ML20217N213
| ML20217N213 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 04/24/1998 |
| From: | Handrick M, Johnson W, Rich J SARGENT & LUNDY, INC. |
| To: | |
| Shared Package | |
| ML20217N207 | List: |
| References | |
| SL-5203, NUDOCS 9805050256 | |
| Download: ML20217N213 (14) | |
Text
2 wnes,. sus 2e m.
l l Big Rock Point Nuclear Plant l Zircaloy Oxidation Analysis Prepared for Consumers Energy Prepared by !A 4//y/9 9 ;
John M. Rich Date' Senior Project incer, Shielding and Radiological Safety l
l Prepared by Y/;;?y[j7p Mark C. Handrick Date Engineer, Power Systems Engineering Division
/ A Revicwed by /g/ lf Lz Y/cpf William f/ofds'o'n' A Consultant, Shiciding and Radiological Safety Date Reviewed by _
~
Zg/hg Robert M. Field Date ' Senior Project Engineer, Power Systems Engineering Division Approved by (tu @ ' N!fI Steve Raupp Date l Project Manager 9805050256 900429 PDR W ADOCK 05000155 PDR , strul-5203. doc /042498
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. t 11 SL-5203 i- Jor ges Lundy
Zircaloy Oxidation Analysis Contents Section bage Exec u t ive S u m m a ry ..... ..................... .............. ............. ..... ................. .... ........ ... ......... . ..... I 1 Introduction............................................................................................................................. 3 1.1 Purpose.................................................................................................................. 3 1.2 Background................................................................................................................... 3 2 Description of Fuel Pool and Pool Con ten ts ....................... .. .. ................... .......... ................ 5 2.1 FuelPool............................................................................................................................... 5 2.2 Spentfuel......................................................................................................................... 5 2.3 Other Materi al s in the Pool . ............................................ ............................. .......... ....... ..... .... 6 3 Decay Heat G en e ra tio n Rate.............................. ........................................ .............................. 7 l 4 Zircalo y Clad d ing Te m pera tu re .............. .. ........................ .............................................. 9 t i ! 5 Conclusions.......................................................................................................................... I1 l
- 6 Refe rences a n d Co m p u ter Codes .... ............ ............................................ ......................... ..... 12 1
l r i 1 Tables ! t j L 2-1 Spent Fuel Pool Fuel Inventory ....... ................... . ........................ ........ ............................ ............. 5 4-1 Peak CladLin g Tem perature .................... .................................................................. . .................... 10 Figures 3-1 Decay Heat G enerati on Rate .............. .... ... .................................... .... ...................................... .. . 8 4-1 Peak Cladd in g Tem perature ................ . .. ........................ ....- ............................... ........... ... ....... 10 t l l l slM.5203. doc /042498
l 5 03 See Lun@ $ 5 * { l Consumers Energy Zircaloy Oxidation Analysis ; EXECUTIVE
SUMMARY
Re final reactor shutdown at Big Rock Point took place on August 29,1997. He reactor vessel is completely defueled, and all fuel is now stored in the spent fuel pool. Under these circumstances, the scenario with the potential for the greatest fission product release involves fuel in the spent fuel pool. As stated in NUREG/CR-6451 (Reference 1]: l After the reactor vessel is defueled, the traditional accident sequences that dominate the operating l plant risk are no longer applicable. De remaining source of public risk is associated with the accidents that involve the spent fuel. Previous studies have indicated complete spent fuel pool
~ drainage is an accident of potential concern. Certain combinations of spent fuel storage configurations and decay times could cause freshly discharged fuel assemblies to self heat to a temperature where the self sustained oxidation of the zircaloy cladding may cause cladding failure. ,
1 ne critical temperature for self sustained oxidation of zircaloy cladding is 565 *C (1049 'F) (Reference 1).
' With the parameters used for this report, approximately 220 days of decay is sufficient to reduce the spent fuel decay heat generation rate to a level below that required to raise the zirealoy cladding to 565 *C (1049 *F)
[ Reference 5].- Two-hundred-twenty days from the final reactor shutdown is April 6,1998. Herefore, it is concluded that April 6,1998, is the approximate date after which self-sustained oxidation of zircaloy cladding no longer needs to be considered. He analysis assumed bounding conditions, which include the following:
.- no water in the pool, e a cover over the pool, e assembly with the highest decay heat generation rate, and e pin peaking factor of 1.2.
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2 sir 5203 W Lun@ u e I I I To ensure that the maximum assembly decay heat generation rate was used in the clad temperature analysis, six assemblies were selected for analysis based on maximum bumup and maximum burn rates. The upper bound of the resulting calculated heat generation rates was used as input for the clad temperature analysis. l To ensure that the maximum clad temperature was calculated, the temperature of the surroundmg air was calculated based on the pool being dry and covered, and the fuel rod with the highest local peaking factor was modeled with the surrounding air as the ~only applicable heat sink. ) l l l l l l l l
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SL 5203 Serge La.sncey"* t
'1 INTRODUCTION i
1.1 Purpose j %e purpose of this report is to summarize analyses that, using conservative data and assumptions, determine ! the date when self-sustained zircaloy oxidation due to decay heat generation is no longer credible. De critical cladding temperature for zircaloy oxidation is 565 *C (1049 'F) as stated in NUREG/CR-6451 [ Reference 1). A set of bounding conditions is used with the computer codes ORIGEN2 [ Reference 2] and HEATING 7.2f [ Reference 3] to calculate the cladding temperature as a function of time. He point in time at which the cladding temperature cannot attain 565 *C (1049 'F) under these conditions is the point in time when i-l zircaloy oxidation due to decay heat is no longer credible. t Additional data are also generated for times well after zircaloy oxidation is possible. Rese data provide information that may be useful in analyses to support spent fuel handling operations that could take place within the next five years. L i
1.2 Background
De date of the final reactor shutdown at Big Rock Point was August 29, 1997. De reactor vessel is completely defueled, and all fuel has been decaying in the spent fuel pool since that time. Under these circumstances the scenario with the potential for the greatest fission product release involves fuel in the spent fuel pool. As stated in NUREG/CR-6451 [ Reference 1]: After the reactor vessel is defueled, the traditional accident sequences that dominate the operating L plant risk are no longer applicable. The remaining source of public risk is associated with the
- accidents that involve the spent fue!
- Previous studies have indicated that complete spent fuel pool drainage is an accident of potential concern. Certain combinations of spent fuel storage
! configurations and decay times could cause freshly discharged fuel assemblies to self heat to a j temperature where the self sustained oxidation of the zircaloy fuel cladding may cause cladding failure. L slAsi-5203. doc /04249s
- r. .
SL-5203 Serge Lun@ "* Note that NUREG/CR-6451 [ Reference 1] also defines four spent fuel configurations. 'Ihis report calculates . the date at which Configuration 1 ends, and provides additional data that may be useful for analyses that support Configurations 2,3, and 4. l Confleuration 1 - Hot Fuel in the Spent Fuel Pool Configuration 1 encompasses the period commencing immediately after the offload of the core to a point in time when the decay heat of the hottest assembly is low enough such that no substantial zircaloy oxidation takes place (given the pool is drained) and the fuel cladding will remain intact (i.e., no gap releases). l !' Confieuration 2 - Cold Fuel in the Soent Fuel Pool I Configuration 2 begins at the point in time when the decay heat loads can no longer initiate zircaloy oxidation. l Confleuration 3 - All Fuel Stored in an Independent SDent Fuel Storare Installation t Confleuration 4 - All Fuel Removed from the Site ! i slM.5203. doc /042498
serge L.uncty u s sir 5203 2 DESCRIPTION OF FUEL POOL AND POOL CONTENTS l 2.1 Fuel Pool l l ne spent fuel pool is approximately 20 ft wide,26 ft long, and 29 ft deep. He pool is modeled as being dry with a loosely fitting cover over the top of the pool. The effect of covering the pool is to reduce heat removal by convection, and this effect is included in the analyses. He actual details of the cover are not included in this report because it is not necessary to model them in the analyses. i l 2.2 Spent Fuel A total of 442 assemblies have been offloaded, of which 441 currently reside in the pool. De average burnup and offload dates are shown below in Table 2-1. ! Table 2 Spent Fuel Pool Fuel Inventory Average Burnup Offload Date Assemblies (mwd /mtu) 06/02/74 18 13,244 01/31/76 22 17,141 07/23/77 20 19,955 07/02/79 26 21,869 10/31/80 22 24,747 02/05/82 22 25,403 05/13/83 22 25,275 05/31/84 16 20,375 09/06/85 20 25.808 01/02/87 24 27,611 04/08/88 12 26,949 06/09/89 30 25,016 09/21/90 20 26,569 11/29/91 22 26,735 06/26/93 20 26,91I str\st-5203. dodo 42498
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,,, SL-5203 Average Burnup Offload Date Assemblies (mwd /mtu) 10/01/94 20 26,029 01/05/96 22 27,445
)
08/29/97 84 17,606 Six assemblies from the final full core ofiload (FFCO) on 8/29/97 were analyzed to determine the maximum assembly decay heat generation rate. The selection criteria include the a=sembly with the maximum burnup, the assembly with the maximum burn rate in the final power run, and four additional assemblies with high bmiup/ burn rate histories. The six selected assemblies are:
. 1515 - His assembly has the highest total bumup in the FFCO,29,668 mwd /mtu.
. 1811 - This assembly has the highest last cycle bum rate in the FFCO,21.57 MW/mtu.
- 1613 - His assembly has a last cycle bum rate and bumup of 16.46 MW and 28,826 mwd /mtu.
. 1617 - This assembly has a last cycle burn rate and burnup of 17.85 MW and 27,309 mwd /mtu.
. 1717 - This assembly has a last cycle burn rate and bumup of 19.81 MW and 25,499 hBVd/mtu.
+ 1703 - This assembly has a last cycle burn rate and burnup of 20.51 MW and 22,939 mwd /mtu.
2.3 Other Materials in the Pool In addition to spent fuel, the pool contains control rods, fuel channels, a.sd some miscellaneous equipment. He decay heat generated by these components is insignificant compared t3 that from the sp nt fuel. Herefore these components were not modeled in the analyses. str\st-5203. doc /042498 z - i
serges sir 5203 Luncty"* l l 3 DECAY HEAT GENERATION RATE De details and results of the decay heat generation rate calculations are contained in Sargent & Lundy Calculation MECH-0150, " Spent Fuel Decay Heat," [ Reference 4]. He computer program ORIGEN2, version 2.1 [ Reference 2) is used to calculate decay heat based on fuel operating history, nree sets of operating histories are analyzed in the ORIGEN2 calculations:
. He specific assembly history is used for 6 fuel assemblies; 1515,1811,1613,1617,1717, and 1703.
. The average FFCO core history is used for the FFCO (84 assemblies).
. A conservative history is developed and used for the fuel from earlier offloads (357 assemblies in the partial core offloads prior to the FFCO).
He decay heat from the FFCO and partial core offloads is summed to obtain the total decay heat for the fuel pool, ne decay heat from each of the six individual assemblies is included in the FFCO decay heat calculation, so it is not added in a second time. The calculated decay heat generation rate, as a function of time is shown in Figure 3-1 below.4 He data that were used to generate the figure are contained in Table 7-1 Sargent & Lundy Calculation MECH-150 i i [ Reference 4] . As can be seen from Figure 3-1, the analyzed assemblies have approximately the same peak values over time. Because the enveloping assemblies were analyzed (maximum burn and maximum burn rate), it is concluded that these assemblies envelope the peak assembly decay heat generation rate. i
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Figure 3 Decay Heat Generation Rate , 1 (Time is days after final retetor shutdown, 8/29!97) 1.0E+06 mzmu . 1
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'%h_ _i DI51 L ww rripet 1.0E+05 l
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# N_ i i Utarewn. omoeirI 27 % l 1.0E+04 7analFu'lGo e i
load (FFCO7 ^q IA ---- - L ---~
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'W7 1.0E+02 _
l 1.0E+01 1.0E+00 1 10 100 1000 10000
"';w After Last Shutdown in Days i
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I g ,,, sir 5203 l . , _ l l 4 ZlRCALOY CLADDING TEMPERATURE The details and results of the decay heat generation rate calculations are contained in Sargent & Lundy l l Calculation 1998-00300, " Maximum Cladding Temperature For Spent ruel Rod" (Reference 5]. The computer l program HEATING 7.2f[ Reference 3] is used to calculate maximum clad temperature based on decay heat and fuel rod geometry. Clad temperature is calculated at 140,220,300, and 365 days after the final reactor shutdown, August 29,1997. The heat transfer model includes the effects of covering the pool with a protective cover that precludes air flow -i into the spent fuel pool cavity. Conduction heat transfer is assumed to occur at the cladding surface / air l 1 interface. This is conservative with respect to calculated cladding temperature since convection and radiation I heat transfer at the cladding surface / air interface are neglected. The maximum calculated cladding temperature, as a function of time, is shown in Figure 4.1 below (rounded ! up to the nearest degree). The data that were used to generate the figum are contained in Table 4-1. l As can be seen from Figure 4 1, the peak cladding temperature is below the critical oxidation temperature when spent fuel has cooled for 220 days or longer after the final core shutdown. ! ) I i l shW5203. doc /042498
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. l 10 Serge Lu% u, SL-5203 i
Figure 4 Peak Cladding Temperature (Time is days after final reactor shutdown,8/29/97) 1300 - T N 1250 1200 1150
!!00 1050 -
1000 f 950 \
\
{900
,3 850 S 800 750 700 650 600 4 100 150 200 250 300 350 400 Time After Final Reactor Shutdown (days)
Table 4 Peak Cladding Temperature (Time is days after final reactor shutdown,8/29/97) Maximum Calculated Days After FFCO Cladding Temperature 140 1300 #F 220 1046*T 300 901 'F 365 822 'F e i ow.52cuocm249: j J
I i L5203 w wncey"* I l 5 CONCLUSIONS t Analyses have been performed that calculate the peak fuel cladding temperature due to decay heat as a function of time after the final reactor shutdown [ References 4 and 5]. The analyses used a ut of bounding conditions that include the following: e no water in the spent fuel pool, e a cover over the spent fuel pool, l e assembly with the highest decay heat genen don rate, and e pin peaking factor of1.2. i Using these bounding conditions, the analyses show that the peak cladding temperature is below the critical oxidation temperature when the spent fuel has cooled for approximately 220 days or longer after the final l reactor shutdown. Therefore, self-sustained oxidetion of zircaloy cladding (as a result of a sudden loss of spent fuel pool cooling water) need not be considered credible when the decay period between the final reactor ( shutdown and the postulated accident is approximately 220 days or longer. Two-hundred-twer.sy days from the final reactor shutdown (8/29/97) is April 6,1998, whid is the approximate ! date after which self-sustained oxidation of zircaloy cladding no longer needs to be considered. i
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F serge I uncey"= L-5203 i j 6 REFERENCES AND COMPUTER CODES
. 1. "A Safety and Regulatory Assessment of Generic BWR and PWR Permanently Shutdown Nuclear Power Plants," NUREG/CR-6451, BNL-NUREG-52498.
- 2. ORIGEN2," Isotope Generation and Depletion Code," Sargent & Lundy Program No 03.7.533-2.12. i i
[ ORIGEN 2.1 is a point depletion code system for calculating the buildup, decay and processing of i radioactive materials in nuclear fuel and associated non-fuel materials. It was developed, and is
- maintained by Oak Ridge National Laboratory. It has been verified and validated under Sargent &
l Lundy's Quality Assurance program and maintained as Sargent & Lundy program number 03.7.533- 1 2.12.
- 3. HEATING 7.2f, Multidimensional, Finite-Difference Heat Conduction Analysis, Sargent & Lundy Program No. 03.7.564-7.2.
l HEATING 7.2f is a multidimensional, finite-diffe ence heat conduction code developed by Oak Ridge National Laboratory, verified and validated under Sargent & Lundy's Quality Assurance l program, and maintained as S&L program number 03.7.564-7.2.
- 4. " Spent Fuel Decay Heat," Sargent & Lundy Calculation MECH-0150, Revision 0, 04/16/98 (193 pages).
!. 5. " Maximum Cladding Temperature for Spent Fuel Rod," Sargent & Lundy Calculation 1998-00300, l Rev.1,04/23/98 (142 pages). l I su s203.wm2es}}