ML20040E019
| ML20040E019 | |
| Person / Time | |
|---|---|
| Site: | Waterford  |
| Issue date: | 01/31/1982 |
| From: | ABB COMBUSTION ENGINEERING NUCLEAR FUEL (FORMERLY |
| To: | |
| Shared Package | |
| ML19297F303 | List: |
| References | |
| C-CE-7452, NUDOCS 8202020493 | |
| Download: ML20040E019 (18) | |
Text
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Encl. (1) to C-CE 7452 l: METHOD of COOLABILITY ANALYSIS for
. DEFORMED GRIDS in PERIPHERAL ASSEMBLIES l
WATERFORD STEAM l ELECTRIC STATION UNIT NO. 3
.. JANUARY,1982 l f22go O 2 SS MS A ccaaeustcN ENGNEIPING .NC u
LEGAL NOTICE THIS REPORT WAS PREPARED AS AN ACCOUNT OF WORK SPONSORED BY COMBUSTION ENGINEERING, INC. NEITHER COMBUSTION ENGINEERING NOR ANY PERSON ACTING ON ITS BEHALF:
A. MAKES ANY WARRANTY OR REPRESENTATION, EXPRESS OR IMPLIED INCLUDING THE WARRANTIES OF FITNESS FOR A PARTICULAR PURPOSE OR MERCHANTABILITY, WITH RESPECT TO THE ACCURACY, COMPLETENESS, OR USEFULNESS OF THE INFORMATION CONTAINED IN THIS REPORT, OR THAT THE USE OF ANY INFORMATION, APPARATUS, METHOD, OR PROCESS DISCLOSED IN THIS REPORT MAY NOT INFHINGE PRIVATELY OWNFD RIGHTS;OR B. ASSUMES ANY LIABILITIES WITH RESPECT TO THE USE OF, OR FOR DAMAGES RESULTING FROM THE USE OF, ANY INFORMATION, APPARATUS, METHOD OR PROCESS DISCLOSED IN THIS REPORT.
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ENCLOSURE - NON-PROPRIETARY ,
Wat rford-3: M thod of Non-Proprietary Enclosure Coolability Analys.is.for Deformed to C-CE-7452 Gtids in Perpoheral Assemblies
-1.0 Introduction C-E has initiated an emergency core cooling system (ECCS) performance analysis to demonstrate the coolability of peripheral fuel assemblies. The analysis
- . will assume that these assemblies experience the maximum hypothetical reduction in coolant channel flow area due. to a combination of loss-of-coolant accident (LOCA) and seismic lateral loading. At present, C-E is performing an asymmetric loads analysis of Waterford Unit 3 which will determine if-deformation can result. This coolability analysis is being performed as a contingency in case the final asymmetric load analysis predicts permanent deformation of the peripheral assemblies. The purpose of the coolability analysis is to show that the allowable peak linear heat generation rate (PLHGR) in the deformed assemblies is greater than the actual maximum PLHGR that can occur in those assemblies consistent with the maximum allowable 3-D PLHGR
- which is achieved in the interior of the core. In thi: way it will be shown that the hot rod in an interior undeformed assembly remains the limiting case for acceptable ECCS performance. The maximum allowable 3-D PLHGR,13.7 kw/ft, is reported in the Waterford-3 FinalSafety Analysis Report (FSAR)(1) ,
i The purpose of this report is to provide a description of the method which is being used in this analysis. The method utilizes C-E's Large break ECCS
- - evaluation model(2) with specific input changes from the FSAR analyses to represent the assumed deformed configuration of the peripheral assemblies. The srccific changes are described in Section 3.0.
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This analysis will utilize an existing blowdown hydraulic results which are applicable to all C-E 34xx plants. The NSSS design and operating parameters as well as the maximum hypothetical grid deformation values for the 34xx reactors are the same. The 34xx blowdown analysis was performed for a 1.0 DE break in the reactor coolant pump (RCP) discharge leg. For Waterford-3, the limiting break is a 0.8 DE break in the same location. Section 2.0 compares the results
. of 1.0 DE and 0.8 DE breaks for Waterford-3 and provides a justification for performing the coolability analysis for the 1.0 DE break.
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2.0 Comparison of 1.0 DE and 0.8 DE Break Results Figures 1 to 3 compare the blowdown core flow, peak clad temperature, and peak clad oxidation transient; at the hot spot for the 1.0 DE and 0.8 DE breaks in
- the RCP discharge leg. Table 1 summarizes the significant performance results for these two breaks. As shown in the figures, the blowdown and refic3d
'. transients are almost identical. The clad heat-up transients proceed in the same manner with clad rupture predicted at the same location and at about the same time during the early reflood period. This rupture location has been determined to be limiting for all large breaks for Waterford-3. The peak clad temperature (PCT) is also predicted to occur at the same location for both breaks. The difference in the PCT calculated for these two breaks is only 3 F and the PCT occurs within one second of each other. In both cases the PCT occurs during late reflood after the reflood rate diminishes to below one I
inch per second. The small variations in blowdown hydraulics have little influence on PCT in late reflood since the controlling parameters, decay heat and reflood heat transfer coefficients, are independent of break area.
l Because of the similarity of the transients, the effect of a reduction in flow area in the peripheral assemblies would have the same influence on both breaks. Therefore, the conclusiors of the coolability analysis performed for the 1.0 DE break would be applicable to the 0.8 DE break.
3.0 Method of Analysis The analysis will utilize the same model(2) as used in the FSAR analysis except for input modifications necessary to represent the geometry of the deformed coolant channel, and the local power distribution characteristics of
. the peripheral assemblies. It will be asssumed that the peripheral assemblies experience the maximum hypothetical flow area reduction along the entire length of the core. The average area reduction of the assembly and the additional area reduction of the hot-channel will be explicitly modeled. It will be assumed that the assembly, containing the highest powered rod of any peripheral assembly, deforms to the maximum assembly average value and the hot sub-channel deforms to the maximum sub-channel value. The maximum area reduction values were conservatively derived and correspond to occurrance of solid-solid load path through adjacent grid strip and fuel rods.
The following provides the method of analysis to demonstrate the coolability of such deformed peripheral assemblies. The analysis will be performed in five stos which are described in Sections 3.1 through 3.5.
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t 3.1 Blowdown Analysis The CEFLASH-4A(3) computer code was utilized in determining the blowdown behavior. In CEFLASH 4A the hot assemaly is modelled separately as required by Appendix K to 10CFR50. In addition, hydraulic communication with its nearent neighbors is modelled explicitly. For this analysis, it was assumed that the hot assembly deforms to the maximum assembly average deformation of the peripheral assembly along the entire length of the core.
The hot assembly flow area was reduced by the maximum assembly average area reduction value. The initial flow in the assembly was calculated by equating the pressure drop through the deformed assembly to the same value predicted for the rest of the core. Also, the increased flow resistance for the smaller flow area and hydraulic diameter were represented to define the initial flow rate in the assembly. The flow area and initial flow rate of the nearest neighbors were increased by the amount these quantities were reduced in the hot assembly. This resulted in a maximum flow diversion away from the hot assembly.
The hot assembly was initialized at the same power density as assumed in the licensing analysis of the limiting interior assembly. The blowdown information, i.e. flow rate and enthalpies, generated for the deformed hot assembly will be used in the -STRIKIN-II (See Section 3.5) representation of the peripheral assembly. This is conservative since the hot assembly yielded higher transient fluid enthalpy quality, hence poorer heat transfer effectiveness, than would be calculated at.the peripheral assembly if it were modelled explicitly. Application of this poorer heat transfer effectiveness to
the peripheral assembly in STRIKIN-II will result in a higher fuel stored energy at the end of blowdown.
3.2 Determination of_ Reflood Rates COMPERC-II(4) is utilized in calculating the refill /reflood behvior (i.e.
reflood rate, two-phase levels etc.). However, these are core average parameters and are controlled by the containment pressure. The containment pressure is determined by the integral leak energy, and remains unaffected by the deformation of the peripheral assemblies. Therefore, the eflood rates and two-phase levels calculated for the FSAR analysis will be utilized as described in Sections 3.3 and 3.4 to calculate the reflood heat transfer coefficients.
3.3 Calculation _ of FLECHT Reflood Heat Transfer Coefficients The third step is to run COMPERC-II with HTC0F option (4) to calculate the FLECHT heat transfer coefficients using the reflood rates from Section 3.2.
The reduction in the flow area and the hydraulic diameter due to deformation of the coolant channel degrades the FLECHT heat transfer coefficients.
C-E applies a FLECHT based reflood heat transfer correlation (designated MOD-IC) for use on the C-E 14 x 14 fuel assembly. Reference 5 provides a ,
procedure, approved by the NRC in Reference 6, for applying the M00-1C correlation to other fuel geometries. The procedure applies a non-constant correction factor to the M00-1C correlation to account for geometric differences betwee' the 14 x 14 and other fuel geometries assuming the axial
power distributions are the same. The procedure accounts for independent variations in the-fuel rod diameter, hydraulic diameter, and coolant channel flow area. This procedure, as discussed below in detail, will be utilized to account for reduction in the channel flow area and hydraulic diameter due to the assumed grid deformation. _
Two sets of FLECHT heat transfer coefficients will be calculated using the maximum sub-channel and the maximum assembly average deformations in the peripheral assembly. Thase two sets will be used in STRIKIN at the hot-rod and the average rod of the peripheral assembly, respectively (STRIKIN models
-* the hot rod and average rod of an assembly).
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i 3.4 Calculation g Steam Cooling Heat Transfer Coefficients The steam cooling heat transfer coefficients will be used at the hot rod at and above the rupture plane for the period during which the reflood rate is less
'. than 1.0 inch /second. As a simplification, the analysis would conservatively
. utilize a minimum steam cooling heat transfer coefficient. (As described in
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CENPD-132 Supplement 1(2)), this value is based on radiation to steam only.
However, if this simple approach is determined to be unnecessarily conservative, the coefficients calculated by the PARCH (7) computer code will be utilized. PARCH calculates convective steam cooling heat transfer coefficients.
The PARCH code will be initialized with the core conditions calculated by COMPERC-II at the time when the reflood rate diminishes to below one inch /second. The only parameter, other than the chanal geometry, which is affected by the deformation of the assembly is the steam flow rate in the channel, which is input to PARCH. This steam flow rate in the deformed sub-channel will be hand calculated using the core parameters t'.wo-phase levels etc) from COMPERC-II (Section 3.2) and taking into account the reduction in steam flow due to the initial deformation of the sub-channel as well .as the additional local blockage due to clad rupture. It will be conservatively assumed that the hot-channel is deformed by the maximum sub-channel deformation
.- value, and is in parallel with an undeformed channel. The steam flow rate in the blocked region of the hot channel will be calculated by equating the axial pressure drop in the blocked channel with that in the unblocked adjacent subchannel.
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h 3.5 Temperature Calculation STRIKIN-II(O) will be utilized to determine clad temperature and clad oxidation transients. The blowdown and refill /reflood informatic.' described in Section 3.1 and 3.2 along with the FLECHT and the steam cooling heat transfer coefficients described in Sections 3.3 and 3.4 will be input to STRIKIN. The STRIKIN analysis will be performed for the maximum power pin in the peripheral assemblies. It will be assumed that the hot subchannel (i.e.
channel around the maximum power pin) experiences the maximum hypothetical sub- g channel deformation. The analysis will use the radial pin power distribution appropriate for a peripheral assembly. The peripheral assembly has a much moie radially skewed pin-to-pin power distribution than interior assemblies. This results in greater redistribution of hot pin decay heat to neighboring pins and better heat transfer due to rod-to-rod thermal radiation.
STRIKIN-II will be run at a peak linear heat generation rate (PLHGR) which is higher than the actual maximum PLHGR that can occur in the peripheral assemblies for the maximum allowable 3-D PLHGR (13.7 kw/ft) reported in the FSAR. The peak clad temperature (PCT) and peak local clad oxidation (PLO) for that assumed PLHGR in the peripheral assemblies will be shown to be lower than those for the limiting case described in the FSAR. Thus it will be demonstrated that the limiting case for compliance with 10CFR50.46 remains the higher powered rod located in an interior assembly in the core.
References
- 1. Waterford Steam Electric Station Unit No. 3, Final Safety Analysis Report.
- 2. " Calculative Methods for the C-E Large Break LOCA Eva;uation Model", CENPD- ,
132, August 1974, (Proprietary).
" Updated Calculative Methods for the C-E Large Break LOCA Evaluation Model", CENPD-132, Supplement 1, December,1974 (Proprietary).
" Calculational Methods for the C-E Large Break LOCA Evaluation Model",
CENPD-132, Supplement 2, July 1975, (Proprietary).
- 3. "CEFLASH-4A, A FORTRAN-IV Digital Computer Program for Reactor Blowdown Analysis", CENPD-133, April,1974, (Proprietary).
"CEFLASH-4A, a FORTRAN-IV Digital Computer Program for Reactor Blowdown Analysis (Modification)", CENPD-133, Supplement 2, December 1974, (Proprietary ).
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- 4. "COMPERC-II, A Program for Emergency Refill-Reflood of the Core", CENPD-134, April 1974, (Proprietary).
"COMPERC-II, A- Program for Emergency Refill-Reflood of the Core", Modifi-cation), CENPD-134, Supplement 1, December 1974, (Proprietary).
- 5. "Reflood Heat Transfer: Application of FLECHT Reflood Heat Transfer Coefficients to C-E's 16 x 16 Fuel Bundles", CENPD-213, January 1976, (Proprietary).
- 6. Letter from Karl Kniel (NRC) to A. E. Scherer (C-E), August 2,1976.
- 7. " PARCH, A FORTRAN-IV Digital Program to Evaluate Pool Boiling, Axial Rod and Coolant Heatup", CENPD-138, August 1974, (Proprietary).
" PARCH, A FORTRAN-IV Digital Program to Evaluate Pool Boiling, Axial Rod and Coolant Heatup", CENPD-138, Supplement 1, February,1975, (Proprieta ry) .
" PARCH, A FORTRAN-IV Digital Program to Evaluate Pool Boiling, Axial Rod and Coolant Heatup", CENPD-138, Supplement 2, January,1977, ..
(Proprietary).
.- 8. "STRIKIN-II, A Cylindrical Geometry Fuel Rod Heat Transfer Program",
CENPD-135, April 1974, (Proprietary).
"STRIKIN-II, A Cylindrical Geometry Fuel Rod Heat Transfer Program
- (Modification)", CENPD-135, Supplement- 2, December 1974, (Proprietary).
"STRIKIN-II, A Cylindrical Geometry Fuel Rod Heat Transfer Program",
' CENPD-135, ~ Supplement 4, August 1976, (Proprietary).
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TABLE 1 Parameter 1.0 DE Break 0.8 DE Break End of Bypass 19.7 Seconds 20.1 Seconds Clad Rupture Time 60.5 Seconds 59.0 Seconds
. Clad Rupture Location 65%* 65%*
Peak Clad Temperature 2115*F 2118*F Peak Clad Temperature Time 238 Seconds 239 Seconds Peak Clad Temperature Location . 65%* 65%*
Peak Clad Oxidation 16.6% 16.7%
Peak Clad 0xidation Location 65%* 65%*
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- Percentage of active core height measured from the bottom of the active core.
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