0.5 06 0.7 0.8 Emak Size (ft2) 12
Framatome Technoloales. Inc. - 86-5002073-02 l
- 3. Plant Parameters and inputs i This section provides a u ief summary of the specific plant parameters and inputs ,used in the LBLOCA and SBLOCA analyses. Most of the plant parameters and inputs are discussed in detailin Reference 7.
3.1 LBLOCA Analyses The inputs and assumptions used in the Mark-B9 LBLOCA analyses are identified in Table 3-1 through Table 3-7 and in Figure 3-1. l 3.2 SBLOCA Analyses The inputs and assumptions used in the Mark-B9 SBLOCA analyses are summarized in Table 3-8 through Table 3-15. l l 1 I I l Il l 13
r i Frapatome Technoloaies. Inc. 86-5002073-02 1 Table 3-1: LBLOCA inputs and Assumptions for Burnups of 0-60 Gwd/mtU Parameter LBLOCA Value Core Power (MWt) 1.02*2772=2827.44 Pump Power (MWt/ pump) 5 (Note 3) l Primary Side Ave Fluid Temp. (F) 579 Cold Leg Fluid Temp. (F) 552 Hot Leg Fluid Temp. (F) 606 RCS Press. @ HL Tap (psia) 2170 Total RCS Mass Flow Rate (Ibm /hr) 133.9x10' Core Bypass Flow (percent) 7.50 indicated Pressunzer Level (in) 220. (on 400 scale) Main Feedwater Temperature (F) 465. SG inventory (Ibm /SG) -52,000 (Note 4) Main Feedwater Flow per SG (Ibm /sec) - 1700 (Note 4) Main Feedwater isolation LOOP + 2 s constant flow + 12 s linear coastdown Turbine Header Pressure (psia) 875 (Note 4) Turbine Tnp Delay Time (s) 0.5 Turbine Stop Valve Stroke Time (sl 0.5 Turbine Stop Valve isolation (s) LOOP + 0.5 s MSSV Capacity (Ibm /s) 486 @ 1065 psia Steam Generstor Tube Plugging 25% BL /15% IL ESAS Setpoint (psia)(HPI) 1495 (Note 5) l Low-Low ESAS Setpoint (psia)(LPl) 355 ECCS (LPI) Delay Time (s) . 35 s after HPl setpoint and 10 s after LPl setpoint. l BWST Liquid Temperature (F) 120 Containment Pressure Figure 3-1 (Note 1) CFT Liquid Volume (ft3/ tank) 985 CFT Liquid Temperature (F) 140 CFT Gas Pressure (psla) 580 CFT Surge-Line K-Factor 6.926 Volume of CFT Surge Line (ft') 40.0 Elev. Change of CFT Surge Line (ft) 18.3 LPI Flow Table 3-3 Reactor Coolant Pump Tnp OnLOOP SS Fuel Pin Energy Deposition 0.973 (1.0 @ EOL) Transient Fuel Pin Energy Deposition 1.0 (1.1 @ EOL) (Note 2) Decay Heat 1.2 times ANS 1971 standard Actinide Coefficients B&W Heavy isotopes Doppler Coefficient (Ak/k/F) -1.69x10* (dji420 F Delayed Neutron Fraction (p.n) 0.007102 Prompt Neutron Generation Time (s) 0.248x10' Operator Actions Table 3-2 N.)tes 1. The RB spray flow used to generate the containment pressure response was 1500 gpm (see Section 4.3), which is below the maximum flow rate for the ANO-1 unit of 1800 gpm.
- 2. The transient energy deposition factor does not correspond to the EDF specified in the AIS (Ref. 7).
See Section 4.1 for a discussion of the effect on the results.
- 3. The pump thermal power is not specifically an input value, but is calculated based on the pump component input and the duid properties to be approximately 5 MWt/ pump.
- 4. The MFW flow, SG inventory and turbine header pressure were adjusted during steady-state to obtain the appropriate heat balance and T.g.
- 5. HPl flow versus pressure is not explicitly analyzed, however it is considered for calculation of minimum containment pressure (using maximum ECCS injection), momentum loss at ECCS injection from steam-water interaction and long-term cooling.
14
Framatome Technoloaies. Inc'. 86-5002073-02 Table 3-2: Operator Actions for LBLOCA Appropriate operator actions to mitigate a LBLOCA event are detailed in the Generic Emergency ' Operating Guidelines (Reference 35) or as provided in the plant-specific EOP. The following specific I operator actions are assumed in the analyses: Trip RCPs within 2 minutes of LSCM if there is no loss-of-offsite-power. Verify proper ES operation and alignment of HPl and LPl. Transfer ECCS suction source from the BWST to the sump for long-term cooling. Maintain appropriate core boron concentration control to prevent boron precipitation. Table 3-3: LBLOCA Doppler Reactivity Fuel Temperature Doppler Reactivity (F) (S) 100.0 3.1411 l 1420.0 0.0 3500.0 -4.9496 l I Table 3-4: LBLOCA LPI Flows l Pressure Flow (psig) (gpm) 0 3150 109 2700 145 1830 160 1350 ; 169 900 ; 178 0 15
Framatome Technolooies. Inc. 86-5002073-02 Table 3-5: LBLOCA Moderator Dens 8ty Reactivity inputs Reactivity Density ($) (Ibm /ft )
-63.36 0.0 l
-38.44 10.089
-24.50 13.452
-12.25 17.936
-7.46 22.421
-4.51 26.905
-2.746 30.043
-2.197 31.389
-1.436 33.631
-1.169 35.873
-0.493 38.115
-0.380 40.357
-0.070 42.599 0.0 44.841
-0.070 49.325 l -0.422 53.809 l -1.690 62.777 l
-1.690 89.682 Note: The moderator reactivity at 100% power corresponds to $0.0 l 16
Framatome Technoloaies. Inc. 86-5002073-02 Table 3-6: Mark-B9 Radial Power Profile History Fuel Rod Burnup Linear Heat Rate (mwd /mtU) (kW/ft) 0 10.93 299 10.38 9999 10.31 11999 10.25 13999 10.17 ) 18999 10.04 j 20999 10.11 1 23999 9.90 27499 9.76 30999 9.56 34999 9.36 37999 9.22 41299 _ 8.67 47999 8.20 i 51999 7.99 62000 7.99 L i l 1 l 17
i Framatome Technoloaies. Inc. 86-5002073-02 Table 34: Containment Assumptions - LBLOCA Minimum Backpressure Analysis Parameter Value initial containment pressure, psig -1.0 Initial containment temperature, F 110 Humidity, % 100 Outside ambient temperature, F 40 Containment free volume, ft' 2.205x10* Paint thickness, mils 10.0 ECCS injection maximum HPl injection through spray 1 HPI train at runout condition RB areas, thicknesses Generic values used in Reference 29 are listed on Table 6.6-5 of TMI-1 FSAR. Thermal conductivities and heat Generic values as determined in Reference 29. capacities RB cooler performance curves 3 fan coolers RB cooler delay, sec 0.0 RB cooler water temperature, F 40 Maximum RB spray flow rate, gpm 1800 (used 1500, see Section 4.2.4.3) Number of RB spray headers 2 RB spray delay, sec 65.0 RB spray water temperature 40 18
Framatome Technoloaies~ Inc. . 86-5002073-02 4 Table 3-8: SBLOCA Inputs and Assumptions Parameter SBLOCA Value Core Power (MWt) 1.02*2772=2827.44 Pump Pcwer (MWt/ pump) 5 (Note 3) l Primary Side Ave Fluid Temp. (F) 579 Cold Leg Fluid Temp. (F) 552 Hot Leg Fluid Temp. (F) 606 RCS Press. @ HL Tap (psia) 2175 (Note 5) Total RCS Mass Flow Rate (Ibm /hr) 134.4x10'(Note 5) Core Bypass Flow (percent) 7.5 indicated Pressunzer Level (in) 220. (on 400 scale) Main Feedwater Temperature (F) 465. Main Feedwater Flow per SG (Ibm /s) -1700 (Note 4) SG Inventory (Ibm /SG) -52,000 (Note 4) Main Feedwater isolation LOOP + 2 s constant flow + 12 s linear coastdown Turbine Header Pressure (psia) 858 (Note 4) Turbine Trip Delay Time (s) 0.5 Turbine Stop Valve Stroke Time (s) 0.5 MSSV Capacity (ibm /s) 486 @ 1065 psia Sec. Side Cooldown/ Blowdown vic ADVs Not Used. l Steam Generator Tube Plugging 25% BL /15% IL EFW Flow (gpm) 200 EFW Temperature (F) 120 EFW Delay (s) 120 SG Level Control (Note 1) EFW to 50% , then raise to 290* at LSCM+20 min ESAS Setpoint (psia)(HPI) 1495 HPl Delay Time (s) 35 s after HPl setpoint Low-Low ESAS Setpoint (psia)(LPI) 350 LPI Delay Time (s) 35 s after HPl setpoint and 10 s after LPI setpoint. l BWST Liquid Temperature (F) 120 CFT Liquid Volume (ft'/ tank) 985 (Note 2)(895 for CFT break) l CFT Liquid Temperature (F) 140 CFT Gas Pressure (psia) 580 (Note 2) (650 for CFT break) l CFT Surge-Line K-Factor 692.6 Volume of CFT Surge Line (ft') 40.0 Elev. Change of CFT Surge Line (ft) 18.3 LPI Flow Table 3-10 Containment Pressure (psia) 70 LHR Limit at 9.536 ft 16.8 Transient Energy Deposition Factor 0.973 Decay Heat and Actinide 1.2 times ANS 1971 standard, with B&W Actinides Reactor Trip (psia) 1795 Reactor Trip Delay (s) 0.6 RC Pump Trip On LOOP Doppler Coefficient (Ak/k/F) -1.69x10* @1420 F Delayed Neutron Fraction (b) 0.007102 Prompt Neutron Generation Time (s) 0.248x10* Scram Curve Table 3-11 Control Rod Drop Time 1.4 see to 2/3 insertion l Fullinsertion Rod Worth (%Ak/k) -2.26 Operator Actions Table 3-9 l 19
l Framatome Technoloales. Inc. 86-5002073-02 Notes on Table 3-8
- 1. The analysis modeled 65 percent (290") on the operate range, however TMI-1 and ANO-1 must maintain the secondary SG level at the LSCM tetpoint (above the RCP spillover elevation) to maintain a condensing surface for the smallest SBLOCAs tnat indicate LSCM or actuate ESAS but are not explicitly analyzed. For TMi-1 the RCP spillover elevation is 23.58 ft above the upper face of the lower tube sheet, which is equivalent to 62% on the OR. For ANO-1 the RCP spillover elevation is 24.35 ft, which is equiva!ent to 65% on the OR. Therefore, the secondary SG level must be maiMained above 62% and 65% for TMI-1 and ANO-1, respectively.
- 2. A sensitivity study was performed that determined that a maximum CFT liquid volume and minimum pressure produces more conservative results for the SBLOCA, except for the CFT line break which uses a minimum CFT liquid volume and maximum pressure. This is discussed in Section 5.1.2.
- 3. The pump thermal power is not specifically an input value, but is calculated based on the pump component input and the fluid properties. The value was calculated by the computer code to be approximately 5 MWt/ pump.
- 4. The MFW flow, SG inventory and turbine header pressure was adjusted during steady-state to obtain the appropriate heat balance and Tug.
- 5. The SBLOCA steady-state initialization was developed independent from the LBLOCA case.
Therefore, the SBLOCA final steady-state RCS mass flow rate and pressure are slightly different than that reported in Table 3-1 for LBLOCA. however they are still representative of appropriate steady-state conditions. Table 3-9: Operator Actions for SBLOC,i Appropriate operator actions to mitigate a SBLOCA event are detailed in the Generic Emergency Operating Guidelines (Reference 35) or as provided in tne plant-specific EOP. The following specific operator actions are assumed in the analyses: Trip RCPs within 2 minutes of LSCM if there is no loss-of-offsite-power. Verify proper ES operation and alignment of HPl and LPl. (ANO-1 ONLY) Balance the high flow HPl line to within 20 gpm of the next highest flow line at 10 minutes after ESAS for a HPI line break. Raise SG level to LSCM setpoint with EFW at 20 minutes after LSCM and control at that level. For those smallest breaks that are not explicitly analyzed (partial HPl line and CLPD < 2 0.01 ft ), manual initiation of HPl is assumed at 10 minutes after LSCM to assure that the consequences of these breaks are less severe than those break sizes that are explicitly analyzed. Transfer ECCS suction source from the BWST to the sump for long-term cooling. Maintain appropriate core boron concentration control to prevent boron precipitation. I 20
l Framatome Technoloaies. Inc. 86-5002073-02 j P Table 3-10: SBLOCA HPI Flow Rates - CLPD Break l RCS Pressure intact Leg Flow Broken Leg / Spill Flow (psig) (gpm) (gpm) 0 306.6 131.4 600 301.7 129.3 i l 1200 266.0 114.0 l 1500 242.2 103.8 1600 234.5 100.5 1800 217.0 93.0 2400 133.0 57.0 Note: After ESAS is actuated, the RCS pressure remains below 2400 psig. l Table 3-11: SBLOCA HPl Flow Rates - HPI Line Break RCS Pressure intact Leg Flow Broken Leg / Spill Flow (psig) (gpm) (gpm) Before 10 Minutes 0 272.09 191.11 600 272.09 191.11 1200 197.97 ,207.00 1500 156.56 217.97 1600 138.96 223.67 , 1800 104.22 233.25 After 10 Minutes 0 308.96 154.24 l ; 600 308.96 154.24 I 1200 230 150 1500 182 164 , 1600 165 170 1600 130 180 l Note: After ESAS is actuated, the RCS pressure remains below 1800 psig. l < 21
]
l Framatome Technoloains. Inc. 86-5002073-02 I l j Table 3-12: SBLOCA HPl Flow Rates - CFT Line Break RCS Pressure intact Leg Flow Broken Leg / Spill Flow (psig) (gpm) (gpm) 0 438 N/A 600 431 N/A 1200 380 N/A 1500 346 N/A 1600 335 N/A 1800 310 N/A 2400 190 N/A Note: After ESAS is actuated, the RCS pressure remains below 2400 psig. l Table 3-13: SBLOCA LPI Flows Pressure Flow (psig) _ (gpm) 0 3150 98 2700 133 1830 148 1350 157 900 1 163.9 0 , j i l 22 l
1 1 Framatome Technoloaims. Inc. 86-5002073-02 Table 3-14: SBLOCA Scram Curve Time Reactivity l (s) (%AK/K) 1 0.0 0.0 l 0.2 0.58 0.3 0.99 I 0.4 1.83 l 0.6 5.29 0.8 12.33 1.0 21.41 1.2 33.09 1.4 50.75 1 1.6 72.96 1.8 91.30 2.0 99.26 2.2 99.99 2.3 100.0 10.0 100.0 Table 3-15: SBLOCA Moderator Density Reactivity inputs l ! Ak/k % Density
-0.4500 0 i
-0.2730 22.5 l l
-0.1740 '
30
-0.0870 40
-0.0530 50
-0.0320 60 '
-0.0195 67
-0.0156 70
-0.0102 75 l
-0.0083 80
-0.0035 85
-0.0027 90
-0.0005 95 0.0 100
-0.0005 110 l
-0.003 120 1
-0.012 140
-0.012 200 Note: The moderator reactivity at 100% power corresponds to 0.0 ak/k.
23 i i
Efamatome Technoloales. Inc. 86-5002073-02 Figure 3-1: LBLOCA Minimum Containment Pressure Response z
, -f i . . . . ]
i 3 .- -
- - - ,- d- .-- . .--,---,--- -- - -,-....r- --
t t t $ a 4 e i R i l l l 0 0 % $ $ f 25 - - ---a -
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e s i e i t i I 9 4 4 4 1 4 h 4 1 3 1 0 10 20 2 40 50 80 70 20 90 100 110 120 130 140 150 Tarts, e j 24
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l Framatome Technoloaies. Inc. 86-5002073-02
- 4. LBLOCA Sensitivity Studies and Analyses The LBLOCA sensitivity studies and analyses are presented in the following sections. A
. sequence of events for the LBLOCA analyses are presented in Table 4-1 through Table 4-4.
4.1 LBLOCA Sensitivity Studies LBLOCA snalyses require that various sensitivity studies be performed with the evaluatica inodel to demonstrate model convergence and to identify the most limit?ng set of boundar/ conditions or break locations that should be used in demonstrating j compliance with the five criteria in 10 CFR 50.46. As part of the LBLOCA EM, FTl l performed numerous LBLOCA sensitivity studies to confirm modeling techniques and methods.' Although the EM was based on a slightly different plant design, the safety evaluation report for BAW-10192-PA supports the application of the EM to the 177-FA plants. FTl has determined that the generic LBLOCA sensitivity studies performed in the EM are directly applicable to and appropriate for use in the ANO-1 and TMI-1 analyses. FTl also performed the necessary plant-type specific sensitivity studies to confirm that tb most limiting set of plant boundary conditions were applied to the licensing , anages. Section 4.1.1 provides a discussion of the generic sensitivity studies that have been applied from the reference EM report (Reference 4). Section 4.1.2 describes the plant-type specific studies. Section 4.2.1 identifies the LBLOCA base model used for the LBLOCA LHR limit analyses. 4.1.1 LBLOCA Evaluation Model Generic Sensitivity Studies The majority of the LBLOCA sensitivity studies presented in the EM topical report (Reference 4. Volume 1) are generic and apply to any LBLOCA analysis for the B&W-
. designed nuclear steam system. An example is the RELAPS/ MOD 2-B&W time-step studyf which showed that the automatic time step selection in RELAP5/ MOD 2-B&W would produce converged results. This demonstration need not be repeated for plant-specific applications in which the modeling techniques used are represented by those in the EM studies. The following is a listing of the sensitivity studies considered to be i generic with a discussion of why the conclusions of the study are applicable to this l
' LBLOCA applications report. For convenience, each discussion is referenced to the l section in the EM topical report where the study is documented. i
- 4.1.1.1 RELAP5/ MOD 2-B&W Time-Step Study The study using the generic EM, documented in BAW-10192-PA, Volume I, Appendix A, Section A.2.1, verified that, for light water reactor geometry, the RELAP5 time-step controller governs the code solution sufficiently to assure convergent results. In 25 m __
r Framatome Technoloales. Inc. 86-5002073-02 RELAPS/ MOD 2-B&W, the user specifies a maximum time step that can be modified I internally by the code in the event of convergence or Courant limitations. The LBLOCA EM time-step studies justified use of a 2.5-millisecond maximum time-step size for the first two seconds of the transient and a 25-milisecond maximum time-step size thereafter as appropriate for B&W-plant LBLOCA analyses. The EM controls the plant input models such that no significant deviation in the number or size of the control volumes or hea structures critical to the model results can be included between piant designs. Since the LBLOCA analytical model is similar to the model used for the EM ; time-step study, and the maximum time-step size is the same in the ANO-1/TMI-1 j LBLOCA analyses as in the EM time-step study, then the RELAP5/ MOD 2 time-step ) controller will also adequately control the problem advancement for these applications. The EM study remains valid, therefore, and this study does not have to be repeated. 4.1.1.2 RELAPS/ MOD 2-B&WPressurizerLocation Study Previous configuration studies performed with the LBLOCA EM (BAW-10192-PA, Volume I, Appendix A, Section A.2.2) showed that there is little difference in results when the pressurizer is connected to the broken loop instead of the intact loop. This result is expected since the LBLOCA transient is dominated by such factors as leak flow and initial fuel stored energy. Therefore, the pressurizer location study performed with the EM is applicable to the ANO-1/TMI-1 LBLOCA analyses and need not be repeated. 4.1.1.3 RELAP5/ MOD 2-B&WBreak Noding Study This study (BAW-10192-PA, Volume I, Appendix A, Section A.2.3) verified that hydraulic stability is achieved by providing at least one control volume in the pipe between any l adjacent component and the break node and by maintaining an L/D near 4.0 in the break control volumes. The calculated UDs for the LBLOCA model are 2.8. These values are below the recommended limit of 4.0, but greater than the minimum value of 1.5 suggested by the benchmarks to the Marviken Tests (Reference 18). Therefore, the break noding study performed with the EM is applicable to the ANO-1/TMI-1 LBLOCA analyses and need not be repeated. 4.1.1.4 RELAP5/ MOD 2-B&WCore Crossflow Study Core crossflow is modeled in the base model through the use of RELAPS/ MOD 2-B&W crossflow junctions between the hot and average channels in the core region. This study (BAW-10192-PA, Volume I, Appendix A, Section A.2.4) verified that a crossflow K-factor of 72.0 in a B&W-type reactor produced converged results and is reasonable l- for two-channel EM applications. The study is dependent only on the basic aspects of the fuel design, which did not change for this evaluation. Therefore, the stud;es performed for the EM remain applicable and do not need to be repeated, 26 L ;
Framatome Technoloaies. inc. 86-5002073-02 4.1.1.5 RELAP5/ MOD 2-B&W Core Nading Study In conjunction with the core crossflow study, this study-(BAW-10192-PA, Volume I, Appendix A, Section A.2.5) verified that modeling the reactor core with two channels adequately predicted the blowdown transient. As the basic core arrangement and fuel design are not altered across the range of designs to be considered, the results of the study are applicable to all plants considered by the evaluation model. Therefore, the study is applicable to the ANO-1/TMI-1 analyst s. 4.1.1.6 RELAPS/ MOD 2-B&WECCS Bypass Study This study (BAW-10192-PA, Volume I, Appendix A, Section A.2.8) verified a non-mechanistic bypass model based on Upper Plenum Test Facility (UPTF) test results to remove the ECCS liquid injected during blowdown. This study is applicable to all plants ' with dcwncomer injection and reactor vessel vent valves. Therefore, the study is applicable to the ANO-1TTMl-1 analyses. 1 4.1.1.7 REFLOD3B Loop Noding Study This study (BAW-10192-PA, Volume I, Appendix A, Section A.3.1) verified the noding detail used in_ the REFLOD3B code (Reference 19). It is applicable to all plants censidered by the evaluation model. A minor change from the EM noding arrangement was included in this lowered-loop noding arrangement. The intact cold legs were combined in the EM, but were separated for this application to accommodate a single i blocked loop seal if predicted. The analyses performed for ANO-1/TMI-1, however, did not predict any locp seals. Although this is a minor modeling change, the spatial details were preserved. Since this change is inconsequential, the results of the EM study are applicable to the ANO-1/TMI-1 analyses. 4.1.1.8 REFLOD3B RCP Locked Versus Free-Spinning Rotor Study This study (BAW-10192-PA, Volume I, Appendix A, Section A.3.2) showed a considerable reduction in flooding rate under a locked-rotor assumption. The study affirms the generally held understanding of loop resistance effects on reflooding rates and is applicable for all plant types covered by the evaluation model. Therefore, the study is applicable to the ANO-1/TMI-1 analyses. 4.1.1.9 BEACH Time Step Study. This study (BAW-10192-PA, Volume 1, Appendix A, Section A.4.1) verified that the BEACH (RELAPS/ MOD 2-B&W) time-step controller would check and adjust time step size sufficiently to assure converged results provided the set of inputs described as the 27
p
' Framatome Technoloaies. Inc.- 86-5002073-02 1
" Decrease'd Time Step" case on Table A-10 of BAW-1019.' 'A is used. In response to
' NRC Question 16 on the evaluation model (BAW-10192-P.' , Volume 111), a reanalysis of
.the BEACH . time-step study was performed with the BEACH inlet' subcooling methodology. The results of the revised study also confirm th?t the time-step inputs ;
given in Table A-10 of- Volume 1 of BAW-10192-PA produce converged results. ! Alternate system designs within the range of designs covered by the evaluation model l will not change these results. Therefore, the study is applicable to the ANO-1/TMI-1 l analyses.
' 4.1.1.10 BEACH Axial Fuel Segmentation Study j l
This study (BAW-10192-PA, Volume I, Appendix A, Section A.4.2) verified that the use of eight fine-mesh intervals 'was sufficient to produce converged results. Alternate system designs within the range of designs covered by the evaluation model will not change that result. Therefore, the study is applicable to the ANO-1/TMI-1 analyses. 1 4.1.1.11 Axial Versus Radial Core Peaking Factor Study I This- study (BAW-10192-PA, Volume I, Appendix A, Section A.5) showed that
- representative LOCA limits were obtained with a method that specifies a constant axial peak of 1.7 and adjusts the radial peaking factor to give the maximum allowable linear heat rate limit. Typical core maneuvering analyses obtain radial and axiai peaking factors similar to those used in the EM. Therefore, FTl views this technique to be reasonable for all EM applications. The NRC has imposed a restriction on this method.
. See Section 7 for further information. > ~ 4.1.2 LBLOCA Evaluation Model Plant-Specific Sensitivity Studies ~ 'Although a considerable portion of the analysis inputs and assumptions are set or controlled by the evaluation model and its sensitivity studies, some parameters are ' dependent on inputs specific to a plant type and can only be established by separate studies. These studies are performed to identify a limiting case to use in calculating the LBLOCA LHR limits. This section presents the studies performed with the LBLOCA evaluation model for the Oconee 177-FA LL plants that helped to define the final plant model configuration used in the ANO-1/TMI-1 LOCA LHR limit analyses.
4.1.2.1 'RELAP5/ MOD 2-B&WPump Degradation Study This_ study was performed as part of the generic evaluation model sensitivity studies contained in' BAW-10192-PA (Volume -1, Appendix A, Section A.2.6). The results established a limiting, maximum pump degradation multiplier set (M1) to be used in all EM analyses. Preliminary Safety Concern (PSC) 1-99 (Reference 17) identified that the 28 3
Framatome Technoloaies. Inc. - 86-5002073-02 lowered-loop plants could produce significantly higher PCTs when a minimum two-phase pump degradation model is used (M3). The subsequent supporting analyses performed for the ANO-1/TMI-1 units confirmed this assertion (Reference 17). The results of the ANO-1/TMI-1 study clearly demonstrated that the minimum two-phase degradation produces more severe results than the maximum degradation case. The minimum doradation multiplier reduces the resistance of the pumps. As a result, the core flow revt ses direction later in the transient and produces lower core flow rates. The decrease in removal of fuel stored energy leads to fuel temperatures at end of blowdown that are approximately 6 F higher than for the maximum degradation case. Furthermore, there is less liquid available for input to REFLODSB in the lower plenum of the reactor vessel. As a result, the adiabatic heatup time will be longer resulting in a PCT increase of approximately 26 F. From these results it is concluded that for all 177-LL B&W plants, the minimum pump two-phase degradation will produce more severe results than the maximum pump degradation. 4.1.2.2 RELAP5/ MOD 2-B&WRC Pump Power Study in the evaluation model (BAW-10192-PA, Volume I, Appendix A, Section A.2.7), this study indicated that the RCS response with the pumps powered is less severe from a core cooling perspective than the configuration with the pumps unpowered. To confirm this pump configuration for the Oconee units, a pumps-powered analysis was performed (Reference 28). The results of the Oconee study clearly demonstrated that the pumps-tripped case produces more severe results than the pumps powered case. With the pumps powered, the core flow was n' ore positive in the first few seconds of the blowdown because the pumps produced h'gher loop flows. During this first portion of blowdown, the increase in the core flow allows for removal of additional fuel stored energy, decre, ping end-of-blowdown fuel temperatures 50 to 90 F. Also, in the pumps-powered case. 'tre liqW was available for input to REFLOD3B in the lower plenum of the reactdr vessel, so the adiabatic heatup time will be less. From these results it is concluded that for all 177-LL B&W plants, the pumps-tripped configuration will produce more severe results than the pumps-powered configuration. Therefore, it is not necessary to demonstrate these results for the ANO-1/TMI-1 application analyses. 4.1.2.3 Containment Pressure and ECCS Configuration Study The results of Volume I, Appendix A, Section A.10 of BAW-10192-PA recommendeo that this study be performed for each plant classification for specific LOCA applications studies. The base case for the Oconee study assumed maximum ECCS injection and a corresponding minimum containment backpressure based on the results of BAW-10192-PA. In addition to the base case, this study considered both maximum (two trains) and minimum (one train) ECCS injection with a corresponding containment pressure (Reference 20). 29 1
Framatome Technoloaies. Inc. 86-5002073-02 The first step was to complete a CONTEMPT (Reference 21) maximum ECCS injection / minimum containment pressure analysis with the mass and energy release from the base blowdown model and from a REFLOD3B analysis that included flow from two ECCS trains. A minimum containment pressure was obtained by incorporating the
- assumptions identified in 'Section 4.3.6.1 of BAW-10192-PA. The CONTEMPT containment pressure was provided to the REFLOD3B analysis as a new boundary condition from which an iteration was performed. The mass and energy release from REFLOD3B was compared with the original input to CONTEMPT for convergence.
Convergence was obtained after one iteration. A similar technique was used for a maximum containment pressure / minimum pumped ECC injection case. The maximum containment pressure was obtained by changing the containment volume and areas of the building heat sinks to the nominal values identified in the FSAR. The results demonstrated that the minimum containment pressure cases produce PCTs about 20 F higher than the maximum containment pressure case at the ruptured elevation. This difference in PCT is related to the location of rupture. The minimum pressure cases ruptured at the peak power location, while the maximum pressure case ruptured at the elevation below the peak power. The small difference in the PCT is attributable to the small difference in power between these two elevations. This is evident because the carryout at the time of the peak is the same for the cases. The maximum containment pressure case produced a higher PCT at the unruptured elevation. This is due in pa.. to the difference in power levels between the peak power location and the elevation above it. But, more important, the carryout at this time in the maximum pressure case is significantly lower than the other two cases. The long-term effects of the minimum ECCS injection coupled with minimum containment pressure are the most severe of the three cases examined. The single train of ECCS injection is unable to keep the downcomer full, so the flooding rate is below two inches per second until after the average channel is quenched. This low flooding rate leads to a slow quench front advancement. The high inlet subcooling suppresses carryout. The whole-core oxidation increase reflects the higher fuel and clad temperatures. It is apparent from the results that the minimum ECC injection consistent with minimum containment pressure configuration produces the most limiting results considering both PCT and long-term metal-water oxidation effects. This configuration is applicable to all 177-LL B&W plant analyses. Therefore, it is not necessary to demonstrate these results for the ANO-1/TMI-1 application analyses. The minimum containment pressure response for the ANO-1/TMI-1 analyses was generated specific 6lly using the CONTEMPT inputs described in Table 3-7 and shown in Figure 3-1. ] 4.1.3 LBLOCA Break Soectrum Studv 10 CFR 50, Appendix K requires that a spectrum of breaks be considered in determining the worst-case break size, configuration, and location. Results of analyses i 30 u ___ _________
Framatome Technoloales. Inc. 86-500f073-02 performed using the previous EM (Reference 1) and the current EM (Reference 4) determined that the typical worst break is a full-area double-ended guillotine break with a discharge coefficient (CD) of 1.0 located in the CLPD piping. This break location causes a significant reduction in the core flow and fuel pin heat removal during the first third of the blowdown period. The proximity of the break to the ECC injection location also maximizes the potential for ECC bypass during the later stages of blowdown. These two effects result in less fuel pellet stored energy removal and an increase in the reactor vessel lower plenum re'ill time. To confirm these results for all 177-LL B&W plants, a break spectrum analysis, which considered break size, configuration, and location, was performed for the Oconee plants using the LOCA evaluation model (Reference 22). The results of these studies are summarized below. Discharae Coefficien+ Analysis - The case with a CD of 1.0 resulted in the smallest positive hot spot core flow between one and eight seconds of the blowdown phase. The smaller flow reduced the fuel pin surface heat transfer. The liquid mass remaining in I the lower plenum at the end of blowdown was also a minimum for this analysis, requiring a longer refill time during which the fuel pins heat up adiabatically. The calculated hot rod PCT of 1989 F was produced by the ruptured cladding segment. The calculated PCTs declined with decreasing discharge coefficient and switched to an unruptured segment, directly adjscent to the ruptured location. Further reductions in the discharge coefficieni would result in additional surface heat transfer that would continue ; to reduce the calculated PCT. Therefore, no other calculations with st" aller discharge ' coefficients were warranted. These results also confirmed that the tn:nsition break { sizos discussed in the LBLOCA EM did not need to be analyzed. The full-area, DEG j CLPD break with a discharge coefficient of 1.0 produced the most limiting results of the ' discharge coefficients studied. Since the results of this study can be applied to all 177-LL B&W plants, it is not necessary to demonstrate these results for the ANO-1/TMI-1 ) application analyses. l Break Tvoe Analysis - Appendix K requires that instantaneous double-ended guillotine ! and longitudinal spilt break configurations be considered. The guillotine break is ! modeled as an instantaneous severance of the pipe, allowing separate discharges ! through the full pipe area from each side with no mixing of the flows from the two sides of the break allowed. The split break assumes discharge from the pipe through an area ; up to twice the cross-sectional pipe area. M;xing at the break location is allowed. The blowdown rates and system flow splits are somewhat.different for the two break types, which can lead to differences in core flows and fuel pin heat removal. Both breaks use discharge coefficients of 1.0. The split break produced higher core downflows during the later portion of blowdown, leading to better cooling and lower end-of-blowdown fuel pin and clad temperatures. The lower pin temperatures produce less boiling, decreasing the upper plenum pressure and increasing the core flooding rate. Consequently, the calculated PCT for the full-area split break with a discharge coefficient of one is lower than that produced by the guillotine break. Split breaks performed with smaller discharge coefficients would increase the positive core flows during the first portion of blowdown. These higher flows would improve the 31
o Framatome Technolooies. Inc. 86-5002073-02 cladding heat removal and cause additional reductions in the calculated PCTs.
' Therefore, CLPD split breaks will not produce core thermal-hydraulic conditions that can result in a PCT higher than that calculated for the guillotine break with a discharge coefficient of one. Since the results of this study can be applied to all 177-LL B&W plants, it is not necessary to demonstrate these results for the ANO-1/TMI-1 application analyses.
Break Location Analysis - There are three locations to consider for the large break LOCA: the hot leg piping, the cold leg pump suction piping, and the cold leg pump discharge piping. The hot leg break has been consistently shown to result in peak cladding temperatures far below those predicted for cold-leg breaks (see BAW-10192-PA, Section A.6.5). The large positive core flow and no ECC' bypass combine to provide high fuel pin heat removal for all hot leg breaks. Therefore, a hot leg LOCA analysis is not required to demonstrate that a hot leg break is not limiting for the 177-LL plants. The pump suction break was analyzed to compare with the cold leg pump discharge break to determine the worst break-location. The broken leg pump provided a significant resistance to flow trying to reach the break through the broken leg (RV side). The liquid was forced to reach the break via the hot legs, leading to positive core flows ' throughout blowdown and significantly increased hot pin heat removal. The lower pin temperatures allowed a higher core flooding rate and faster quench front advancement, and the amount of liquid remaining in the reactor vessel at EOB led to a significantly shortened adiabatic heatup time. The PCT for the pump suction break was 160 F lower { than that for the pump discharge break. Therefore, a break in the CLPD will produce ' more severe results. Since the results of this study can be applied to all 177-LL B&W plants, it is not necessary to demonstrate these results for the ANO-1/TM101 application analyses. 4.1.4 CFT Initial Conditions Study-This study was not performed as part of the generic evaluation model sensitivity studies contained in' BAW-10192-PA. A study was performed (References 23 and 24), however, for the Oconee plants to investigate which combination of CFT initial pressure and liquid inventory was most conservative for use in the LBLOCA analyses being ! performed with the evaluation model. Four cases were included in the Oconee study: (1) minimum inventory with minimum pressure, (2) maximum inventory with minirrium pressure, (3) maximum inventory with maximum pressure, and (4) nominal inventory ; with nominal pressure. The results of the Oconee study showed that the maximum inventory with minimum ! pressure case produced the most conservative set of initial CFT conditions. These e
' initial conditions combine to produce the smallest initial gas volurns and mass. As the CFT empties, the nitrogen overpressure reduces more quickly, resulting in a lower CFT :
flow during the lower plenum refill or adiabatic heatup period. Because the PCTs at all l core elevations are ruptured-node limited and occur at approximately 30 seconds, the 32
I Framatome Technologies. Inc. 86-5002073-02 beneficial effects of more CFT liquid on long-term reflooding rates and the clad cooling l are not realized. l Two additional cases were examined in which the CFT liquid inventory was further i reduced below the current minimum value for the Oconee plants. The smaller inventories were combined with maximum initial CFT pressure to investigate the potential for a significant delay between the CFTs emptying and the LPI initiating. This scenario can allow the downcomer level to drop sufficiently such that the long-term flooding rates are lower and a higher PCT occurs later in the transient. The reduced inventories investigated were 920 ft and 860 ft 3(Oconee minimum inventory was 980 3 ft ). The 920-ft case resulted in a delay of 16 seconds t:etween the CFTs emptying and the initiation of LPI flow, but the PCT still occurred near the beginning of core l recovery and was less than the PCT calculated for the maximum inventory / minimum ; pressure case. Reducing the inventory further to 860 ft , however, produced a delay of l 18 seconds, which produced a PCT at approximately 70 seconds that approached the PCT calculated for the maximum inventory / minimum pressure case. The results of the Oconee analyses can be directly applied to the 20 percent tube , plugging analyses by confirming that the variation in the CFT liquid volumes and initial pressures have bee sufficiently considered in the Oconee study. The CFT initial pressure variation for TMI-1 and ANO-1 is almost identical to the Oconee plants and does not need to be considered. The nominal inventory, however, is 100 ft less than that considered for the Oconee units, providing a rnaximum volume of 985 ft3 and a minimum volume of 895 ft 3 (For ANO-1, the maximum CFT liquid volume is to be 3 3 revised to between 895 ft and 985 ft .) The maximum volume case is covered in the Oconee analyses, but the minimum inventory falls between the two values considered in the reduced inventory study. The parameter of interest is the delay between the CFTs emptying and the beginning of LPl. Since the 20 percent tube plugging analyses modeled a shorter ECCS de!ay time of 35 seconds, versus 48 seconds for Oconee, the delay between the CFTs emptying and LPI initiation would be on the order of 3 to 5 seconds instead of 16 to 18 seconds. Since the Oconee analyses showed that a delay of 16 seconds would still produce results bounded by the maximum inventory / minimum pressure case, the minimum CFT inventory with maximum CFT pressure results are also bounded by the maximum inventory /minimurn pressure assumption. Therefore, the Oconee study sufficiently covers the CFT initial conditions modeled in the analyses, and it is not necessary to redemonstrate these results for the ANO-1/TMI-1 analyses. 33
x Framatome Technoloaies. Inc. 86-5002073-02 4.2 LBLOC.A An' alyses at 2772 MWt The' LOCA analyses are performed to show compliance with 10 CFR 50.46 for the limiting core power and peaking conditions that are used to set core operational limits and trip setpoints (i.e.' the LOCA limits). These LBLOCA analyses serve as the bases for the allowable local power. Numerous cases are performed to determine a curve of allowable peak linear heat rate (LHR) as a function of core elevation for times in life of fuel . operation. This curve'is either contained in or referenced by the plant technical specifications. Plant operation is controlled such that the local peaking and power do not exceed these allowable LHR limit values. 4.2.1 Base Model The results of the evaluation model and plant classification sensitivity studies define the base model configuration for the ANO-1/TMI-1 LBLOCA LHR limit analyses. The base case is a full double-area, guillotine break in the cold leg pump discharge piping at the elevation of the reactor vessel inlet nozzle. A discharge coefficient of 1.0 is used to maximize the break flow. A loss of offsite power is assumed at the time of break opening, so the reactor coolant pumps and main feedwater pumps are not powered. The Westinghouse homologous head flow curves with RELAP5 two-phase head difference curves and head degradation using the M3 two-phase multiplier maximizes the PCT (minimizes core cooling during blowdown). Tube plugging in the steam generators is considered. The broken loop steam generator is 25 percent plugged and the intact loop is 15 percent plugged, for a plant-average 20 percent tube plugging. The non-mechanistic ECCS bypass method is used during blowdown to discard the ECCS liquid injection prior to predicting the end of bypass. The maximum time delay of 35 seconds after low (HPI) setpoint or 10 seconds after low-low (LPI setpoint is assumed to initiate pumped ECCS injection (LPI). HPl flow versus pressure)is not modeled LBLOCA transient analysis, however it is considered for calculation of minimum containment pressure (using maximum ECCS injection), momentum loss at ECCS injection from steam-water interaction and long-term cooling. For the refill and reflood system analysis, the reactor coolant pump rotors are assumed to be in a fixed position. The maximum ECC fluid temperature is assumed to minimize the core cooling potential. i Minimum (one train) ECCS with a minimum containme.nt pressure response was used to produce more conservative PCTs and whole-core hydrogen generation. The CFT initial conditions are set to maximum inventory and minimum initial gas pressure to
- assure a conservative calculation of the PCT. Additional plant conditions specific to the ANO-1/TMI-1 20 percent tube plugging analyses are summarized in Table 3-1.
Operator actions specifically considered in the LBLOCA analyses are listed in Table 3-2, however all actions specified in the plant specific EOPs should be performed to successfully mitigate the consequences of the LOCA. 34 e
E Framatome Technc wies. Inc. 86-5002073-02 4.2.2 -Transient Proaression 1 Large break loss-of coolant accidents can be treated analytically in three separate i phases: blowdown, refill, and reflood. The blowdown phase is characterized by the ! rapid depressurization of the reactor. coolant system to a condition nearly in pressure l equilibrium with its immediate surroundings. Core flow is variable and dependent on the nature, size, and location of the break. Departure from nucleate boiling (DNB) is calculated to occur very quickly, and core cooling is by a film boiling process. Since film I boiling amounts to only a -small fraction of the steady-state cooling, the cladding I temperature . increases by 600 F to 1200 F. CFT flow begins after the RCS depressurizes' below the CFT fill pressure. The condensation on the CFT liquid accelerates the negative core flows and reduces the fuel pin temperatures during the middle blowdown period. During the last phases of blowdown, cooling is by convection to steam, and the cladding temperature begins to rise again. Following blowdown, a period of time is required for the CFTs to refill the bottom of the reactor vessel, before final core cooling can be established. During this period, core cooling is marginal, and the cladding experiences a near-adiabatic heatup. This period is designated as the refill phase. Wnen the CFT water reaches the bottom of the core, the reflood phase begins. Core cooling is by steam generated below the rising core water level. The cladding temperature excursion is generally terminated before a particular elevation is covered by water since the steam-water mixture is sufficient to remove the relatively low decay heat power being generated at this time. The core is , eventually covered by a two-phase mixture, and the path to long-term, cooling is ! established through initiation of LPI near the time that the CFTs empty. The RELAF5/ MOD 2-B&W (Reference 25) code calculates system thermal-hydraulics, l core power generation, and the clad temperature response during blowdown. The l REFLOD3B (Reference 19) code determines the length of the refill period and the core
. flooding rate during reflood. BEACH (Reference 26), which is the RELAPS/ MOD 2-B&W core model with the reflood fine-mesh rezoning option activated, determines the clad temperature response during the reflood period with input from REFLOD3B. The CONTEMPT code (Reference 21) is used to determine the minimum containment pressure response based on the mass and energy release from the RCS as predicted by RELAP5 and REFLOD38.
12.3 LHR Limits The case identified in Section 4.2.1 was used as the base case for the ANO-1/TMI-120 percent tube plugging, Mark-B9 LOCA limit analyses. TACO 3 (Reference 27) steady-state fuel pin data were obtained for each case based on the assumed axial and radial peaking factors. The TACO 3 input was based on an initial prepressure of 355 psia and a maximum fuel enrichmant of 5.1 "'/o. 35
1 Framatome Technologies. lac. 86-5002073-02 Five axial power peaks centered at the middle of ihe five grid spans (at elevations of 2.506 ,4.264 , 6.021, 7.779 , and 9.536-ft) were analyzed with a constant axial peak of 1.7; the radial peak was adjusted to obtain an allowabb LHR limit. Figure 4-2 identifies the axial power shapes analyzed. Generally, the maximum LOCA LHR limit was established within a PCT range of 19,50 F to 2050 F. This PCT range was chosen as reasonable, given the sensitivity of the PCT to the metal-water reaction energy contributions at elevated temperatures. l l Results for all five elevations for the beginning-of-life (BOL) and middle-of-life (MOL) i analyses are presented in tables for each elevation. Figures for each elevation are included for the BOL analyses. The figures comprise five sets with eight figures in each set. The figures show (1) the pressure in the upper plenum during blowdown, (2) the 1 mass flow rate through the break during blowdown, (3) the mass flow rate at the l ruptured and peak unruptured locations in the hot channel during blowdown, (4) the reflooding rate, (5) the hot channel fuel and clad temperatures at the ruptured location, (6) the hot channel fuel and clad temperatures at the peak unruptured location, (7) the hot channel heat transfer coefficients at the ruptured and peak unruptured locations, and (8) the quench front advancement in the hot and average channels. Results for the end-of-life (EOL) case at the 2.506-ft elevation are presented in Table 4-4. The results of this study are extended to the other elevations based upon extrapolations of the results and trends obtained in the BOL and MOL analyses. A discussion relating to the applicability of the full-power LHR limits to reduced core I power levels is contained in Section 4.2.3.4. 4.2.3.1 BOL LOCA LHR Limits The BOL hot pin initial condit!ons for each clevation are presented in Table 4-1. The results of the BOL LOCA limit analyses are tabulated in Table 4-2 and shown in Figures 4-2 through 4-9 for the 2.506-ft location, Figures 4-10 through 4-17 for the 4.264-ft location, Figures 4-18 through 4-25 for the 6.021-ft location, Figures 4-26 through 4-33 br the 7.779-ft location, and Figures 4-34 through 4-41 for the 9.536-ft location. 4.2.3.2 MOL LOCA LHR Limits Previous LOCA analyses have shown that BOL LHR limits can be held constant until the MOL burnup where the fuel volume-averaged temperature approaches the SOL , value. These time-in-life studies, documented in BAW-10192-PA, Section AL.h are j apprcpriate provided mid-blowdown rupture is not predicted. The time-in-life analyses performed for ANO-1/TMI-1 justified maintaining the BOL allowable LHR limits for all core elevations at constant values up to a burnup of 40 GWd/mtU. The conditions at this time in life produced PCTs sufficiently bounded by the BOL PCTs. The hot pin initial conditions obtained from TACO 3 for each elevation at MOL are shown in Table 4-1 The results of the MOL LOCA limit analyses are tabulated in Table 4-3. Reported 36 l
Fr Framatome Technoloaies. Inc. 86-5002073-02
~
results for all elevations consider the expected increases based on the corrected grid data evaluated in Reference 32, a 4.2.3.3 EOL LOCA LHR LimitsI The EOL limits were established'at 60 GWd/mtU for the Mark-89 fuel design. A linear interpolation between the MOL burnup and the design burnup is appropriate based on the fuel performance input parameters during this burnup penod. Once an internal pin pressure of 3000 psia is reached for a burnup, the LHR, and subsequently the initial fuel
-temperature, is reduced as burnup increases to maintain the pin pressure below 3000 psia. The reduction in the LHR limit decreases the initial fuel stored enerw, which produces calculated PCTs that are well below the 2200 F limit. The hot pin initial conditions obtained from TACO 3 for the 2.506-ft elevation at EOL are identified in Table 4-1.
The allowable LHR limit at the 2.506-ft location was determined to be 11.6 kW/ft with a PCT of 1548 F... The LHR is limited by the TACO 3 maximum allowable internal pin i pressure of 3000 psia,800 psia above system pressure. The allowable LHRs for the other four locations were set to 11.6 kW/ft without running specific cases. The initial internal pin pressures are similar, and the transient progression for each of the other peak. power locations will be similar to the 2.506-ft location. . Table 4-4 identifies approximate values for specific parameters based on the 2.506-ft case and previous experience with the LOCA transient progression. Reported results for all elevations considerL the expected increases based on the corrected grid data evaluated in Reference 32. The LHRs after 40,000 mwd /mtU include increased uncertainty factors on the fuel volume average temperatures to account for decreases in the fuel thermal conductivity as discussed in Section 4.2.~ For the 60000 mwd /mtU case, the uncertainty factor used
. was 1.18 (Reference 11).
4.2.3.4 Partial-PowerLBLOCA'LHR Limits Core power distribution analyses generally assume that the 100 percent full power LHR
- limit. is preserved for all core power levels above 50-percent full power. Various
- LBLOCA analyses were performed for the Oconee plants with the 100' percent full-power LHR limit to determine the maximum allowable MTC as a function of core power 1 that results in a PCT that is below the PCT calculated for the 100 percent full-power
- case (Reference 28). The results of these analyses are summarized below for both 4-pump and 3-pump operation. The conclusions from the partial-power analyses are
~
applicable to ANO-1 and TMI-1 because they are all 177-LL B&W plants. Four power levels,95,75,65, and 50 percent full power, were analyzed to ratermine an appropriate MTC curve for four RC pump operation. The MTC curves used for each i 37 y s.
-- -. .- - - - - - - - - - - -- - - - - -- i
framatoma Technoloales. Inc. 86-5002073 02 analysis Were 0, +1, +2, and +5 pcm/F for the 95, 75, 65, and 50 percent full power, respectively. The results of the study demonstrated that the calculated PCT for the 100 percent full-power case would bound the partial-power operation with the specified MTC - curves. Figure 2-3 presents the allowable MTC as a function of percent full power with the key assumption of preserving the full power LHR.
- 1 1
An Oconee study was also performed with three operating RCPs at 80 percent full power with an .MTC curve of +1 pcm/F. The inoperative pump was modeled in the ~ broken cold leg to maximize the' calculated PCT for this mode of operation based on the , results of Section A.8. cf Reference.4. The results of the study showed that the { calculated PCTs for the most limiting three-pump case would be bounded by the 100 ! percent full-power case, it is concluded from these analyses that the full-power ANO-1/TMI-1 LOCA LHR limits can be maintained for partial-power and three-pump operation as long as the allowable MTC as a function of percent full power given in Figure 2-3 is preserved. ! 4.2.3.5 Core inlet and Exit LHR Limits The five LBLOCA elevations analyzed allow LHR limits to be determined anywhere between .them by linear interpolation. However, outside of these end points, extrapolation of an appropriate LHR is not defined. Reference 23 contains a sensitivity study that was performed to establish the LHR limits below the 2.506-ft core elevation and above the 9.536-ft elevation. The conclusion of that study determined that the allowable LHR at the top and bottom of the core can be calculated as 95 percent of the LHR at the 2.506- and 9.536-ft elevations, respectively. Any limit in between can be ' determined by linear interpolation. The 95 percent , extrapolated limits are not dependent on the fuel type, because any core flow effects l resulting from differing pin dimensions are already included in the 2.506- and 9.536-ft analyses on which the inlet and outlet limits are based. Additionally, the flow effects resulting from changes in assembly dimensions are uniform and not typically elevation dependent. However, core modeling changes which appear to be elevation dependent i must be considered to determine whether the 95 percent extrapolation remains bounding. The change in the modeling of the pump homologous data and two-phase multiplier was determined to be elevation dependent in Reference 17. The magnitude of the i pump resistance resulted in changes in both the positive and negative flow periods during blowdown, which resulted in changes in the PCTs and allowable LHRs. The largest effect wa's seen at the 4.264-ft and 6.021-ft elevations where the least amount of l cooling was predicted. A decrease in cooling at the 2.506-ft elevation was also seen, however the effect was not as severe because of a small benefit from setting the initial quench volume to the lowest channel volume. The quench volume selection would be an even greater benefit at the core inlet where the negative effect of the pump resistance would be further reduced. Therefore, 95 percent of the 2.506-ft elevation 38
p Framatome Technoloaies. Inc. 86-5002073-02 LHR would remain bounding for these analyses. At the top of the core, little effect on PCTs is seen as a result of the change in pump resistance, therefore 95 percent of the 9.536-ft elevation remains bounding. t j Based on these considerations, the 95 percent extrapolation developed in Reference 23 remains valid and precludes the necessity of performing calculations based on { difference fuel types or at elevations beyond the five elevations commonly analyzed. ! 4.2.4 Recent LBLOCA Issues l Two recent issues affecting generic LBLOCA analysis inputs have been addressed and l incorporated in the current analyses. These issues relate to steady state and transient i energy deposition factors (EDFs) and fuel thermal conductivity decreases with burnup. I These items are discussed in Sections 4.2.4.1 and 4.2.4.2, respectively. Section I 4.2.4.3 contains a summary of studies performed to show that the containment back pressure analyses performed with containment spray flow of 1500 gpm/ spray header are valid for ANO-1. which has the capability of achieving 1800 gpm/ spray header of spray flow. 4.2.4.1 Energy Deposition Factors The energy deposition factor is defined as the energy absorbed (thermal source) in the fuel pellet and clad divided by the energy produced by the pellet (nuclear source). I EDF = Pmerm.i.oure./ Pnucieu soure. The BWNT LOCA evaluation model reports that an EDF of 0.973 will be used for the steady-state initialization and during the blowdown portion of the transient, and an EDF of 0.96 will be used during reflood (Reference 4) for LBLOCA analyses. New methods and I predictions for the EDFs appropriate for use in LOCA analyses at various times in life i have recently been evaluated by FCF (Reference 9). These calculations do not totally support the 0.973 or 0.96 values for high burnup, low power fuel or fuel that may be surrounded by higher power fuel. As a result, the LOCA evaluations may use different EDFs, depending on the time in life and fuel pin type.. For some applications, the EDF may exceed a value of 1.0. The 20 percent tube plugging analyses have considered the new steady-state and transient EDF values developed by FCF in Reference 9. For Mark-B9 fuel assemblies , with UO2fuel rods, a steady-state EDF of 0.973 and a transient EDF of 1.0 is used in the LOCA EM calculations at burnups less than 40,000 mwd /mtU (constant LHR, high power fuel); a steady-state EDF of 1.0 and a transient EDF of 1.1 is used at a burnup of 60,000 mwd /mtU (reduced LHR, low power fuel). For Mark-B9 fuel assemblies with gadolinia fuel rods, a steady-state EDF of 0.973 and a transient EDF of 1.013 is used at burnups less than 40,000 mwd /mtU; a steady-state EDF of 1.0 and a transient EDF of 1.1 is used at a burnup of 60,000 mwd /mtU (Table 3). The RELAPS-based LOCA LHR 39
i Framatome Technoloaies. Inc. 86-5002073-02 limits are reported based on nuclear source power and the EDF is accounted for in the j LOCA EM transient calculations. Therefore, the LHR limits provided in Table 2-1 represent the total power generated by the fuel pin. The EDFs used in the Mark-89 analyses are summarized below for beginning of life , (BOL), middle of life (MOL), and end of life (EOL) conditions. i BOl and MOL EOL Steady-State EDF 0.973 1.0 Transient EDF 1.0 1.1 In the core maneuvering analyses, the LOCA LHR limit should be greater than or equal to the LHR calculated at the limits of normal operation in the peaking analysis. LHR tocA 2 LHRoeaans analys,. = Fqp. g
- F.ug
- LHR.v. l l
where Fqp.,x is the product of the axial peak and the radial peak, F:ug is the product of all augmentation factors (including committed LOCA target margin), and LHR ve is the core average LHR as calculated by LHRav. = [(P, r.o
- FOP ) / (Noin
- N s y
- Leu.i))
- EDF .
In this equation Prar.o is the 100 percent rated power, FOP is fraction of the core power, Npin is the number of fuel pins in an assembly, Nas3y is the number of fuel assemblies in I the core, Lru.i is the length of the active fuel, and EDF is the energy deposition factor. The LHR v. (and hence the LHR in the peaking analysis) is in terms of the energy produced (Pnucie,, oure.) when the EDF is not applied (or EDF = 1.0). The LHR limits are reporteu in this document in terms of energy generated by the pin (nuclear source). As long as the limits are defined this way, an EOF would not be useci in calculating the core average linear heat rate that is used in a peaking margin calculation to convert the peak calculated by the nuclear design code to a calculated LHR. Therefore, the maneuvering analysis should set the EDF to 1.0 for an appropriate calculation of margin to the reported LOCA LHR limits. 4.2.4.2 Bumup Fuel Thermal Conductivity The NRC-approved TACO 3 fuel performance code uses a conductivity model that varies only with temperatute and not with burnup. Recently, SIMFUEL data has become available that demonstrates that fuel thermal conductivity decreases with extended burnup (Reference 10). Since the TACO 3 model is based on a beginning-of-life conductivity curve, LOCA. initialization fuel volume-average temperatures calculated at high burnups are nonconservative. Justification for not using a variable thermal conductivity versus burnup model in TACO 3 is supported by increasing the fuel volume-average temperature uncertainty factor for pin burnups exceeding 40,000 mwd /mtU. 40
1 Framatome Technoloaies. Inc. 86-5002073-02 The NRC, as discussed in the tecMical evaluation report (TER), has approved this method for BAW-10186 (Reference 10). The value of the increased uncertainty factors used in the Mark-89 LHR calculations at burnups greater than 40,000 mwd /mtU are discussed in Section 4.2.3.3. l t 4.2.4.3 Increased Containment Spray Flow (1800 gpm vs.1500 gpm) The CONTEMPT containment analysis results that were used in the BWOG 20 percent SG tube plugging LDLOCA models were generated in Reference 8. The containment analysis was performed using inputs intended to minimize containment backpressure. This is conservative since the lower pressure leads to greater steam production in the core and increased steam binding. The steam binding effect conservatively limits the flooding rate. An important parameter for the containment analysis is the containment spray flow rate. The current analysis results are based on a spray flow of 1500 gpm per train. This spray flow is not strictly conservative for the BWOG 20 percent SG tube plugging analysis since ANO-1 has a maxirnum spray flow of 1800 gpm per train. Studies were performed to determine the effect of the higher (1800 gpm/ train) flow rate on analysis results and also to document whether further analysis is needed to bound ANO-1 containment spray flow rates (Reference 13). The containment spray flow study showed that the change in the containment spray flow rate had little effect on peak clad temperatures. Peak ternperatures occur early in the transient, around 30 seconds, while containment spray does not activate until 65 seconds. Long-term whole core hydrogen generation may be affected by a change in spray flow, but the studies performed shcwed that changing spray flow from 1500 ppm to 1800 gpm does not significantly affect core flooding rates. Also, the methods used to calculate and report whole core hydrogen generation are extremely conservative, and despite the conservatism, BOL and MOL cases show that there is a large margin available in the calculation of whole core hydrogen generation. The results of this study show that a higher spray flow rate does not adversely affect the BWOG 20 percent SG tube plugging results with respect to PCT or whole core hydrogen generation. The results presented in Section 2 are then bounding for ANO-1 with containment spray flows up to 1800 gpm/ train. ' 4.2.4.4 Pump Type in response to PSC 1-99, the Reference 13 analyses were reviewed and found to be based on the Bingham pump type. The Reference 13 analyses were based on the combination of the TMI-1 and ANO-1 limiting parameters, however, neither plant operates a Bingham type RCP. TMI-1 operates Westinghouse type RCPs, while ANO-1 operates Byron-Jackson type RCPs. In' order to model the most limiting pump configuration, a sensitivity study was performed in Reference 17. Based on the comparison of pump types and two-phase multipliers, the W pump with use of the M3 two-phase multiplier is limiting for the 20 percent tube plugging analysis 41
Framatome Technolooies. Inc. 86-5002073-02 considering the TMI-1 and ANO-1 177-LL B&W units. The LBLOCA cases from R?ference 13 were then reanalyzed with the appropriate pump data, with the final results reported in Reference 17. 4.2.4.5 FuelDimensions The base calculations performed in Reference 13 modeled the Mk-B10 pin dimensions instead of the Mk-B9 pin dimensions. The reanalysis for the pump data performed in Reference 17 contained the correction to the fuel radius and cladding inner radius to correctly model the Mk-B9 pin dimensions. 4.2.4.6 Grid Type After the analyses performed for PSC 1-99 v.ere completed (Reference 17) an error was discovered in the hot channel reficcd grid data. The error was discovered in response to completing a restrictions and limitations checklist (Reference 30) based on the NRC Safety Evaluation Report on the BWNT LOCA EM (BAW-10192-PA). The hot channel grid was inappropriately modeled as the Mk-B11 mixing vane grid instead of the Mk-B9 non-mixing vane grid. The analyses in Reference 17 were reevaluated based on the correct hot channel reflood grid data for Mk-B9 assemblies. In response to the increased PCTs resulting from the lower Mk-89 grid enhancement, the 7.779-ft elevation LHR was reduced from 17.3 kW/ft to 16.8 kW/ft. Appropriate PCT penalties were applied to the time in life analyses to consider the corrected grid data. The results from the grid reevaluation (Reference 32) provide the final BWOG 20% tube plugging LBLOCA results. 4.2.4.7 Replacement 15% Tube Plugging LHR Limits for TMI-1 Reference 36 outlines a list of precautions associated with the LBLOCA analyses performed for the TMI-1 15% SG tube plugging analyses that support current plant operation. The 20% SG tube plugging analyses model additional tube plugging conservatism and address the precautions associated with the 15% SG tube plugging analyses. Therefore, the 20% SG tube plugging LHR limits and PCTs summarized herein (and Reference 37) are bounding for the current operation of TMI-1 and replace the results of Reference 36. 4.2.4.8 Extension of LBLOCA Results to Greaterthan 60,000 mwd /mtU The end-of-life analyses were performed at 60,000 mwd /mtU. Extension of the LHR limits to times-in-life beyond 60,000 mwd /mtU must be evaluated separately. 42
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FIGURE 4-2. 2.506 FT, BOL LOCA LIMIT CASE - REACTOR VESSEL UPPER PLENUM PRESSURE. 2400 . 160 2000- *
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86-5002073-02 FIGURE 4-4. 2.506-FT, BOL LOCA LIMIT CASE - HOT CHANNEL MASS FLOW RATES. 40 LEGEND R"rtured, Seg 7 15 30 - --- Unruptured, Seg 6 E 10 20- Q e 6 6 I 10 5 b
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86-5002073-02 FIGURE 4-6. 2.506-FT, BOL LOCA LIMIT CASE - HC FUEL & CLAD TEMPERATURES AT RUPTURED LOCATION. 400 LEGENO l 1500 i Fuel, Seg 7 2000- - - - Clad, Seg 7 4
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l
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O O 40 80 120 160 200 240 TIME, SECONDS { FIGURE 4-7. 2.506-FT, BOL LOCA LIMIT CASE - HC FUEL & CLAD TEMPERATURES AT PEAK UNRUPTURED LOCATION. 00 LEGEND 1500 Fuel, Seg 6 2000- - - - Clad, Seg 6
./ '\ s w 1250
,1600< . I S.s a
\ 1000
- 1200-k l
l 750 800, I Y 400- k 500
*~---_.~.~._._._. . . , ,
0 0 40 80 120 160 200 240 ; TIME, SECONDS
~
50 3
86-5002073-02 FIGURE 4-8. 2.506-FT, BOL LOCA LIMIT CASE - FILTERED HC HEAT TRANSFER COEFFICIENTS. 10 ' !- LEGEND Ruptured, Seg 7 h - --- Unruptured, Seg 6 10' y b 10'i N 8 N o g .r* %. n
. _ - =. wA 10' $,
10 ' : f d E i h o
- g. f , :10, 10*: t/ g ,
f z I H 30 H H 10*1 6 h , I 10* 10* O 40 80 120 160 200 240 TIME, SECONDS FIGURE 4-9. 2.506-FT, BOL LOCA LIMIT CASE - QUENCH FRONT ADVANCEMENT. 12 - I 3.2 10- , j' l g 8- , 2.4 2
./ g 5
6' g* r/ E g
./ 1.6 u
'~/ f
- a. ~-
f 2, . r'/./ LEGEND 0.8
/
/ Hot Channel 0
[ ----- Average Channel 0.0 0 40 80 120 160 LOO 240 TIME, SECONDS 51.
- w. . ,,
( 86'-5002073-02 l o-
,o FIGURE 4-10. 4.264 FT, BOL LOCA LIMIT CASE -
-F .
REACTOR VESSEL UPPER PLENUM PRESSURE. 2400 ' 160 1
- 2000 <
1 120 1600< 1200<
,80 u) 800< <
40 400< i 0 0 0 4- 8 12 16 20 24 28 32
)
. TIME, SECONDS FIGURE 4-11. 4.264-FT, BOL LOCA LIMIT CASE -
x10' BREAK MASS FLOW RATES. xi O' 10 t.EGEND
'4 RV Side 8 --- - Pump Side ,
W 3 g 6-g ti g 2 I 4'
' I- 3:
o y W E. 2<
N. 3 $
g I \'N.s,N o~___ _ . . _ . % D' - 0 2 0 4 8 12 16 20 24 28 32 TIME, SECONDS g
~a L
;;; 1 g& _._
86 - 5 0020 73 -oy FIGURE 4-12. 4.264-FT, BOL LOCA LIMIT CASE - l HOT CHANNEL MASS FLOW RATES. LEGEND i Ruptured, Seg 9 15 1 30, ----- Unruptured, Seg 10 - i j G 10 E 20-3 14 \\
'l N
g .5 10- ' 3: { 6 3 ' e 0: - - 0 i 10< 5 1 0 4 8 12 16 20 24 28 32 TIME, SECONDS FIGURE 4-13. 4.264-FT, BOL LOCA LIMIT CASE - CORE FLOODING RATE. 6 14 5 12 N E 2 4 .jo o i E f 3- $ e B 2' f e l8 u 4 u l' 2 0 0 0 40 80 120 160 200 240 TIME, SECONDS 53
86-5002073-02 FIGURE 4-14. 4.264-FT, BOL LOCA LIMIT CASE - HC FUEL & CLAD TEMPERATURES AT RUPTURED LOCATION. 2 00 LEGEND 1500 Fuel, Seg 9 2000< . - --- Clad, Seg 9
\.
T -N,m.v. 1250
/
1600< / fi 5= %
- u. .
g g
. 1000
\,
11200 ) & 800< < 400< 500 I L.-._.-.._.. ._.
'O O 40 80 120 160 200 240 TIME, SECONDS FIGURE 4-15. 4.264-FT, BOL LOCA LIMIT CASE - 3' HC FUEL & CLAD TEMPERATURES AT PEAK UNRUPTURED LOCATION.
2400 LEGEND 150Q Fuel, Seg 10 2000, --- - Clad, Seg 10 ,
,/N'a .n*%,,% % , 1250 1600<
j
'N.g u
, 1000 1200< - ,
l' .
\*
750 gg,
)
)
400, 500 l 0 0 40 80 120 160 200 240 TIME, SECONDS 54
. ~ .
86-5002073-02 FIGURE 4-16. 4.264-FT, BOL LOCA LIMIT CASE - FILTERED HC HEAT TRANSFER COEFFICIENTS. l 10 LEGEND g Ruptured, Seg 9 4 - - - Unruptured, Seg 10 : 10' w k 10': A
'Q N g /N- 3 4
- / 104
.,/
10 'i 'g I e
' J
- 10* }
10*j '
.// h g
I H : 10' p i H g.10*p ~ $ { z ! ( 10* 10' O 40- 80 120 160' 200 240 l l TIME, SECONDS i FIGURE 4-17. 4.264-FT, BOL LOCA LIMIT CASE - QUENCH FRONT ADVANCEMENT. j 12 - l 1 3.2 l
/
- b. 8< ,/ 2.4 2
/
6< /.A 1.6 E
/s' f l
4- ,
, ,7*/* i
, .d 2<
/./
f /'" LEGEND O.8 Hot Channel g -- - - Average Channel 0 40 80 ' 120 160 200 240 l
. TIME, SECONOS !
~
11
, 55 :
i f
86-5002073-02 FIGURE 4-18. 6.021-FT, BOL LOCA LIMIT CASE - REACTOR VESSEL UPPER PLENUM PRESSURE. 2400 g ,gg 2000-120 1600-1200- 80 B B E E 800 40 100-0 - 0 - 0 4 8 1'i 16 20 24 28 32 TIME, SECONDS FIGURE 4-19. 6.021-FT, BOL LOCA LIMIT CASE - x10* BREAK MASS FLOW RATES. x10* 10 -- LEGEND RV Side '4 8, - - - Pump Side , le 3 g h 6- g ti g 2 4' ' g o ' 3 d ' w u.. % . $
'1 j 2- N, -
g
\
(. 0-
~. 4
-0 l
1 2' -- - l 0 4. 8 12 16 20 24 28 32 TIME, SECONDS 56 L #
86-5002073-02 FIGURE 4-20. 6.021-FT, BOL LOCA LIMIT CASE - HOT CHANNEL MASS FLOW RATES. 40 LEGEND Ruptured, Seg 12 15 30< - --- Unruptured, See 13 i
$ il -
10 Q g 20< P, g i I 6 I 10< .5 i < g. d I 3 u.
,h o, . . . . . . _
.o h
- 2. .[ . 3 10' '
. -5
-20 ---
,. 0- 4 8 12 , 16 20 24 28 32
, TIME, SECONDS iU FIGURE 4-21. 6.021-FT, BOL LOCA LIMIT CASE -
, L CORE FLOODING RATE.
6 14 5< 12 e G E 3 4- .jo u Y' 5 f , g i
, 3, .
'S l
'O
__ d g 2< 0 g 1 o 1 4
'. j, 2
0 -- - O e O' 40 '. 80 120 160 200 240
.a. !, TIME, SECOWS !
\
57
.;h
86-5002073-02 FIGURE 4-22. 6.021-FT, BOL LOCA LIMIT CASE - HC FUEL & CLAD TEMPERATURES AT RUPTURED LOCATION. 2400 LEGEND l
~
1500' Fuel, Seg 12. (. 2000- A, %- - Clad, Seg 12 , .
.j.,j%.) N. A. 1250
. ' s= f\/ ' s_ \
1000 5 a W
'N;\
- 1200 *- <
g h' l 800< 750 $: l 400, \. 500 L . ._. . _ . . 0 0 40 80 120 160 200 240 TIME, SECONDS FIGURE 4-23. 6.021-FT, BOL LOCA LIMIT CASE -
.HC FUEL & CLAD TEMPERATURES AT PEAK UNRUPTURED LOCATION.
2400 LEGEND 1500 Fuel, Seg 13 2000, - - - Clad, Seg 13 O'\.^d*/*%.A. .% 1250
,1600- ,
**s,s.%
o
, "N, 1000
\.
11200 800< 750 400, 500 0 40- 80 120 160 200 240 TIME, SECONDS
,_ 58
7: ' 86-5002073-02 ) i FIGURE 4-24. 6.021-FT, BOL LOCA LIMIT CASE - FILTERED HC HEAT TRANSFER COEFFICIENTS. 1 ' 1: LEGEND Ruptured, Seg 12 h - - - Unruptured. Seg 13 10' w g 3o. - A
$. N 10*: .I
! : 10 '
{c , w E
't ./ w I
$ ! :10' 10' .
.y.
I 10 H : H i g 10 ' . $ j M I
)
; 10
- l 10* l
- 0. 40 80 120 160 200 240 TIME, SECONDS FIGURE 4-25. 6.021-FT, BOL LOCA LIMIT CASE -
QUENCH FRONT ADVANCEMENT. 12 - g
. 3.2 10- /
/
/
g 8' 2.4 2 j- ,
/* y
. o 6 / E 1.6 5
/,/
i 4 p*f, h a
,/
l ,/ 0.8 l 2 [ LEGEND
,! Hot Channel
- - - Average Channel 0 0.0 0 40 80 120 160 200 240
[ TIME, SECONDS f 59 L. _ _
r-
- m. i 86-5002073-02 L FIGURE 4-26. 7.779-FT, BOL LOCA LIMIT CASE -
REACTOR VESSEL UPPER PLENUM PRESSURE. 2400 160 [ 2000-120 1600-
@ .- 2 1200- N 80 m M i
- h. '. E 800- N 40 400 <
0 - 0
. 0 4 8 12 16 20 24 28 32 TIME, SECONDS FIGURE 4-27. 7.779-FT, BOL LOCA LIMIT CASE -
xi O' BREAK MASS FLOW RATES. xio- . 10 t.EG END RV Side '4 8
- - - Pump Side .
. i Q 3 g 6- g h
6 2 k y 4- y o O d u. g E $
'\'s,
'l 2-
$ 2 0- ' -
0
~
0 4 8 12 16 20 24 2B 32 TIME, SECONDS 60
86-5002073-02 FIGURE 4-28. 7.779-FT, BOL LOCA LIM!T CASE - HOT CHANNEL MASS FLOV'"^TF.S. 40 p ,, NO
. >$- ed, Seg 15 15 30, - - - Unruptured, Seg 16
@ 10 E l 20< o a x N I 10- 5 g o 3 d - w m w 1 m O. 0 < l l 2 i 10' 5 I 1 20 0 4 8 12 16 20 24 28 32 TIME, SECONDS
) > FIGURE 4 29. 7.779-FT, BOL LOCA LIMIT CASE -
CORE FLOODING RATE. 6 14 5 ) 12 n g E 4< u
.jo N N
& E E 3- < E 5 a '8 8 d ' e 2 ^~ u 4 o 1< 2 0 0 0 40 80 120 160 200 240 TIME, SECONDS g;t-
1 l 86-5002073-02 FIGURE'4-30. 7.779-FT, BOL LOCA LIMIT CASE - HC FUEL & CLAD TEMPERATURES AT RUPTURED LOCATION. 2400 1500 i 2000-h(,gf. ,^ %.N.m. y 1250
,1600- ,j N %,
j ' ~.w e 1000 g 1200- ' N g d I a 800 400' LEGEND 500
- Fuel, Seg 15
- - - Clad, Seg 15 0l '
1 0 40 80 120 160 200 240 TIME, SECONDS j
)
FIGURE 4-31. 7.779-FT, BOL LOCA LIMIT CASE - HC FUEL & CLAD TEMPERATURES AT PEAK, UNRUPTURED LOCATION. 2400' JEGEND 1500 Fuel, Seg 16 2000 - - - Clad, Seg 16 l y'%A A.J's.#.% ~._'- 1250
]
- u. 's"
{\.] ~ ~. a l s .s._' N . g - j 1000 , 1200- l l' I 2 . 750 W l 800-400, 500 O - 0 40 80 120 160 200 240 l TIME, SECONDS l 62
86-5002073-02
- FIGURE 4-32. 7.779-FT, BOL LOCA LIMIT CASE -
FILTERED HC HEAT TRANSFER COEFFICIENTS. 10 LEGEND Ruptured, See 15
+ --- - Unruptured, Sep 16 ( 10' g
.10' h e e E '
88
;3o, P 5
l ' 10 : 0
> C C tb 10' .o. :
5
^
-[
f H 10*:
- 10' [
6 I _'f
; 10+
0 40 80 120 160 200 240 TIME, SECONCS FIGURE 4-33. 7.779-FT, BOL LOCA LIMIT CASE - QUENCH FRONT ADVANCEMENT. 12 ; 3.2
/
10< j
.s' g 8- j '.f*/ 2.4 lE
./
j ' 1.6 5 8 0 4 //
./.
,/ 0.8 ,
2, [ _ LEGEND Hot Channel
--- - Average Channel 0 00 0 40 80 120 160 200 240 TIME, SECONDS 63 1 d
86-5002073-02 FIGURE 4-34. 9.536-FT, BOL LOCA LIMIT CASE - REACTOR VESSEL UPPER PLENUM PRESSURE.
'2400 160 2000<
120 1600-1200- '
- -80
=
h E 800-40 400-0 4 8 12 16 20 24 28 32 TIME, SECONDS FIGURE 4-35. 9.536-FT, BOL LOCA LIMIT CASE - x10' BREAK MASS FLOW RATES. x10' 10 LEGEND
- RV Side '4 8, - --- Pump Side ,
E! 3 g g 6< g 6 Y
-{ 2 I g 4' ,
g
'o 3
" s.
$ A m
k 2-
\'\, s
'l E
2
^ '
0 - * - - - - 0 0 4 8 12 16 20 24 28 32 - TIME, SECONDS 64
86-5002073-02 ! FIGURE 4-36. 9.536-FT, BOL LOCA LIMIT CASE - HOT CHANNEL MASS FLOW RATES. 40
! LEGEND
'{' Ruptured, Seg 18 15 30, --- - Unruptured, Seg 17 ,
20- 10 fe < g g d i 10 '0 h j i o 1 0' -- 0 10-
-5 0 4 8 12 16 20 24 28 32 TIME, SECONDS FIGURE 4-37. 9.536-FT, BOL LOCA LIMIT CASE -
CORE FLOODING F. ATE. 6 14 5-12 4<
<. 10 0
t N Y 3 h $ 8 e e g 2< - a u g 4 u 1-2 0 0 40 80 0 120 160 200 240 TIME, SECONOS
. 65
86-5002073-02 FIGURE 4-38. 9.536 FT, BOL LOCA LIMIT CASE - i
!!C FUEL & CLAD Tr'VIPERATURES AT RUPTURED LOCATION.
2400 , 1500
'2000- , f. -
.$,I\[\,A g,m, 1250 j 1600-
- u. .f g '~ A.%,s.s ~ v. u ,
i
.g ,
1000 p l j 11200-,gg. 750 e h- 1 l 400' LEGENa 500 ; Fuel, Seg 18
~~
0 --- ' 'O O l 0 40 80 120 160 200 240 Til AE, SECONDS d FIGURE 4-39. 9.536-FT, BOL LOCA LIMIT CASE - h FUEL & CLAD TEMPERATURES AT PEAK UNRUPTURED LOCATION. LEGEND 1500 Fuel, Eeg 16 2000- ~ ~ - - - - ---- Clad. Seg 10
*py,jw%J'~ *%.,
/ , N % ,, 12f0 1600-
- j' 'N w.
1200-f* ' N g ,' 1000 800- 50 400- c 500 o 0 0- 40 80 120 160 200 240 TI' 4E, SECONDS 6G l
)
86-5002073-02 FIGURE 4-40. 9.536-FT, BOL LOCA LIMIT CASE - FILTERED HC HEAT TRANSFER COEFFICIENTS. LEGEND g Ruptured, Seg 18 4 ---- - Unruptured, Seg 16 :10' g tt 10- 4 4
- R 4
- 10 '
10 ' : 8 ' g
- 10' E
lg 104:' ,
- W W ,[ '
9 g q l 10* @ g H 10*: b
- 10*
10* O 40 80 120 160 200 240 TIME, SECONDS FIGURE 4-41, 9.536-FT, BOL LOCA LIMIT CASE - QUENCH FRONT ADVANCEMENT. 12 ,
.]
/ 3.2 10'
/ ,,[ .
,./.
g 8' p 7 #' 2,4 g 9-s' Z E 6' /* ' 5 3
.r' 1'e 5
./ 5 1 S 4 .I , $
f 0.8 2< [ LEGEND Hot Channel
- - - Average Channel 0 0.0 0 40 80 120 160 200 240 i TIME, CECONDS 67
Framatome Technoloales. Inc. 86-5002073-02 I
- 5. SBLOCA Sensitivity Studies and Analyses The SBLOCA sensitivity studies and analyses are presented in the following sections.
. A summary of inputs and sequence of events for the SBLOCA analyses are presented In Table 4-1 through Table 4-4 and Figure 2-4.
5.1 SBLOCA Sensitivity Studies SBLOCA analyses require that .various sensitivity studies be performed with the evaluation model to demonstrate model convergence and to identify the most limiting set of boundary conditions or break locations that should be used in demonstrating compliance with the five criteria in 10 CFR 50.46. As part of the SBLOCA EM, FTl performed numerous SBLOCA sensitivity studies to confirm modeling techniques and methods. Although the EM was based on a slightly different plant design, the safety evaluation report for BAW-10192-PA supports the application of the EM to the 177-FA plants, and FTl has determined that the SBLOCA sensitivity studies performed in the EM are directly applicable to and appropriate for use in the ANO-1/TMl-1 analyses. One additional sensitivity study was performed using ANO-1/TMI-1 plant-specific parameters to determine the appropriate CFT initial conditions for use in the CFT line break
. analysis. Section 5.1.1'provides a discussion of the generic sensitivity studies from the reference EM report that have been applied. Section 5.1.2 describes the plant-specific CFT sensitivity study and Section 5.2.1 identifies the SBLOCA base model used for the SBLOCA break spectrum analyses.
5,1.1 SBLOCA Evaluation Model Generic Studies
.The majority of the SBLOCA sensitivity studies presented in the EM topical report (Reference 1, Volume ll) are generic and apply to any SBLOCA analysis for the B&W-designed nuclear steam system. An example is the RELAP5/ MOD 2-B&W time-step study, which showed that the automatic time step selection in RELAP5/ MOD 2-B&W would produce converged results. This demonstration need not be repeated for plant-specific applications in which the modeling techniques used are represented by those in the~ EM studies. The following is a listing of the sensitivity studies considered to be generic, with a discuss' ion of why the conclusions of the study are applicable to this ;
SBLOCA applications report. For convenience, each discussion is referenced to the ! section in the EM topical report where the study is documented.
'5.1.1.1 SBLOCA Time-Step Study The study using the generic EM, documented in BAW-10192-PA, Volume 11, Appendix A, Section A.2, verified that, for light water reactor geometry, the RELAPS time-step controller govems the code solution sufficiently to assure convergent results. In 68
Framatome Technoloaies. Inc. 86-5002073-02 i RELAP5/ MOD 2-B&W, the user specifies a maximum time step that can be modified internally by the code in the event of convergence or Courant limitations. The SBLOCA EM time-step studies justified use of a 20-millisecond maximum time-step size as i appropriate for B&W-plant SBLOCA analyses. The EM controls the plant input models such that no significant deviation in the number or size of the control volumes or heat i structures, critical to the model results, can be included between plant designs. Since the SBLOCA analytical mode! is similar to the model used for the EM time-step study, and the maximum time-step size is 20 milliseconds in the SBLOCA analyses, then the RELAPS/ MOD 2 time-step controller will also adequately control the problem advancement for these applications. The EM study remains valid, therefore, and this study does not have to be repeated. 5.1.1.2 SBLOCA Pressun'zer Location Study Previous configuration studies performed with the SBLOCA EM (BAW-10192-PA, Volume 11, Appendix A, Section A.3) showed that there is little difference in results when the pressurizer is connected to the broken loop instead of the intact loop. This result is expected since the SBLOCA transi<ent is domina+ed by such factors as leak flow, decay heat generation rate, initial primary liquid inventory, and ECCS injection rates. Therefore, the pressurizer location study performed with the EM is applicable to the SBLOCA analyses and need not be repeated. 5.1.1.3 SBLOCA Core Crossflow Resistance Study Core crossflow is modeled in the base model through the use of RELAPS/ MOD 2-B&W crossflow junctions between the hot and average channels in the core region. The crossflow areas are calculated based upon the actual flow area exposed by the three-by-four matrix of fuel assemblies in the hot channel, and the junction form loss factors are input based on the method discussed for the EM base case (BAW-10192-PA, Volume ll, Appendix A). This scheme was found to increase the flow diversion out of the hot channel while restricting the flow of lower tamperature steam from the average to the hot channel during core uncovering, thereby, maximizing the hot channel peak clad temperature prediction. There are no significant differences between the cases used for the EM and the ANO-1/TMI-1 analyses. Therefore, the studies performed for the EM remain applicable and do not need to be repeated. 5.1.1.4 SBLOCA Core ChannelModeling Study The core noding in the ANO-1/TMI-1 model used 20 axial nodes to model the heated fuel assembly region with twelve assemblies in the hot channel and the remaining assemblies lumped into the average channel, in addition, each channel included an unheated segment at the inlet and exit. The EM study (BAW-10192-PA, Volume ll, Appendix A, Section A.5) used a similar model, which was shown to ensure calculation 69
Ergmatome Technoloaies. Inc. 86-5002073-02 of a conservative peak clad temperature for those cases in which the mixture level descends into the heated core region. Therefore, this study does not have to be repeated for this application. 5.1.1.5 SBLOCA CFTLine Resistance Study The core flooding system consists of two pressurized CFTs that are each connected to , the reactor vessel downcomer by a surge line containing two check valves and an ! isolation valve. Following an SBLOCA, the primary system may depressurize to the l CFT fill pressure, allowing flow from the tanks and lines to enter the RV downcomer at a variabie rate, depending on the CFT line resistance and the pressure drop between the CFTs and the RV downcomer. The CFT line resistance study performed with the EM (BAW-10192-PA, 2 Volume 11, Appendix A, Section A.7) included analyses of the base 0.1-ft break and a larger 1.0-ft break. This study confirmed that a CFT line resistance of one-hundred times the nominal value is appropriately conservative and acceptable for use for all SBLOCA analyses, except for the CFT line break. The CFT line break analysis uses the nominal resistance as stated in Section A.7 of the SBLOCA EM. Since the geometry, phenomena, and mode!ing of the reactor vessel downcomer region are similar between the currant applications and the EM cases, the EM CFT line resistance study remains apprcpriate and applicable. 5.1.1.6 SBLOCA Break Discharge Coefficient Study . i The break discharge coefficient study perfomted with the EM (BAW-10192-PA, Volume ll, Appendix A, Section A.8) confirmed that all classical EM applications should be performed with the set of high break void discharge coefficients. In the ANO-1/TMI-1 analyses, all break flow model discharge coefficients were set equal to 1.0. The classical EM applications include the reactor coolant pump discharge location with the reactor coolant pumps tripped. The break discharge coefficient studies performed with the EM confirmed that, during the critical boiling pot phase of a CLPD SBLOCA, the break volume void fraction was approximately 98 to 99 percent. The data verify that , the CD should be 1.0 at these high void fractions. The process, identified in DAW-10192-PA, Volume ll, Section 4.3.2.4, states that the high break voiding discharge l coefficient range should be used for all classical EM SBLOCA applications. It also includes provisions for reanalysis of the case if the SBLOCA transient evolves to, or spende the critical portion of the transient with the break inlet conditions within, the intermediate void fraction range. These conditions may be encountered in a hot leg SBLOCA or in a pumps-on simulation of any other SBLOCA break location. Since none of these special analyses were performed for this application, the EM results for the high void discharge coefficient method remain applicable. 70
Framatome Technoloales. Inc. 86-5002073-02 5.1.2 SBLOCA Evaluation Model Plant-Soecific Sensitivity Studies FTl determined that two additional sensitivity studies were required for these plants. The studies were to confirm' the worst CFT level for a CLPD break and the worst combination of level and pressure in the CFT for a CFT line break. These studies are discussed in Sections 5.1.2.1 and 5.1.2.2, respectively. 5.1.2.1 CFT Levelfor CLPD Breaks Historically, it has been assumed that a minimum CFT volume generates the most
- conservative results for a SBLOCA, because the minimum ECCS liquid !: m!!fuie for i core cooling. However, it has recently been noted that the PCTs for tM limiting bra % <
occur shortly after the CFTs begin to discharge. Therefore, the total 6 mouM J "' liquid may not be important. Instead, how quickly the CFT liquid reaches R,e cors m:.; be the deciding factor. A sensitivity study was run (Reference 6) for a 0.08-ft2 CLPD break where combinations of the CFT level and pressure were varied. Two cases were studied. The first considered a minimum CFT liquid volume with a minimum gas' pressure. This is the historic modeling configuration of the CFT. The second case considered a maximum CFT liquid volume with a minimum gas pressure. Considering a minimum gas pressure will produce the minimum injection rate in all cases, but the liquid volume will change the injection rate substantially based on the perfect gas law. The larger gas volume associated with the minimum liquid volume will inject liquid into the system at higher rate that the smaller gas volume associated with the larger liquid volume over the same differential pressure range. This phenomenon ends the temperature excursion and quenches the entire core earlier in the transient. Therefore, the case with a minimum CFT volume has a lower PCT than the maximum CFT volume case. Based on this study, the entire CLPD spectrum was performed with a maximum CFT liquid volume and a minimum gas pressure. The CFT parameters for the ANO-1/TMI-1 SBLOCA analyses are summarized in Table 5-1. 5.1.2.2 CPTLine Break The CFT-line break prevents one CFT and one LPI train from injecting into the reactor vessel. A failure of one diesel to start disables the other LPI pump and an HPl train. The remaining ECCS available for core cooling consists of one HPI train and one CFT. Historically, it has been assumed that a minimum volume in the CFT would produce the
- most limiting results, because the liquid available for core cooling is minimized To confirm this assumption, a sensitivity study was performed on the CFT initial conditions to determine the worst initial configuration of liquid volume and gas pressure.
Three case were analyzed to examine the following combinations of CFT liquid volume and pressure: (1) maximum volume with minimum pressure, (2) minimum volume with minimum pressure, and (3) minimum volume with maximum pressure. The results j l 71 u_-- _ _ - - --_
Framatome Technolooies. Inc. 86-5002073-02 indicated that the calculated PCT for each of the cases did not exceed the initial cladding temperature of 715 F, but core uncovering was predicted in all cases except the case that considered the maximum volume with the minimum pressure. The case that considered the minimum volume with the maximum pressure predicted the most core uncovering. In this case, the combination of high pressure and small liquid volume empties the CFT relatively quickly. Consequently, there is a substantial amount of time before the HPl begins during which core inventory is boiled off. As a result, approximately two feet of core uncovering is predicted. This study determined that the combination of minimum liquid volume and maximum gas pressure in the CFT predicts the most severe results for a CFT line break. However, the predicted PCT remains less than the initial clad temperature. The CFT parameters for the ANO-1/TMI-1 SBLOCA CFT line break analysis are summarized in Table 5-1. Table 5-1: SBLOCA CFT Volume and Pressure inputs SBLOCA Break Type CFT Liquid Volume CFT Gas Pressure (ft3/ tank) (psia) CLPD Spectrum 985 580 HPl Line Break (max) (min) CFT Line Break 895 650 (min) (max) l I i
)
1 ! 72 ' j
I Framatome Technoloaies. Inc. 86-5002073-02 5.2 SBLOCA Analyses at 2772 MWt This section presents the results of the SBLOCA spectrum analysis performed for ANO-1HMI-1 at an initial core power level of 2772 MWt with 20 percent average steam generator tube plugging. The model described in Section 4 of this report was used in the analysis. This spectrum consists of a number of break sizes analyzed at the cold leg reactor coolant pump discharge (CLPD) location, plus a CFT line break and an HPl line break. The specific break areas, in square feet, analyzed at the CLPD location were: 0.01, 0.03, 0.04, 0.05, 0.06, 0.07, 0.08, 0.09, 0.10, 0.15, 0.30, 0.50, and 0.75. A discussion of the SBLOCA base model is presented in Section 5.2.1. Based on similarities in transient response, EBLOCAs can be grouped into three categories: small, intermediate, and large. A general discussion of SBLOCA phenomena common to all three categories is provided in Section 5.2.2. The results of the spectrum analysis are presented in Section 5.2.3. The results of the HPI line break and the CFT line break are presented in Section 5.2.4. 5.2.1 SBLOCA Base Model The EM studies determined that the most limiting SBLOCA break location is in the bottom of the cold leg piping between the reactor vessel inlet nozzle and the HPI nozzle. That break location is used for the current SBLOCA break spectrum analysis, except for two special break cases:' an HPl line break and a CFT line break, which are also included. The high void discharge coefficient method is applied in the model. The steam generator tube plugging is set to 25 percent in the broken loop and 15 percent in the intact loop, with 75 percent of the EFW region tubes assumed plugged. The i pressurizer is attached to the intact loop. The 177-FA lowered-loop plant model uses an entire core of Mark-B915 X 15 fuel assembi.ies with an initial power level of 102 percent'of 2772 MWt and an axial power shape with a 1.7 peak at the 9.536-ft elevation. The hot channel contains twelve assemblies with a peak linear heat rate of 17.5 kW/ft. The remaining 165 assemblies are grouped into the average channel. The entire range of core input parameters are covered by using a O pcm/F moderator temperature coefficient to define the moderator reactivity feedback curve, with an end of life (EOL) ' beta-effective of 0.007102. The beginning of life (BOL) initial fuel temperature, BOL ; oxide thickness, and EOL pin pressures are used to cover all fuel burnup times. A minimal tripped rod worth is used. ' A constant heating ramp rate of 1.0 is applied for the EM pin rupture model. The use of this ramp rate can prevent clad rupture by increasing the clad rupture temperature during the analyses. Clad rupture at cladding temperatures less than approximately 1600 F allows cooling of the inside and outside clad surface without significant metal-water reaction. For these lower cladding temperatures, preventing clad rupture maximizes the peak clad temperature. For higher cladding temperatures where the metal-water reaction contributes to the peak clad temperature, the pin pressure and/or 73
'Framatome Technoloaies. Inc. 86-5002073-02 the constant heating ramp rate are adjusted to obtain clad rupture at the most limiting time during the transient. For the ANO-1/TMI-1 SBLOCA analyses, since the PCTs were less than 1600 F, clad rupture was prevented by using the heating ramp rate of 1.0.
Only one HPI pump and one LPI pump (one ECCS train) are modeled in the analyses due to the single-failure assumptien of the loss of one diesel generator following LOOP. The HPI flows for the HPI line break are split based on operaior action to throttle the high flow after ten minutes. The HPl flows to the broken and intact loops for pump discharge, CFT line and HPI line breaks are listed in Table 3-10 through Table 3-12. Table 3-13 provides the LPI flow distribution used in analyses of pump discharge breaks and the HPI line b'reak. The CFT break prevents one CFT and one LPI pump from injecting ECC, while the single failure disables the other tr)l pump. Therefore, no LPI liquid is available for the CFT line break. The reactor trips on a low primary system pressure of 1795 psia with a 0.6-second delay before control rod insertion begins. A loss of offsite power (LOOP) is assumed to occur at the time of reactor trip, causing the reactor coolant pumps to coast down. The main feedwater pumps also coast down on the LOOP, and the main feedwater flowrate is decreased to zero within 14 seconds after the LOOP. ESAS is triggered when the primary system pressure drops below 1495 psia. A 35-second delay time is assumed before HPI flow begins. At 120 seconds after reactor trip, EFW is actuated and fills the secondary side to a nominal level of 50 percent (20.7 ft) on the operate range, to simulate the natural circulation level. A manually-throttled error-adjusted EFW flow rate of 200 gpm is modeled. This EFW flow value is reasonable based on EOP guidelines which instruct the operator to larget EFW fill at the minimum fill rate. For analyses of breaks in the reactor coolant pump discharge piping, including the HPl line breaks, each CFT has an initial liquid inventory of 985 ft3 and is pressurized to 580 psia. As discussed in Section 5.1.2.2, an initial CFT pressure of 650 psia and an initial 3 liquid inventory of 895 ft is used for the CFT line break analysis. Each CFT line is modeled to contain an additional 40 ft3 of liquid. The LPI flow is also pressure-dependent, and the LPI pumps are activated by a low-low primary system pressure of 355 psia, with a delay of 10 seconds after the low-low setpoint (355) and.35 seconds after the low setpoint (1495). The base analysis assumptions include operator actions as listed in Table 3-8, however all actions specified in the plant' specific EOPs should be performed to successfully mitigate the consequences of the LOCA. Some operator actions are credited at a specific time to minimize the severity of the postulated event on the core cooling. For the SBLOCA analyses, at 20 minutes following a loss of core exit subcooling, the operator is assumed to reset the secondary level setpoint from 50 percent to 65 percent
. on the operate range (24.3 feet). The 65 percent level is a conservative representation of the in-plant inadequate subcooling margin (ISCM) setpoint less instrumentation 74
Framatome Technoloales. Inc. 86-5002073-02 errors. Once the secondary levels reach the ISCM setpoint, the EFW is throttled to maintain this level. One of the two HPI line break analyses incorporates an additional operator action. Manualinitiation of HPlis assumed to occur 600 seconds after the loss of subcooling margin occurs at the core exit. This assumption simulates the operator's response to the operating procedural requirement to initiate HPl flow upon loss of the RCS subcooling margin (Section 5.2.5.2 elaborates further on this assumption). 5.2.7 SBLOCA Transient Proaression The transient progression for SBLOCAs is summarized here to identify the key phenomena and controlling thermal-hydraulic behavior during each phase of the event. The discussions of the break spectrum, which begin in Section 5.2.3, refer to this information to avoid repetition. An SBLOCA generally propsses through five phases: (1) subcooled depressurization (2) reactor coolant pump and loop flow coastdown and natural circulation, (3) loop draining, (4) boiling pot, and (5) refill and long-term cooling. The subcooled depressurization phase begins at the leak initiation. This phase is characterized by the
' period of time before the RCS begins to saturate and voids begin to form in the RV upper head and hot leg U-bends. During this period, the pressuri7.er will begin to empty, the RCS will depressurize to the low RCS pressure reactor trip setpoint, and the turbine will trip, With the assumption of a loss of offsite power coincident with reactor trip, the MFW pumps and RC pumps will trip and EFW will be initiated following a 120-second delay.
Following the RCP coastdown, the RCS flow tends to evolve to a natural circulation flow condition. The energy generated by the core is transferred by convection to the steam generators during the flow phase. The continued loss of the RCS liquid inventory allows steam voids to form in the upper reactor vessel head and the upper hot leg U-bends. Natural circulation ends when the U-bend steam void displaces the hot leg mixture levels below the U-bend spillover elevation. Flow is usually interrupted first in the hot leg containing the pressurizer surge line connection, because of the additional flashing of the saturated pressurizer liciuid thai enters during the subcooled depressurization. Near the end of the flow phase, alternating periods of RCS repressurization can cause intermittent spillovers of hot-leg liquid into the steam generator primary region. With the interruption of the RCS loop flow, the loop-draining phase begins. As the entire RCS approaches saturated conditions, the onset of subcooled and satureted nucleate boiling occurs in the core because of the high decay heat levels and the RCS depressurization. The flashing within the hot legs increases the size of the voids in the U-bends and eventually interrupts RCS flow and decreases the primary-to-secondary heat transfer. For the larger SBl.OCAs, the RCS will continue to depressurize as the loops drain. For smaller breaks, however, the reduced heat transfer can interrupt the RCS depressurization. Also for these smaller breaks, the volumetric expansion of the l 75
Framatome Technoloales. Inc. 86-5002073-02 RCS, due to continued steam formation, can exceed the volumetric discharge from the break, causing the RCS pressure to temporarily stabilize or increase, in the reactor: vessel, the steam void in the upper head displaces enough liquid to uncover the reactor vessel internals vent valves (RWVs), creating a manometric
' imbalance between the core and the downcomer. The imbalance forces the RVVVs to open and pass _ steam into the reactor vessel downcomer. The downcomer steam vo'ume grows until the cold leg nozzle is exposed to steam. As soon as the downcomer liquid level decreases below the cold leg nozzle spillunder elevation, a steam venting ,
path develops from the core through the RVVVs to the cold leg break, enhancing the RCS depressurization. During the loop draining phase, the steam voids that develop in the U-bends can become large enough that the primary liquid level is displaced into the steam generator tube region below the EFW nozzles. The improved primary-to-s?condary heat transfer can then be restored, through condensation on the tubes wetted by the EFW. This heat transfer process within a once-through steam generator (OTSG) is referred to as boiler-condenser mode (BCM) cooling. When BCM cooling takes place near the location of the EFW nozzles, it is referred to as high-elevation BCM cooling. If high-elevation BCM cooling occurs, the RCS depressurization rate will be increased. Later in the loop draining phase, a different form of BCM cooling can occur if the RCS tuba liquid level decreases below the secondary liquid level. This cooling process is referred to as pool BCM cooling, and will continue if (1) RCS condensation and ECCS injection do not cause the RCS liquid level to inersase above the secondary level, (2) the secondary fluid temperature is maintained below the temperature of the steam on the primary side of the OTSG tubes, and (3) the secondary liquid level is high enough that the secondary OTSG thermal center remains several feet above the RCP spillover elevation. For the smaller breaks, the combination of leak flow (with upper-RV venting through the RVVVs), BCM cooling, and HPI cooling will cause the RCS pressure to again decrease. Also during the loop draining phase, the reactor vessel outlet annulus mixture level will decrease to the hot leg nozzle spillunder elevation. If the top of the hot leg nozzles void, steam will flow up the hot leg riser section, and liquid fram the hot leg risers will drain back into the vessel. This hot leg draining allows the mixture level in the outlet annulus to remain near the top of the hot leg nozzle until the hot leg level drops Mio the RV exit nozzle horizontal piping. - After the hot legs empty, another path for the direct venting of steam to the break can be opened if the loop seals in the RCP suction piping are cleared. The suction piping of the four RCPs contains a large total volume and the spillunder elevation for this piping is approximately 23 feet below the top of the core. For the larger SBLOCAs, the RCS depressurization can be rapid enough to cause significant flashing in the suction piping, causing the liquid level to decrease below the suction pip:ng spillunder elevation. The loop seals will then be clear, creating another steam relief path, in addition to the path through the RVVVs. I 76
Framatome Technoloaies. Inc. 86-5002073-02 l When loop draining ends, the break site void fraction will be based on core steam plus I broken loop HPl. At that point, the only RCS liquid available for core cooling is the liquid remaining in the reactor vessel and the ECCS flow from the intact loops. This portion of the transient is defined as the " boiling pot" phase. The increased void fraction at the break will further ir. crease the RCS depressurization rate. The reactor vessel levels will continue to decrease, however, if the ECCS injection cannot match the vactor vessel liquid loss from flashing, decay heat, and passive metal heat. The break flow allows the RCS depressurization to' progress until either the CFT pressure is reached or the HPl flow rate exceeds the liquid loss rate, allowing the RCS to refill to the break elevation. Before either of these conditions occur, the mixture levels may descend into the core heated region resulting in a heat up of the fuel cladding in the uncovered portion of the core. The clad temperature increases calculated for the upper elevations are conservative , because the assumed power. shape in the model places the peak power at the 9.536-ft core elevation. This powe'r shape bounds the positive imbalance limits for core operation. During the period of partial core uncovering, the clad may swell and possibly rupture if the clad temperatures exceed 1300 F. The potential for clad rupture is increased in the SBLOCA analytical model by assuming an initial internal pin pressure typical of the end of fuel life (EOL). If clad rupture is calculated, a sensitivity study is needed to show that the calculated PCT will bound the fuel pin conditions at any time-in-life condition. An SBLOCA transient analysis is normally terminated at some point after the entire core is refilled and the cladding temperatures returned to within a few degrees of RCS saturation temperature. For the level to increase, core inflow (ECCS plus SG condensate) must exceed the liquid loss rate. _ Continued RCS depressurization permits higher ECCS injection rates that hastens core refill. The additional ECCS flow assures that the core can be kept covered. Once the core has been completely quenched, the analytical results are checked to ensure a path to long-term cooling is established. For long-term cooling io be assured, the HPl flow and/or LPl flow must match core boiling due to decay heat and wall metal heat plus flashing. When long-term cooling is assured, the LOCA analysis is terminated. The results of the specific small breaks ; analyzed are provided in the following sections. 4 77
Framatome TQchnoloaies. Inc. 86-5002073-02 5_2] Break Soectrum Analysis at RCP Discharae The specific break sizes analyzed at the RCP discharge location are listed at.the
- beginning of Section 5.2. The results for the SBLOCA break spectrum analyses are summarind in Tables 5-1 through'5-3. These analyses used the base case described in Section 5.2.1, except as noted, with the indicated break areas. The maximum calculated PCT was 1412 F for the 0.05-ft2 break. Plots of primary pressures, collapsed liquid levels, peak cladding temperatures for the RCP discharge analyses are located in Figures 5-1 through 5-9.
5.2.3.1 Small SBLOCAs at RCP Discharge The smaller LOCAs, between approximately 0.01 ft2 and 0.08 ft 2, present the greatest challenge to the HPI system to replace lost liquid inventory before significant fuel or clad damage occurs. These break sizes are not large enough to rapidly depressurize the RCS to the CFT fill pressure. Consequently, the HPl flow reaching the reactor vessel must be sufficient to match the core decay heat. 2 The smallest break size analyzed was 0.01 ft . Breaks smaller than 0.01 ft2 w ll produce a slower RCS inventory loss rate and will be less challenging to the capacity of the ECCS to : provide adequate core cooling. This break size demonstrates the effectiveness of continuous SG heat removal via EFW preservation and level control. The smaller break sizes will also allow more time for mitigative operator actions, such as a manual depressurization of the secondary system, restoration of an additional HPl pump to service, or initiation of HPl/LPI piggyback operation. None of those potentially beneficial operator actions were assumed in these analyses. The benefits of such actions are readily apparent after examining the results for the break sizes in this calsgory.
-Even for these relatively small break sizes, the RCS depressurization and voiding quickly interrupted loop flow. With primary-to-secondary heat transfer interrupted, an RCS repressurization was predicted for the 0.01- and 0.08-ft2 breaks in this category.
For all the break sizes in this category, the RCS pressure response caused delays in ESAS actuation, and, consequently, in the start of HPI flow. All of the small SBLOCA cases, except for the 0.01-ft2 break, experienced partial core uncovering. The duration of core uncovering was a direct result of the sustained high RCS pressure, which limited the HPl flow and delayed the start of CFT injection. For the seven small SBLOCA cases analyzed, the RCS pressure did not depressurize sufficiently to allow LPI flow ic enter the reactor vessel. In each case, at the time the analysis was ended, the core was completely recovered, the downcomer level was increasing', and the HPI flow matched flashing plus core boil-off due to decay heat and wall metal heat contributions. These conditions confirmed that the HPl flow was 78
p 1 Framatome T@chnoloaies. Inc. 86-5002073.02 l adequate to prevent significant cladding temperature excursions and to ensure long-term cooling via preservation of HPI and EFW. 5.2.3.2 Intermediate SBLOCAs at RCP Discharge The intermediate small breaks were analyzed in the RCP discharge piping with break areas of 0.09,0.10, and 0.15 ft2. The . aults of these breaks are summarized in Table
.5-2.
The intermediate break sizes caused the RCS to depressurize faster, and to enter into , the boiling pot mode sooner, than the smaller breaks discussed in Section 5.2.3.1. The intermediate breaks continuously depressurized the RCS after the forced flow phase ended, although the depressurization rate slowed temporarily. The 0.09- and 0.10-ft case experienced a slight RCS repressurization, however it did not significantly delay ESAS actuation. The RCS depressurization rate increased after the steam venting path through the RWVs was established. ESAS was actuated relatively early in the transient. The RCS depressurization rate also was not significantly affected when the RCS pressure decreased below the secondary pressure and secondary-to-primary (reverse) heat transfer began. The intermediate breaks were able to discharge fluid mass, energy, and volume so effectively that they cooled and depressurized both the primary and the secondary systems. Even though the RCS depressurized fairly quickly, allowing the CFT injection to begin relatively early in the event, partial core uncovering _ did occur for the intermediate breaks. As a result, clad heatup occurred in the upper core regions. When the analysis of the OA9,0.1, , and 0.15-tt abreaks ended, the low-low pressure , setpoint had been reached but the RCS pressure remained above the LPI shutoff head.. i Furthermore, the pressure was decreasing and sufficient liquid remained in the CFTs to sustain CFT injection until LPI flow could begin. Therefore, the analyses confirmed that, for intermediate breaks, the ECCS capacity was sufficient to recover the core and to provide adequate long-term cooling. 5.2.3.3 Large SBLOCAs at RCP Discharge - Three large SBLOCAs, with break areas of 0.30,0.50, and 0.75 ft2, were also analyzed. For these large SBLOCA sizes, the effects of the break dominated other factors (such ! as the timing and magnitude of EFW flow, BCM cooling, and reverse heat transfer) that could potentially affect the RCS depressurization rate.
. For all three cases, the low-pressure reactor trip setpoint was reached within 1.5 seconds after break opening, and ESAS was actuated within the first 16 seconds. The transients progressed relatively quickly through the SBLOCA phases identified in Section 5.2.2. In comparison with the smaller breaks, the HPI, CFT, and LPI delivery 79 u
Framatome Technoloaies. Inc. 86-5002073-02 during the large SBLOCAs was enhanced by the lower transient RCS pressures, i.e. the actuation times were earlier and the flow rates were higher. CFT injection started within ten seconds after the onset of. core uncovering in both cases. The rapid RCS depressurization rates caused by the large SBLOCAs produced significant core voiding. The improved ECCS performance significantly shortened the duration of core uncovering and produced lower PCTs. The calculated PCTs for the 0.30 ,0.50 , and 0.75-ft2breaks were well below the values calculated for the limiting breaks in the small SBLOCA category. LPI flow was initiated for the three cases ~, but well after the HPI and CFT flow had limited the clad temperature . increase and completely refilled the core. However, while the HPl and CFTs had demonstrated adequate short-term core cooling, the availability of the LPI ensured that adequate long-term cooling would continue after the CFTs emptied. 5.2.4 Soecial Breaks A rupture in either a CFT line or an HPI line, between the first check valve and the RCS connection, is categorized as a special break. These breaks present unique challenges to the core cooling capacity of the ECCS because they result in reduced ECCS flow to the RCS. Therefore, these breaks are treated as special cases and subjected to specific analyses, which are described in the following sections. 5.2.4.1 CFTLine Break For the CFT line break, a more severe degradation of the ECCS capacity must be considered than for a CLPD break. During a CFT line break, the break location prevents one CFT and one LPl train from injecting into the reactor vessel. The other LPI train and one HPl train are assumed to be unavailable due to a single failure. Therefore, only one CFT and one train of HPl remain available for core cooling. However, since the HPI piping remains completely intact, all of the flow from the available HPl pump is able to enter the RCS. As was previously noted in Sections 5.1.1.5 and 5.1.2.2, for the CFT line break, the nominal CFT line resistance, minimum CFT inventory and maximum fill pressure were modeled. The break was modeled to occur at the connection of the CFT line nozzio to the reactor vessel. The break area was limited to 0.44 ft2by the cross-sectional area of the nozzle insert. The results of the CFT line break analysis are summarized in Table 5-4. Plots shows system pressures, collapsed liquid levels, peak cladding temperatures and core mixture levels are located in Figures 5-10 through 5-13. The relatively large break size caused a rapid RCS depressurization. The timing of CFT actuation and the available HPl flow were sufficient to maintain core covering throughout the event. The results of this case also demonstrated that one CFT and one HPI pump provide adequate ECCS flow to mitigate the consequences of a CFT line break. 80
Framatome Techr.oloaiejL Inc. 86-5002073-02 5.2.4.2 HPI Line Break The HPl line break location was modeled in an HPl line just upstream of the HPI nozzle, and the thermal sleeve in the nozzle was assumed to be blown out coincident with the break opening. The break area is limited by tie cross-sectional area of the pipe without z the thermal sleeve (0.02463 ft ). HPI flow from the broken line was assumed to spill into the containment and was, therefore, not modeled. For the CLPD breaks and the CFT line break, HPl was modeled to initiate automatically upon actuation of the low-RCS-pressure ESAS signal plus a 35-second delay. However, because of the slow RCS depressurization during the HPl line break, HPl was not available until ~1251 seconds. ' A second case was run that modeled operator action to manually start the available HPl pump at 10 minutes after the loss of core exit subcooling providing HPI flow at ~675 seconds. Both cases assumed that the other HPl pump was unavailable for the entire transient due to a single failure, and the makeup pump was also unavailable. The results of both HPl line break analyses are presented in Table 5-5. Plots of system pressures, collapsed liquid levels, clad temperatures and core mixture levels are located in Figures 5-14 through 5-17 (manual HPl initiation) and Figures 5-18 through 5-21 (HPI initiation on low-RCS-pressure ESAS plus delay). The timing of CFT actuation and the avai!able HPl flow were sufficient to maintain core covering throughout the event. Therefore, the HPI system has sufficient capacity to provide adequate short- and long-term core cooling for the HPl line break. 5.2.5 Recent SBLOCA issues ; Recent SBLOCA issues pertain to the energy deposition factor, the assumption of manual ESAS actuation, MSSV lift tolerance and EFW flow rates. 5.2.5.1 EDF i l l The energy deposition factor is defined as the energy absorbed (thermal source) in the fuel pellet and clad divided by the energy produced by the pellet (nuclear source). EDF = Pm,mi.oua/ Pnuew.rsouro. The BWNT LOCA EM specifies a steady-state and transient EDF of 0.973 for SBLOCA analyses. New methods and predictions for the EDFs appropriate for use in LOCA analyses at various times in life have recently been evaluated by FCF (Reference 9). These calculations do not totally support the 0.973 or 0.96 values for high burnup, low power fuel or fuel that may be surrounded by higher power fuel. FTI's current regulatory position on the energy deposition factor concludes that the EDF that was used for the 81
Framatome T@chnoloaies. Inc. 86 5002073-02 SBLOCA EM analyses was calculated with methods that were state-of-the-art at the time l the calculations were performed. FTl clearly stated the EDF values used in the EM and 1 recognizes that conservatism in the Appendix K methodology are more than adequate to address the slight increase in fuel pin thermal energy deposition for SBLOCA analyses.
'In anticipation of potential future changes imposed by the NRC, a study was performed (Reference 7) to determine the effect of using a transient EDF of 1.0 in the BWOG SBLOCA analyses. This study was performed for a 0.05 ft2 CLPD break. The results ,
show an increase in the PCT of 45 F. While the PCT increased, the EDF change did not I make a significant alteration in the timing of the transient. Therefore, the use of a transient J EDF of 0.973 or 1.0 does not make a significant difference in the SBLOCA transient. 5.2.5.2 Assumption of ManualHPI Actuation The purpose of this section is to clarify the assumption made regarding manual HPI 2 actuation for the 0.02463-ft HPl line break 10 minutes after the loss of subcooling margin (LSCM). This assumption was made only in tha ANO-1/TMI-120 percent tube plugging HPl line break and assumed actuation of HPI and isolation of letdown. The HPI line break was small enough that RCS repressurization began before the low RCS pressure ESAS actuation trip was obtained. Based on emergency operating procedure (EOP) guidelines, the operators are instructed to initiate and control HPl on the loss of core exit subcooling margin. The manual initiation of HPI is thus highly probable for the smaller SBLOCAs because of the slower evolution of the transient. For those smallest breaks that are not explicitly analyzed (partial HPl line and CLPD < 0.01 ft 2), manual initiation of HPl at 10 minutes after the LSCM assures that the consequences of these breaks are less severe than those break sizes that are explicitly analyzed. FTl re-evaluated the HPI line break and waited for automatic ESAS actuation to occur when the EFW boiler-condenser mode of heat transfer adequately depressurized the RCS to the low RCS pressure ESAS setpoint and subsequently initiated HPl. The reanalysis concluded that even with the delay in HPI flow and lower mixture level, the PCT increased only 2 F and the whole channel quench was reached at nearly the same time. Plots of system pressure, collapsed liquid levels, clad temperatures and core mixture levels for the case with no operator action to initiate HPl are located in Figures 5-18 through 5-21. 5.2.5.3 Pump Type. l The Reference 7 analyses were reviewed in response on PSC 1-99 and found to be based on the Bingham pump type. However, neither plant operates a Bingham type RCP. TMI-1 operates Westinghouse type RCPs, while ANO-1 operates Byron-Jackson type RCPs. The effect of the pump type and pump two-phase degradation for Oconee were investigated in Reference 33 for SBLOCA. The investigation examined the Westinghouse versus Bingham pump type and the ANC/M1 versus RS/M3 two-phase 82
[f l
- Framatome Technoloaies. Inc. , 86-5002073-02 l- pump degradation. The study concluded that the selection of the pump type and two- - phase degradation is not a sensitive input for the SBLOCA transient because the pump flow approaches zero during the first few hundred seconds and does not sufficiently )
influence the core flow or vessel inventory during the transient. Therefore, the BV\/OG 20% tube plugging SBLOCA analyses results reported in Reference 6 remain valid. 3 l 5.2.5.4 MSSVLift Tolerance 1 l Questions have been raised as to the appropriateness of modeling the MSSV setpoint based on the nominal lift setpoint instead of the 3 percent lift tolerance. There are three points that support the use of the nominal MSSV setpoint of 1065 psia. First, the Evaluation Model (Reference 4) input guidelines in Table 9-2 of Volume 11 specify that nominal operation design levels will be used for system pressures. Second, studies , supporting the EM, which were approved by the NRC, used nominal iASSV setpoints. l Third, information related to more detailed MSSV modeling was examined in Reference l 6. To model the hysteresis effect on the valve open and close setpoints for the four MSSV banks, the MSSV setpoint pressure, plus 3 percent, was used to lift the valve and the nominal MSSV setpoint was used to close .the valve (without crediting blowdown below the nominal setpoint). The results of that examination showed that the MSSV setpoint t.id not significantly influence the RCS pressure response of the most limiting break size. Therefore, it was concluded that setting the MSSV setpoint at the nominal value of 1065 psia, without modeling accumulation or crediting blowdown below
- l. the nominal valve setpoint, is representative of the MSSV behavior and adequate for l SBLOCA analyses.
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Framatome Technoloaies. Inc. 86-5002073-02 Table 5-3. Summary ofIntermediate SBLOCA Results 0.9 ft2 to 0.15 ft2 Parameter 0.09 0.10 0.15 Break Opens, s 0.0 0.0 0.0 LP Reactor Trip, s 9.36 8.10 4.56 ESAS Trip, s 121.18 56.30 32.62 Subc. Blowdown Ends, s - 30 - 30 - 15 HPl Flow Begins, s 156.18 91.32 67.62 EFW, s 129.38 128.12 124.58 Loop Draining Begins, s ~330 ~350 -180 Boiling Pot Begins, s -480 - 440 -300 CFT Actuation, s ~1000 - 860 -530 RV Refill Begins, s - 1500 -1220 -535 LPI Flow Begins, s N/A N/A N/A HC Peak Clad Temp, F 1354 1334 922 Segment Number 19 19 20 Time, sec 1266.7 1045.1 688.32 No. of Dry Segments 4 4 3 AC Peak Clad Temp, F 1079 972 688 Segment Number 19 18 18 1 Time, see 1214.8 1215.7 0.2 j No. of Dry Segments 5 4 3 l Max Local Oxidation, % < 0.55 < 0.55 < 0.55 Whole Core H2 Gen, % < 0.10
< 0.10 < 0.10 Calculation Stopped, sec 4728.3 4082.5 1906.8 85
Framatome Technoloaies. Inc. 86-5002073-02 l l Table 5-4. Summary of Large SBLOCA Results 0.3 ft2 to 0.75 ft2 Parameter 0.30 0.50 0.75 Break Opens, s 0.0 0.0 0.0 LP Reactor Trip, s 1.14 0.88 0.76 ESAS Trip, s 15.82 9.68 7.58 Subc. Blowdown Ends, s -9 -9 - 11 HPl Flow Begins, s 50.82 44.68 42.58 EFW, s 121.16 120.90 120.78 Loop Draining Begins, s ~ 75 - 42 l
~ 25 '
Boiling Pot Begins, s -110 - 75 - 50 CFT Actuation, u ~225 -125 ~ 77 RV Refill Begins, s -235 -125 ~ 71 i LPI Flow Begins, s -500 -235 - 143 HC Peak Clad Temp, F 790 845 860 Segment Number 20 19 17 Time, sec 270.38 150.24 98.42 No. of Dry Segments 3 4 8 AC Peak Clad Temp, F 688 705 738.41 Segment Number 18 19 18 Time, sec 0.1 148.68 104.26 No. of Dry Segments 4 5 8 l l Max Local Oxidation, % < 0.55 < 0.55 < 0.55 l Whole Core H2 Gen, % < 0.10 < 0.10 < 0.10 Calculation Stopped, sec 510.44 279.30 191.20 l 86
Framatome Technoloains. Inc. 86-5002073-02 Table 5-5. Summary of HPI and CFT Line Breaks Parameter HPI Line Break HPI Line Break CFT Line Break (Note 1) (Note 2) Break Size, ft2 0.02463 0.02463 0.4/ Break Opens, s 0.0 0.0 0.L LP Reactor Trip, s 41.16 41.16 0.92 ESAS Trip, s 1216.58 1177.02 10.82 Subc. Blowdown Ends, s - 50 - 50 ~ 10 HPI Flow Begins, s 1251.60 675.18 45.82 (Note 1) (Note 2) EFW, s 161.16 161.16 120.94 Loop Draining Begins, s -250 -250 - 45 Boiling Pot Begins, s -2000 -2250 - 90 CFT Actuation, s N/A N/A ~125 RV Refill Begins, s -4800 -5250 -2600 LPI Flow Begins, s N/A N/A N/A HC Peak Clad Temp, F 1299 1297 715 Segment Number 19 19 18 Time, sec 5036.4 5570.1 0.00 No. of Dry Segments 3 3 3 AC Peak Clad Temp, F 1035 869 688 Segment Number 19 19 18 Time, sec 4480 5303 0.1 No. of Dry Segments 4 3 3 Max Local Oxidation, % < 0.55 < 0.55 < 0.55 Whole Core H2 Gen, % < 0.10 < 0.10 < 0.10 Calculation Stopped, sec 8000.0 8000.0 4000.0 Note 1: No operator action was taken to initiate HPL Note 2: Operator action to manually initiate HPI was assumed at 600 seconds after the loss of subcooling margin (- 75 seconds). 87
86-5002073-02 Figure 5-1; 0.01 - 0.06 FT2 CLPD SBLOCA BREAKS FOR 20% SGTP l COMPARSION OF PRIMARY PRESSURES. ) 2400 LEGEND _ ,160 J 0.01 FT2, CLPD SBLOCA 2000, -- - - 0.03 FT2, CLPD SBLOCA , ,
-------- 0.04 FT2, CLPD SBLOCA
- ---------- 0.05 FT2, CLPD SBLOCA -
.g, ~~~~~~ - 0.06-FT2, CLPD SBLOCA 120 ,
1600< '
\ l j . ,'\\ '\
\
l f 1200' 's, ' g
's,,\ ' \ . g '~~ . ~ . , _
80 v> s, ss .s
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40
. ~ ~ ...
. . .--. -. ~~. - -. . . . ~ . . . . ~ .
400< . . . _ ~ ~. ...
~-..~.._
0 0 0 1000 2000 3000 4000 5000 6000 TIME, SECONDS Figure 5-2: 0.07 - 0.10 FT2 CLPD SBLOCA BREAKS FOR 20% SGTP l COMPARSION OF PRIMARY PRESSURES. ! 2400 LEGEND ,160 0.07 FT2, CLPD SBLOCA 2000, - --- 0.08 FT2, CLPD SBLOCA ,
-----~~ 0.09 FT2, CLPD SBLOCA
---------- 0.10-FT2, CLPD SBLOCA
-120 1600' 1200' .. 80
- g. g !
800< '. 's
\., '\ 40 400' '
' * - ~EiEi'3'.Gi- - .:::..w._.._ --
0 0 0 1000 2000 3000 4000 5000 6000 TIME. SECONDS 1 88
86-5002073-02 Figure 5-3: 0.15 - 0.75 FT2 CLPD SBLOCA BREAKS FOR 20% SGTP COMPARSION OF PRIMARY PRESSURES.
.2400 LEGEND -
160 0.15-FT2, CLPD SBLOCA
~*- - 0.30 FT2, CLPD SBLOCA ,
2000- - --- 0.50 FT2, CLPD SBLOCA
---------- 0.75 FT2, CLPO SBLOCA 120 1600<
1200<. .80
- \.
800- it \.
- \ 40
; 'g 400< i, g \,
', \ s
, s 0
0 1500 1750 2000 0 250 500 750 1000 1250 TIME, SECONOS Figure 5-4: 0.01 - 0.06 FT2 CLPD SBLOCA BREAKS FOR 20% SGTP . COMPARSION OF REACTOR VESSEL COLLAPSED LIQUID LEVELS. 35 LEGEND i'. . 10 C.01-FT2 SBLOCA t 30< - - - -. 0.03.FT2 SBLOCA ' {-
-- --- 0.04-FT2 SBLOCA
---------- 0.05-FT2 SBLOCA
--- --- ~~~ 0.06-FT2 SBLOCA .- 8 25< }.
2 k-20' N. __ . , _ _#-- __
-- - - - ^
- '- 6
*%. _ _ _ ., ggty' ash. 4 Va-s s. . .="*-- TCP OF MEATto CORE 10<
2 6< 0
.0- 3000 4000 5000 6000 0 1000 2000 TIME, SECONDS
~'
89
m q l , 86-5002073-02 ; Figure 5-5: 0.07 0.10 FT2 CLPD SBLOCA BREAKS FOR 20% SGTP l COMPARSION OF ~ ACTOR VESSEL COLLAPSED LIQUID LEVELS. l 35 LEGEND (. -10 I 0.07-FT2 SBLOCA ! 30' -.-.-. 0.08 FT2 GBLOCA ' l
---- 0.Q9 FT2 SBLOCA I
---------- 0.10-FT2 SBLOCA 8 I 25-
]
t :
-f 20< -
,8 5 \\ 5 8 is. yx ,,,.... -p-;^g <
8 ; 8 hx ...-..-e 4 8 g.g.@ _ . TOP OF HEAltD CORE I 2 5' O O O 1000 2000 3000 4000 5000 6000 TIME, SECONDS Figure 5-6: 0.15 - 0.75 FT2 CLPD SBLOCA BREAKS FOR 20% SGTP COMPARSION OF REACTOR VESSEL COLLAPSED LIQUID LEVELS. 3 LEGEND
}. 10 0.15-FT2 SBLOCA 30- ,
---.-. O.30 FT2 SBLOCA '
k ------ 0.50-FT2 SBLOCA p 0.75-FT2 SELOCA 8 25 ), e il = 20-. , 6 l'-n , / 1 I
,, ,v --
5< \j 0 0 0 250 500 750 1000 1250 1500 1750 2000 TIME. SECONDS 90
ft ) 86-5002073-02 Figure 5-7:0.01 - 0.06 FT2 CLPD SBLOCA BREAKS FOR 20% SGTP COMPARISON OF PEAK CLADDING TEMPERATURES. 2000 LEGEND 1 0.01.FT2, CLPD SBLOCA, Seg 18 , 17,50 -.- - 0.03-FT2, CLPD SBLOCA, Seg 19 1200
----- -- 0.04 FT2, CLPD SBLOCA, Seg 20
---------- 0.05.FT2, CLPD SBLOCA, Seg 20 1500,
- ... . . 0.06-FT2, CLPD SBLOCA, Seg 19 j' "'s - \ 1000
,,,, jD, 50- ,
,j Qf3, y,,,, ,
~ '
1000< ,e' ,
\ '\- 800
- :.,... , e
. l 750< e' ,
{ h 600
~~~'
E00-
....-......=.....=...b-.-
l l L..._ 250< 400 i 0 0 1000 2000 3000 4000 50C0 6000 7000 TIME, SECONDS i Figure 5-8: 0.07 - 0.10 FT2 CLPD SBLOCA BREAKS FOR 20% SGTP l COMPARISON OF PEAK CLADDING TEMPERATURES. l 2000 LEGEND l l 0.07-FT2, CLPD SBLOCA, Seg 19 , l 1750,
----- 0.08-FT2, CLPD SBLOCA, Seg 19 1200 ;
------ 0.09-FT2, CLPD SBLOCA, Seg 19 l 1500,
---------- 0.10-FT2, CLPD SBLOCA, Seg 19 ,
^,^ 1000 m e
- w
*[' 4f,b g
,I ,
\
- I ,'v# i.,, A '
a 1000< .
- 800 750<
500< (! ' ' ' - ' e
,.l S
'-wwwwww 250, 400 0
0 1000 .2000 3000 4000 5000 6000 TIME, SECONDS 91
e
.y
. 86-5002073-02 Figure 5 9: 0.15 - 0.75 FT2 CLPD SBLOCA BREAKS FOR 20% SGTP COMPARISON OF PEAK CLADDING TEMPERATURES.
_ LEGEND
- 0.15-FT2, CLPD SBLOCA Sep 20 ,
1750 - - - 0.30 FT2, CLPD SBLOCA, Seg 20 1200
----- 0.50-FT2, CLPD SBLOCA, Seg 19
--------- 0.75-FT2, CLPD SBLOCA, Seg 17 1500 1000 M
1000-l1250-800 750- , [ 500<
.;~..
t.. 250- 400 0 0 250 500 750 1000 1250 1500 1750 2000 TIME, SECONDS 1 l 1 i l l 92
r: 06=5002073-02 Figure 5.10: 0.44 FT2 CFT-LINE SBLOCA BREAK FOR 20% SGTP COMPARSION OF 9YF"EM PRESSURES. 2400--- - LEGEND 160 Hot Leg 2000- - - - Intact SG Secondary .
----- Broken SG Secondary CFT 120 1600-h $
1200' 80 g !?> E E 800- < j
.. )
\ 40 I 400< *%
%v= wn v ,ru v ars== =muna v .
................................=...........................
0- o j 0 500 1000 1500 2000 2500 3C00 3500 4000 ' TIME, SECONOS - 1 Figure 5-11: 0.44 FT2 CFT-LINE SBLOCA BREAK FOR 20% SGTP 1 REACTOR VESSEL COLLAPSED LIQUID LEVELS. ! LEGEND 10 Downcomer 30- .-.-. Coro Composite 8 , 25< < t .! 2 ; 6 I 20 ] ,
$ 15- k. <
8 1 t A .. -::: k CD CORE 4 8 V/ %.mW* *-- -. 10' . . . . "* - 2 5< 0 0 0 500 1000 1500 2000 2500 3000 3500 4000 TIME, SECONDS 93
l.: 86-5002073-02 Figure 5-12: 0.44 FT2 CFT-LINE SBLOCA BREAK FOR 20% SGTP HOT CHANNEL PEAK CLADDING TEMPERATURES. 2000 LEG END _ _ 1750' Seg 21,11.2 - 12.0 FT j - - - Seg 20,10.4 - 11.2 FT 1200 l ----- Sep 19, 9.6 10.4 FT 1500,
---------- Seg 18, 8.8 9.6 FT
-- ~~ - Seg 17, 8.0 8.8 FT 1000 1250'
- 1000- '
. 800 1
600 500'
---_.-- ai 1
~~'
250' 400 0 0 500 1000 1500 2000 2500 3000 3500 4000 TIME, SECONDS Figure 5-13: 0.44 FT2 CFT-LINE SBLOCA BREAK FOR 20% SGTP COMPARISON OF HOT CHANNEL MIXTURE LEVELS. 20.0 ,,,, LEGEND 17.5, , Het Channel ,
- - - Average Channel 4.8 15.0- -
/
h* 12.S ' i, ' avrve ntAseu cune. q t 3.2 10.0' g 7.5- ' 1.6 5.0 ' i 2.5- ' 0.8 0.0 0.0 0 1000 2000 3000 4000 5000' 6000 7000 8000 TIME, SECONOS 94
86-5002073-02 Figure 514: HPl.LINE WITH OPERATOR ACTION SBLOCA BREAK FOR 20% SGTP COMPARSION OF SYSTEM PRESSURES. 2400 LEGEND 160 Hot Leg i
- - - Intact SG Secondary
'2000 ----- Broken SG Secondary
.......... CFT
-120 1600-5 f f 2 !
1200' 80 mew;y E 800- -
................................................................. 40 400' O' 0 O 1000 2000 3000 4000 5000 6000 7000 8000 1 i
TIME, SECONOS Figure 515: HPI-LINE WITH OPERATOR ACTION SBLOCA SREAK FOR 20% SGTP REACTOR VESSEL COLLAPSED LIQUID LEVELS. 28 ,JEGEND
- ' 8.0 Downcomer 24- ---.-* Core Composite 7.2 3.4 I
, y ,r r p ' ' 5.6 16' 'N - 4.8
~.-.C^- 4.0 g 12 g 3.2 ::l 8- 2.4 1.6 4'
0.8 o 0.0 0 1000 2000 3000 40C0 5000 6000 7000 8000 TIME, SECONDS 95
86-5002073-02 Figure 5-16: HPI-LINE WITH OPERATOR ACTION SBLOCA BREAK FOR 20% SGTP HOT CHANNEL PEAK CLADDING TEMPERATURES. 50
, EGEND 1200 Seg 21,11.2 12.0 FT '
1500- - -.-. Sep 20,10.4 - 11.2 FT
- ------- Seg 19, 9.6 - 10.4 FT
---------- See 18, 8.8 - 9.6 FT 1000 1250' ""'"*""~~~* See 17, 8.0 8.8 FT .%
,r--
1000< ~ i '
*S - 800 750-
/ ~~~ y
~ .f < 600 1
~ ~ ~ ~ ~
500 250 . 400 0 0 1000 2000 3000 4000 5000 6000 7000 8000 TIME, SECONDS Figure 5-17: HPI-LINE WITH OPERATOR ACTION GBLOCA BREAK FOR 20% SGTP COMPARISON OF HOT CHANNEL MIXTURE LEVELS. 0.0 LEGEND Hot Channel 17.5- - --- Average Channel, 4.8 15.0 < < W 3g,g, 1UP UP Mt:AIt;u UU l'O 3
' 3.2 _
- 10.0 - _
7.5, 2.4 5.0 , . 1.6 2.5< 0.8 0 1000. 2000 3000 4000 5000 6000 7000 800 TIME, SECONDS 96 u
i 86-5002073-02 Figure 518: 'HPI-LINE NO OPERATOR ACTION SBLOCA BREAK FOR 20% SGTP COMPARSION OF SYSTEM PRESSURES. 2400 LEGEND 160 Hot Leg
- --- Intact SG Secondary ]
2000 , li
- - - - Broken SG Secondary
.......... CFT 120 1600-
'1200' 80 I mw 1 c,v ,
h 800- ]
................................................................ 40 400<
0 0 0 1000 2000 3000 4000 5000 6000 7000 8000 TIME, SECONDS Figuro 5-19: -HPI-LINE NO OPERATOR ACTION SBLOCA BREAK FOR 20% SGTP REACTOR VESSEL COLLAPSED LIQUID LEVELS. 28 l LEGEND 8.0
, Downcomer 24< --- -. Core Composite 7,2
.. 6.4 20<
'} 5.6 g ,
'16< N 4,3
_ _ _ , . .Z ^ Re 4.0 g 12 l 3.2 8' < 2.4 1.6 4< < - 0.8 0 1000 2000 3000- 4000 5000 .6000 7000 800b' TIME, SECONDS gy - L ,
86 - 5 0020 73 -o g Figure 5-20: HPI-LINE NO OPERATOR ACTION SBLOCA BREAK FOR 20% SGTP HOT CHANNEL PEAK CLADDING TEMPERATURES. U50 _ LEGEND 1200 Seg 21,11.2 12.0 FT 1500< -.-.-. Seg 20,10.4 - 11.2 FT .
- - - " - ~ Sep 19, 9.6 10.4 FT
---------- See 18, 8.8 - 9.6 FT 1000 1250< " ~~~~/-):.--ligeg 17, C.0 8.8 FT
, . . _ , , n .u y.
.n./ u 1000< j 'V' % - 800
\.
750< l i L- % _)> I.' '600
* ~~~~~~~~~~~~~~~ ' '
500< 250< 400 0 0 1000 2000 3000 4000 5000 6000 7000 8000 TIME, SECONDS Figure 5-21: HPI-LINE NO OPERATOR ACTION SBLOCA BREAK FOR 20% SGTP COMPARISON OF HOT CHANNEL MIXTURE LEVELS. 0.0 ~~ LEGEND 17.5, Hot Channel ,
- - - Average Channel 4.8 15.0 < <
12.5 < 4 Ul" UI'* MtAI LU GU [ 3 10.0, , 3.2
-M*
7.5,
, 2.4 i
5.0 , , 1.6 ; l 2.5< < 0.8 I 0.0 0.0 0 1000 2000 3000 4000 5000 6000 7000 8000 TIME, SECONDS 98
Framatc.ne Technoloales. Inc. 86-5002073-02 6.10CFR50.46 Criteria The compliance of the LBLOCA and SBLOCA analysis results with the five 10CFR
' 50.46 ECCS criteria is discussed in this section.
'6.1 Peak Cladding Temperature The first criterion of 10CFR50.46 requires that the calculated peak cladding temperature remains below 2200 F. The peak cladding temperature results for the Mark-B9 fuel design are summarized in Table 2-1. The SBLOCA break spectrum PCT results are given in Table 2-2. For all LOCA cases, the PCT was calculated to be less than 2200 F.
The SBLOCA calculations are bounded by the LBLOCA calculations since the predicted PCTs are less than 1900 F. 6.2 Local Cladding Oxidation The second criterion 'of 10CFR50.46 requires that the maximum local degree of
. cladding oxidation not exceed 17 percent. Compliance with this criterion is obtained by evaluating the results of the calculation of peak cladding temperature. In the calculation, local cladding oxidation is computed as long as the cladding temperature remains above 1000 F.
The hot channel local cladding Woation values for the Mark-B9 LBLOCA analyses are summarized in Reference 32. In all cases, the hot channel local cladding oxidation was significantly less than 17 percent. For SBLOCAs (Reference 6), the results confirmed that the amount of local cladding o.:idation for small break LOCAs is also significantly < less than 17 percent. The oxidation values were calculated using a conservative initial oxide thickness to maximize the cladding temperature response due to. metal-water reaction. If a best-estimate initial oxide thickness were used to maximize the local cladding oxidation, the calculated peak cladding temperatures would be less than those reported in Section 2.
. Since the hot channel local cladding oxidation values for the BWOG analyses did not exceed 3 percent for any case, it is concluded that the 17 percent oxidation limit would not be reached even if a maximum initial oxide thickness were used in the large and small break analyses. These results and conclusions confirm that the amount of local
- cladding oxidation for the LOCAs analyzed meet the 10CFR50.46 local cladding oxidation requirement.
99
+ .
Framatome Technoloaims. Inc. 86-5002073-02 6.3 Whole-Core Oxidation and Hydrogen Generation The third criterion of 10CFR50.46 states that the calculated total amount of hydrogen generated from the chemical reaction of the cladding with water or steam shall not exceed 0.01 times the hypothetical amount that would be generated if all of the metal in the cladding cylinders surrounding the fuel reacted, excluding the cladding surrounding the plenum volume. As discussed in Section 6 of the LOCA evaluation model topical report (Reference 4, Volume 1), if the hot channel average oxidation increase is less than or equal to one percent, then the whole-core hydrogen generation increase is reported as less than one percent, and no additional calculations are performed. The results for the LBLOCA and SBLOCA analyses all predicted hot channel oxidation increases less than one percent, which precluded the need to perform the RELAP5/ MOD 2-B&W whole-core calculations outlined in Reference 4. For additional information, however, an equation was developed to estimate the whole-core hydrogen generation using a simple approximation of the detailed method outlined in Section 6.0 of the evaluation model. The estimates of the whole-core hydrogen generation for each of the LBLOCA limit analyses are summarized in Reference 32. The maximum whole-core hydrogen generation for the Mark-B9 fuel was estimated to be less than 0.4 percent for all cases. For the SBLOCA analyses (Reference 6), the maximum whole-core hydrogen generation rate was estimated to be less than 0.1 percent. The LOCA cases performed and documented in this report cover the entire range of possible power distributions and fuel burnup conditions that can occur in the plant. The maximum possible oxidation increase that can occur during a LOCA has been enveloped for the TMI-1 and ANO-1 plants, and a significant margin has been demonstrated to the one percent limit contained in the third criterion of 10CFR50.46. Therefore, this criterion is satisfied. 6.4 Core Geometry The fourth acceptance criterion of 10CFR50.46 states that calculated changes in core geometry shall be such that the core remains amenable to cooling. The RELAP5/ MOD 2-B&W PCT calculations directly assess the alterations in core geometry at the most severe location in the core that result from a LOCA. These calculations demonstrate that the fuel pin cooled successfully. As discussed in Section 7 of the BWNT evaluation model report (Reference 4), clad swelling and flow blockage due to rupture can be estimated based on NUREG-0630. For the TMI-1 and ANO-1 plants, the hot assembly flow area reduction at rupture is less than 72 percent for all large break LOCA cases, while the small break LOCA cases did not predict rupture. Furthermore, the upper limit of possible channel blockage, based on NUREG-0630, is less than 90 100 L _ _ _ _ _ - - - _ - - - - - - - - -
Framatome Technoloaims. Inc. 86-5002073-02 percent. Neither 90 percent blockage nor 72 percent blockage constitutes total subchannel obstruction. Since the position of rupture, in a fuel assembly is distributed within the upper part of a grid span, subchannel blockage will not become coplanar across the assembly. Therefore, the assembly retains its pin-coolant-channel arrangement and is capable of passing coolant along the pin to provide cooling for all l regions of the assembly. The effects of fuel rod bowing on whole-core blockage are considered in the fuel assembly and fuel rod designs, which minimize the potential for rod bowing. The minor adjustments of fuel pin pitch due to rod bowing do not alter the fuel assembly flow area substantially, and the average subchannel flow area is preserved. Therefore, due to the axial distribution of blockage caused by rupture, no coplanar blockage of the fuel assembly will occur, and the core will remain amenable to cooling. Deformation of the fuel pin lattice at the core periphery is allowed to occur from the combined mechanical loading of the LOCA and a seismic event. Using leak-before-break (LBB) methodology, the spacer grid impact loads are within the spacer grid elastic load limit and no permanent grid deformation is predicted (Reference 14). Therefore, the coolable geometry requirements are met for all fuel assemblies within the core. The consequences of both thermal and mechanical deformation of the fuel assemblies in the core have been assessed, and the resultant deformations have been shown to maintain coolable core configurations. Therefore, the coolable geometry requirements of 10CFR50.46 have been met and the core has been shown to remain amenable to core cooling. 6.5 Long-Term Cooling The fifth acceptance criterion of 10CFR50.46 states that the calculated core temperature shall be maintained at an acceptably low value, and decay heat shall be removed for the extended period of time required by the long-lived radioactivity remaining in the core. Successful initial operation of the ECCS is shown by demonstrating that the core is quenched, and the cladding temperature is returned to near saturation temperature. Thereafter, long-term cooling is achieved by the pumped injection systems. These systems are redundant and'are able to provide a continuous flow of cooling water to the core fuel assemblies so long as the coolant channels in the core remain open. Compliance with this criterion is demonstrated for the systems and components specific to the TMI-1 and ANO-1 plants. The initial phase of core cooling has been shown to result in low cladding and fuel temperatures. A pumped-injection system capable of recirculation is available and operated by plant personnel to provide extended coolant injection. For a cold leg break, the concentration of boric acid within the core might induce a crystalline precipitation, which could prevent the coolant flow from reaching certain portions of the core. The concentration of dissolved solids has been shown to 101
i Framatome Technolooies. Inc. 86-5002073-02 be' limited to acceptable levels through timely initiation of the dump-to-sump j recirculation mode or through passive recirculation through the RVWs and the hot leg nozzle gaps (Reference 15). Therefore, the capability of long-term cooling has been established and compliance to 10CFR50.46 has been demonstrated. 1 d 102
- Frarnatane Technoloales. Inct 86-5002073-02
- 7. RELAP5/ MOD 2-B&W EM SER Restrictions The NRC Safety Evaluation Report (SER) on BAW-10192-PA (Reference 1) contained eleven restrictions related to the use of the RELAP5/ MOD 2-B&W EM. Compliance with these eleven restrictions is demonstrated in Reference 30 and repeated below, j BWNT LOCA Evaluation Model: BAW-10192-PA i
- 1. The LOCA methodology should include any NRC restrictions placed on the individual codes usedin the evaluation model(EM).
Response: Sections 2.2 through 2.5 of Reference 30 detail the NRC restrictions placed on' the codes used in the BWNT LOCA EM. For LBLOCA analyses, the RELAP5/ MOD 2-B&W (includes BEACH), the REFLOD3B and CONTEMPT codes are utilized. For SBLOCA analyses, only the RELAPS/ MOD 2-B&W code is utilized.
- 2. The guidelines, code options, and prescribed input specified in Tables 9-1 and 9-2 in both Volume I and Volume II of BAW-10192-P should be used in LBLOCA and SBLOCA evaluation n'adel applications, respectively.
P.esponse: Table 9-1 in Volume I (LBLOCA) of BAW-10192-PA is verified via use of Table 4 in Reference 30. Compliance to the Table 4 restrictions for the LBLOCA analyses is I;sted in Appendix J of Reference 32. Table 9-2 in Volume ll'(SBLOCA) of BAW-10192-i'A is verified via use of Table 6 in Reference 30. Compliance to the Table 6 restrbion for the SBLOCA analyses is listed in Appendix R of Reference 6. These tables also include input and restrictions placed on the individual codes that make up the BWNT LOCA EM as discussed in detail in Reference 30.
- 3. The limiting linear heat ratu for LOCA limits is detennined by the power level and the product of the axial and radialpeaking factors. An appropriate axial peaking factor for use in determining LOCA limits is ono that is representative of the fuel and core design and that may occur over the core lifetime. .The radial peaking factor is then set to obtain the limFog linear heat rate. For this demonstration, calculations were performed with ths axial peak of 1.7. The general approach is acceptable for demonstrating the LOCA limits ~ methodology. However; as future fuel or vote designa evolve, the basic approaches that were used to establish these conclusions
'may change. FTl must ruvalidate the acceptability of the evaluation model peaking methods if: (1) significant changes are found in the core elevation at which the
. minimum core LOCA margin is predicted or (2) the core maneuvering analyses radial and axi,rl peaks that approach the LOCA LHR limits differ appreciably from those used to demonstrate Appendix K compliance.
103 x .. . - _ - _ _ = _ _ - _ - _ _ _ - __ b
Framotome Technoloaies. Inc. 86-5002073 02 Response: This restriction is related only to LBLOCAs. The axial and radial peak used in the Reference 32 analyses were similar and approximately 1.7 for all elevations and linear haat rates analyzed. The restriction states that FTl must revalidate the acceptability of the evaluation model peaking methods if: (1) significant changes are found in the core elevation at which the minimum core LOCA margin is predicted or (2) the core maneuvering analysee radial and axial peaks that approach the LOCA LHR limit differ appreciably from those used to demonstrate Appendix K cornpliance. Reference 34 defines several layers of screening criter needed to show compliance with the BWNT LOCA EM restriction on peaking. The methods provided are valid for any current or past Mark-B fuel type (including but not limited to Mark-B4Z, Mark-B8, Mark-B9, Mark-B10(F), Mark-B11) that is ruptured-node limited or has similar ruptured-or unruptured-node PCTs predicted with the BWNT LOCA EM. Four criteria were developed in Reference 34 from which to show compliance or to - define a LOCA linear heat rate (LHR) limit penalty associated with LHR limits calculated . based on the RELAP5 EM (Reference 4). These criteria are summarized below.
- 1. The fuel burnup must be compared to the LOCA LHR limits versus burnup. If the burnup is on the PCT-limited portion of the LOCA limit curve (< 40,000 mwd /mtu), then proceed to Step 2. If the burnup range is on the pin-pressure-limited portion of the curve, the restriction is met without any other conditions.
That is, no axial peaking checks or linear heat rate limit adjustments are needed for pin pressure limited LHRs.
- 2. If the burnup is on the PCT-limited portion of the curve, then the power distribution analysis LOCA margins must be checked at all core elevations. If there is less than 5% LOCA margin, proceed to Step 3. If there is more than 5%
margin, the restriction is met and no further checks are needed because the PCT at the maximum power distribution LHR will be lower than the BWNT LOCA EM PCT.
- 3. If the burnup is on the PCT-limited portion of the curve and there is less than 5%
LOCA margin, then varirtions in the augmented peaking factor versus the 1.7 axial used in the LOCA analyses must be considered. The axial peak must be 1.65 or greater for 0 to 4 ft power peak elevations,1.710.5 for 4 to 8 ft elevations, and 1.75 or less for 8 to 12 ft elevations. If these axial peaks are in compliance, the restriction is met and no further checks are needed. If they are not met, then proceed to Step 4 for the LOCA LHR limit reductions.
- 4. If the bumup is on the PCT-limited portion of the curve, there is less than 5%
LOCA margin, and the axial peak is not in compliance, then the power distribution analysis must assign a LOCA LHR limit penalty to ensure that the BWNT LOCA EM PCT (based on the given LHR and APF of 1.7) is not underpredicted. The LHR limit penalty compensates for the known deviation between the augmented axial peak and the required peak. The LHR limit 104 i
- - _ - _ _ _ _ _ _ _ _ IY
E 1 , Framatome Technoloaies. Inc. 86-5002073-02 reductions, ALHR, are core elevation dependent: 1 ALHRoi.4 n = min { 0.0, [ (APFm., a.inw,on.n 9...gm.ni.a p., - 1.65)
- 1.5 kW/ft ] },
ALHR4 to e n = min { 0.0, [ (1.75 - APFpo ,a, ww.on.n.iy.i..gm.ni.e p..)
- 4.0 kW/ft ] }
+ min { 0.0, [(APFpo ,ai mwon.n.ry...gmeni.a p.,- 1.65)
- 1.5 kW ,
and ALHRe to in n = min { 0.0, [ (1.75 - APFpo , a.inw,on .n.iy.. .gm.ni., p..)
- 4.0 kW/ft ] }
i
- 4. The mechanistic ECCS bypass modelis acceptable for cold leg transition (0.75 ft to 2.0 ft') and hot leg break calculations. The nonmechanistic ECCS bypass model must be used in the large cold leg break (2 2.0 ft) methodology since the demonstration calculations and sensitivities were run with this model.
Response: As outlined in BAW-10192-PA Volumes I and 11, different bypass models are used for large break and small break analyses. The nonmechanistic ECCS bypass model is used in large break analyses (2 2.0 ft2). The mechanistic ECCS bypass model is used for cold leg transition (0.75 ft" to 2.0 ft2), hot leg, and all smaller sized cold leg breaks. I
- 5. Time-in-life LOCA limits must be determined with, or shown to be bounded by, a specific application of the NRC-approved evaluation model. }
Response: Time-in-life cases were explicitly examined for the LBLOCA analyses. j Conditions appropriate to the specific time in life were used in the hot channel, while the BOL parameters were maintained in the average channel. The results demonstrated that the BOL LHR limits can be maintained up to 40 GWd/mtU. For 40 GWd/mtU to 60 GWd/mtU, the LHR limit is reduced linearly to 11.6 kW/ft in accordance with the TACO 3 pressure limit of 800 psi above system pressure. ; Time-in-life calculations for SBLOR applications are not required unless the fuel pin heatup is sufficient to cause cladds,9 supture. FTl evaluates the likelihood of rupture by analyzing the SBLOCA with a composite set of pin conditions that provide a conservative PCT prediction. End-of-life pin pressures are used to maximize the cladding hoop stresses, thereby improving the likelihood of rupture for those cases that do experience heatup. To maximize the cladding temperatures, the beginning-of-life (BOL) fuel stored energy and BOL oxide thicknesses are used. However, any case that predicts clad rupture with these conditions would be fuither parameterized by adjusting the time of rupture (via pin pressure or normalized heating ramp rate changes) to push rupture to the time of peak clad temperature. This composite method ensures that the calculated PCT will bound any PCT predicted by a 105
)
Framatome Technoloaies. Inc. 86-5002073-02 consistent time in life (TIL) analysis with appropriate TIL pin parameters. A pure TIL calculation (with fuel stored energy, pin pressure, and cladding oxide thickness consistent with the TIL that produces the worst rupture time) would be performed if the composite case is judged to be overly conservative. The consistent case would also use the plastic-weighted normalized heating ramp rate to predict the fuel pin swell and rupture performance. SBLOCA sensitivity studies performed and documented in Reference 38 indicated that for PCTs less than 1606 F, the most limiting PCT results are produced when rupture is not predicted because rupture tends to cool the node. Reference 39 studies showed that with PCTs near 1800 F, the ruptured node becomes limiting because of the metal-water energy contribution. The SBLOCA analyses from Reference 6 used a constant normalized heating ramp rate limit of one to minimize the likelihood of cladding rupture. Since the limiting SBLOCA PCT was below 1606 F, limiting PCT results are assured. The possibility of cladding rupture can not be ruled out, however the PCT predicted from a ruptured node condition would not be limiting for the 20% tube plugging analyses.
- 6. LOCA limits for three pump operation must be established for each class of plants by application of the methodology described in this report. An acceptable approach is to demonstrate that three pump operation is bounded by fourpump LHR limits.
Response: LBLOCA analysis of a three-pump case at a core power in the range of 65 to 85 percent is performed (or reference to a three-pump analysis for a similar plant design is made) to demonstrate that three-pump operation is bounded by four-pump LHR limits. The hot channel three-pump peak LHR limit is set equivalent to the 100 percent power 4-pump LHR limit. Because this analysis is performed at a power level less than 95 percent, it must consider the possibility of positive moderator temperature coefficient (MTC). An 90 percent full power three-pump analysis with a +1.0 pcm/F MTC was performed for Oconee in Reference 28 to show that the altered configuration j would not predict more limiting results than the 4-pump 100 percent full power case. i These results concluded that the 4-pump LHR limits are appropriate for 3-pump LHR l limits at 80 percent power and an MTC of +1.0 pcm/F or less for the 177-FA LL plants. Three-pump SBLOCA analyses are not performed, because the core power is reduced but the ECCS capacity remains at the 100 percent full power levels. Therefcre, four-pump fuli-power SBLOCA PCTs will bound the PCTs for similar three-pump partial power cases.
- 7. The limiting ECCS configuration, including minimum versus maximum ECCS, must be determined for each plant er class ofplants using this methodology.
Response: This restriction is primarily related to LBLOCAs and is not applicable to the < SBLOCA analyses. The limiting SBLOCA ECCS configuration is a single ECCS train for CLPD breaks. The minimurn containment pressure derived from a maximum ECCS 106 Y
Framatome Technolooies. Inc. 86-5002073-02 l flow was used in the LBLOCA analyses that considered minimum ECCS injection. This composite approach conservatively considers the worst containment pressure with the g minimum ECCS refill capacity to ensure the LBLOCA calculated consequences are l bounding for any combination of available ECCS pumps, l
- 8. For the small break model, the hot channel radial peaking factor to be used should
! correspond to that of the hottest rod in the core, and not to the radial peaking factor of the 12 hottest bundles. Response: There are twelve assemblies modeled in the hot bundle, and each pin is ) peaked to the hot pin radial value.
~
- 9. The constant discharge coefficient model(discharge coefficient = 1.0) referred to as the "High or Low Break Voiding Normalized Value," should be used for all small break analyses. The model which changes the discharge coefficient as a function of void fraction, i.e., the " Intermediate Break Voiding Normalized Value," should not be (
used unless the transiont is analyzed with both discharge models and the ) intermediate void methodproduces the more conservative result. Response: This restriction is related only to SBLOCA analyses. A constant discharge coefficient is used for SBLOCA analyses. Verification of this input is performed for each SBLOCA analysis, 10.For a specific application of the FTI small break LOCA methodology, the break size which yields the local maximum PCT must be identified. In light of the different possible behaviors of the local maximum, FTI shouldjustify its choice of break sizes in each application to assure that either there is no local maximum or the size yielding the maximum local PCT has been found. Break sizes down to 0.01 ft should be considered. Response: This restriction is related only to SBLOCA analyses. The SBLOCA break spectrum (down to at least 0.01 ft*) is performed to determine the local maximum PCT. ! The break sizes analyzed are chosen to ensure 'that the local peak has been ! appropriately defined, The full spectrum of break sizes performed for the BWOG 20% SGTP task covers this requirement.
- 11. B&W-designed plants have intemal reactor vessel vent valves (RVVVs) that provide l a path for core steam venting directly to the cold legs. The BWNT LOCA evaluation model credits the RVVV steam flow with the loop stearn venting for LBLOCA l analyses. The possibility exists for a cold leg pump suction' to clear during l blowdown and then reform during reflood before the evaluation model analyses l predict average core quench. Since the REFLOD3B code cannot predict this j i
107 J
_ w Framatome Technoloaies. Inc. 86-5002073-02 reformation of the loop seal, FTIis required to run the RELAP5/ MOD 2-B&W system model until the whole core quench, to confirm that the loop seal does not reform. This demonstration should be performed at least once for each plant type (raised loop and lowered loop) and bejudged applicable for all LBLOCA break sizes. . Response: This restriction is related only to LBLOCA analyses. This verification analysis was performed using the RELAP5 system model for the 177-FA LL plant design in Reference 16. The results of that analysis confirmed that a loop seal does not reform prior to whole core quench. Since these resu!ts were obtained using the 177-FA LL model, it can be concluded that this restriction of the evaluation modelis met. l l I I j f 108 J
Framatome Te'chnolooies. Inc. 86-5002073-02
- 8. . References
- 1. FTl Topical ' Report, "B&Ws ECCS Evaluation Model", BAW-10104-PA, Rev. 5, l
' November 1986.
- 2. FTl Topical hport, "B&Ws Small-Break LOCA ECCS Evaluation Model", BAW-10154- l P, Rev. O, November 1982.
- 3. FTl Document 32-126624-00, "20% TP LOCA Base Model", FTl Proprietary.
- 4. ' FTl Topical Report, "BWNT LOCA Evaluation Model for OTSG Plants", BAW-10192-PA, l Rev. 0,- June,1998.
- 5. FTl Topical Report, "TMI-12772 MWt ECCS Analysis with R5/M2-B&W", BAW-10222-P, l Rev. O, March 1997.
6; FTl Document 32-1266343-00, "BWOG SBLOCA 20% TP" FTl Praprietary.
- 7. FTl Document 51-1266254-01, "177 LL 20% Tube Plugging LOCA AIS".
- 8. FTl Document 32-1244481-00, "TMI-1 Uprate LBLOCA R5/M2", FTl Proprietary.
- 9. FCF Document 86-1267229-00, " Pre and Post LOCA EDFs", FTl Proprietary.
- 10. FCF Topical Report, " Extended Bumup Evaluation", BAW-10186-P. l
- 11. FTl Document 32-1239313-00, "LOCA Fuel Temperature Changes", FTl Proprietary.
- 12. FTl Document 32-1232704-02, "Oconee Mk-B11 LOCA Limits", Fil Preprietary.
- 13. FTl Document 32-1266332-00, "BWOG 20% TP LBLCCA", FTl Proprietary.
- 14. FTl Topical Report, "Framatome Marx-B Fuel Assembly Spacer Grid Deformation in B&W Designed 177-FA Plants", BAW-2292-P, Rev. O, February 1997. l
- 15. FTl Document 51-1266113-00,." Post-LOCA Boron Concentration Management".
- 16. FTl Document 86-5000368 00, " Loop Seal Clearing for 177-Lt.".
- 17. FTl Document 32-1266332-01, "BWOG 20% TP LBLOCA", FTl Proprietary.
- 18. Schyltz, R. G., Sandervag, O., and Hanson, R. G., "Marv? cn Power Station Critical Flow
. Data: A summary of Results and Code Assessment Applications", Nuclear Safety, Vol.
' 25, No.6, November-December 1984.
- 19. FTl Topical Report BAW-10171-P, Rev. 3, "REFLOD3B - Model for Multinode Core Reflooding Analysis", September 1989.
109
~
n i Framatome Technotoales. Inc 86-5002073-02
- 20. FTl Document 32-1232655-00, "Oconee Pump & ECC/ Cont Pres Studies", FTl Proprietary.
- 21. FTl Topical Report BA'N-100L-A, Rev.1, " CONTEMPT - Computer Code for Predicting Containment Pressure-Temperature Response to LOCA", April 1978.
- 22. FTl Document 32-1232665-00, "Mk-B11 LBLOCA Spectrum Study", FTl Proprietary.
- 23. FTl Document 32-1232704-01, "Oconee Mk-B11 LOCA Limits", FTl Proprietary.
- 24. FTl Document 32-1232704-02, *Oconee Mk-B11 LOCA Limits", FTl Proprietary.
i
- 25. FTl Topical Report BAW-10164-PA, Rev. 3, *RELAP5/ MOD 2-B&W - An Advanced Computer Program for Light Water Reactor LOCA and Non-LOCA Transient Analysis",
July 1996,
- 26. FTl Topical Report BAW-10166-PA, Rev. 4, " BEACH - A Computer Program for Reflood Heat Trarsfer During LOCA", July 1996.
- 27. FTl Topical Report BAW-10162-PA, " TACO 3 - Fuel Pin Thermal Analysis Computer Code", October 1989.
- 28. FTl Document 32-1234828-00, "Oconee Partial Power Study", FTl Proprietary.
- 29. FTl Document 32-0401-02,
- Minimum Back Pressure Study for a Typical 2500 MWt Class Reactor", FTl Proprietary.
I 30. FTl Document 51-5001731-00, "BWNT LOCA EM Limitations and Restrictions".
- 31. FTl Document 51-1203953-01, "RELAPS/ MOD 2 B&W Plant Noding." )
- 32. FTl Document 32-1266332-02, *BWOG 20% TP LBLOCA", FTl Proprietary.
l
- 33. FTl Document 32-1232670-04, "Oconee Mk-B11 SBLOCA Spectrum", FTl Proprietary.
- 34. FTl Document 51-5004541-00, " Radial vs Axial Core Peaking".
I
- 35. FTl Technical Document 74-1152414-08, " Emergency Operating Procedures Technical Bases Document, Volume 1, Generic Emergency Operating Guideline".
- 36. FTl Document 32-1244481-02, "TMI-1 Uprate LBLOCA RS/M2", FTl Proprietay.
- 37. FTl Document 32-1266332-03, *BWOG 20% TP LBLOCA", FTl Proprietary.
- 38. FTl Document 32-1234842-01, "Oconee Mk-B1175% FP SBLOCA", FTl Proprietary.
- 39. FTl Document 32-1234842-02, "Oconee Mk-B1175% FP SBLOCA", FTl Proprietary.
110
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