ML20085N779
| ML20085N779 | |
| Person / Time | |
|---|---|
| Site: | Catawba  |
| Issue date: | 06/19/1995 |
| From: | Tuckman M DUKE POWER CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9506300313 | |
| Download: ML20085N779 (19) | |
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,j h-j June 19,1995 '
- l Document Control Desk :
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. US Nuclear Regulatory Commission
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Subject:
- Catawba Nuclear Station, Units 1 &2 Docket Nos. 50-413 and 414, respectively' d
Response to NRC Req'uest for AdditionalInformation Fuel Enrichment Upgrade Submittal t
Gentlemen:
q Please find attached, Duke Power Coinpany's response your Request for Additional.
j Information (RAI), dated June 7,1995. The RAI concerns our fuel enrichment upgrade.
t submittal, dated September 19,1994. Responses are provided to all NRC. questions.
l We hope these responses are sufficient to satisfy your concerns. Ifyou have any questions,-
j or need additional information, please contact Ms. Judy Twiggs at 704-382-8897.
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. Sincerely, M.S. Tuckman,.
Senior Vice President Nuclear Generation Department jgt/ attachments
- U.S. N.R.C.
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R.E. Martin, Senior Project Manager -
Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission i
Mail Stop 14H25, OWFN i
. Washington, D.C. 20555 -
S.D. Ebeneter, Regional Administrator -
U.S. Nuclear Regulatory Commission - Region II.
101 marietta Street, NW - Suite 2900 l
Atlanta, Georgia 30323 j
- R.J. Freudenberger, Senior Resident Inspector Catawba Nuclear Station i
Max Batavia, Chief Bureau ofRadiological Health l
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South Carolina Dept.of Health and Environmental Control:
2600 Bull St.
Columbia, S.C. 2920I t
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Duke Power Company -
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Catawba Nuclear Station Response to Request for Additional Information -
Regarding Proposed Fuel Enrichment Increase -
Ql)
Section 9.1.3.1.1 of the UFSAR of October 1993 states that the " normal" heat load consists of 1/3 core with full irradiation and 7 days decay, one fuel (sic) core of empty spaces and the remainder of the spent fuel pool (SFP) filled with.
previously irradiated fuel. Further, the " maximum" heat load includes a full core
'l discharge consisting of 1/3 core irradiated 11 days and decayed 7 days,1/3 core j
irradiated one full cycle and decayed 7 days,1.3 (sic) core irradiated two full cycles and decayed 7 days; plus 1/3 core fully irradiated and decayed 25 days with -
l the remainder of the pool filled with fuel from previous yearly refuelings.-
a)
Does the normal case in the UFSAR' correspond to the " normal" case in the submittal of September 19,19947 l
b)
Does the maximum case in the UFSAR correspond to that in the submittal?
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c)
What, in the case of the submittal, does full irradiation correspond to in days ofoperation at power?-
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d)
What is the irradiation time and decay periods for the fuel "from previous j
yearly refuelings" at the times corresponding to the normal and maximum cases in the submittal?
I e)
If the answer is negative to."a" or "b" above, provide the correct answer and explain it.
j A1).
The maximum and normal decay heat load' cases in the submittal correspond to the Catawba Nuclear Station FSAR cases, with qualifications as explained below. The maximum decay heat load case' assumes a full core discharge (193 ' assemblies), 7 days decayed, with batch burnup values ranging from 19.1 GWD/t to 58.2 GWD/t.
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The remainder of the SFP is assumed to be filled with discharge batches of 76 assemblies, decayed for consecutively longer periods of time based on projected l
cycle lengths, irradiated to high burnup values (42.5 to 47.2 GWD/t). The value of.
.193 assemblies plus consecutive discharged batches of 76 assemblies fills 1409 fuel assembly cells with an assembly. In this analysis there are 12 empty cells based on 3
the method chosen to achieve the maximum postulated burnup (experience l
demonstrates that burnup is the limiting factor for decay heat evaluations). The
.l empty 12 cells in this analysis is deemed to have an insignificant affect on overall decay heat results, since these cells would be filled with fuel with a long decay time. Additionally, the decay heat is dominated by the recently discharged high i
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burnup fuel, in which the original FSAR definition of 11 days irradiation was seen to be non-conservative. The core assumed to be discharged in this analysis has been fully irradiated, with much of the fuel sustaining a burnup which would only be achieved over two cycles.
The computer code PANTHER was used to evaluate decay heat (see explanation of this code below). PANTHER has the capability to input an initial burnup and an ending burnup, in which the actinide fission power fractions are varied based on the period in operating life. Hence, PANTHER is capable of evaluating fuel with an outage time in between cycles, while maintaining track of actinide fission power fractions. However, this option was not selected when the analysis was performed. Outage time in between cycles was ignored, and irradiation was assumed to be consecutive (i.e., a batch-averaged approach was utilized in the evaluation of decay heat)._ For example, for fuel with a burnup of 44.4 GWD/t, PANTHER would have irradiated this fuel for a total number of days as follows:
EFPDs=(A'$m/)(AM) = (44,400)(88) = 1,145.5 A6Vt 3411 Hence, the method chosen to run the decay heat evaluations yielded conservative results for all the assemblies included in this analysis (see also the discussion below regarding the conservative nature of the fission product neutron capture correction factor).
The normal decay heat load case assumes a third of the core burned to a high
- burnup value (i.e., 58 GWD/t), with consecutive discharge batches of 76 assemblies burned to the same burnup values as with the maximum decay heat load case.
See attachment 2 for details of the fuel assumed to reside in the SFP. Evaluations performed with the 12 empty cells discussed above assumed to be filled with high burnup fuel yield no discernible differences from the decay hat or SFP equilibrium texperature reported in Reference 1. The SFP equilibrium temperatures reported in Attachment 1 includes the values associated with all of the cellsfilled with dischargedfuel.
Q2)
You note that, when the PANTHER code was used to calculate the decay heat, the values of 4.7E7 and 1.85E7 BTU /Hr were calculated for the maximum and normal cases, respectively. Also,4.33E7 BTU /Hr was calculated for the maximum case when using the ORIGEN code. Which value did you use for the maximum case? What are the bases for the values you used?
A2)
The values calculated with the PANTHER computer code were used in the evaluation ofdecay heat and spent fuel pool temperature. ORIGEN2 results are provided to demonstrate that the PANTHER computer code is producing conservative and accurate results. It is our intention to perform licensing calculations with the PANTHER computer code, since this methodology produces conservative results as explained below. However, decay heat cases are evaluated on an infrequent basis for conditions in which the Spent Fuel Pool Cooling System must be removed from service for maintenance. For these conditions, decay heat i
and heatup rate are calculated to ensure that the spent fuel pool design temperature is not exceeded prior to returning the Spent Fuel Pool Cooling System to service.
The design temperature of the spent fuel pool is 150 *F, since this is the design temperature for the concrete at Catawba Nuclear Station (based on ACI 318-71).
For these infrequent evaluations, we intend to utilize the ORIGEN2 computer code for reasons explained below.
ORIGEN is an Oak Ridge National Laboratory computer code which is in world-wide use for performing evaluations related to decay heat, isotopic inventory, etc.
There are some differences in how ORIGEN2 and ORIGEN-S handle cross section burnup dependence (see Reference 2). However, for purposes of decay heat and isotopic inventory studies, ORIGEN2 and ORIGEN-S results variation should not be significant enough to affect the calculations discussed herein.
ORIGEN-S has been noted to be an acceptable method to perform decay heat evaluations for fuel in an independent spent fuel ctorage installation. It is noted that ORIGEN and ANSI /ANS-5.1 produce similar results when the conservatism of the fission product neutron capture correction in ANSI /ANS-5.1 is discounted (Reference 3; NRC Staff acceptance of the ORIGEN matrix-exponential methodology is inferred).
The PANTHER computer code allows calculation of the correction factor for fission product transmutation due to neutron capture for short decay times.
However, the decay heat runs with the PANTHER computer code for the enrichment upgrade submittal were evaluated with the default correction factor (i.e., the values contained in ANSI /ANS-5.1). ANSI /ANS-5.1 uses a correction factor to account for fission product transmutation due to neutron capture taken from Reference 4, Figure 5, for the highest thermal neutron flux evaluated (a 2
thermal flux of 3E14 n/cm -sec). This value is not representative of Catawba core designs. Typical Catawba core designs range from a BOC value of SE13 to an EOC value of 6.5E13. Therefore, use of the default value for this correction factor yields conservative results. Reference 5 notes that for operating times and
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'i other conditions not exceeding the conditions epon which the values are based, the i
ANSI /ANS-5.1 Gm orrection factor overestimates the neutron capture. Also, c
the use of the ANSI /ANS-5.I methodology included 2a uncertainty. Therefore,-
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use of the.OKIGEN2 computer code should yield decay heat results which are -
more appropriate for Catawba Nuclear Station core design conditions.
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It is concluded that both PANTHER and ORIGEN are acceptable methods for performing decay heat evaluations. Duke Power Company intends to use l
PANTHER for licensing calculations and ORIGEN for operational calculations i
. (see also response to question #5).
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i Note: Questions 3 and 7 will be addressed together.
Q3)
What SFP coolant temperatures did you obtain for the normal and maximum conditions?
Q7)
Discuss the values of UAF (overall heat transfer coeflicient
- heat transfer surface area
- area correction factor) found for the SFP heat exchangers. Was each heat exchanger tested separately to determine the UAF7 Was any extrapolation required to apply the test data to conditions of use in normal and maximum conditions?
A3/7) Inputs, mathematical models and results may be found in Attachment 1 of this document (enerpts from Reference 1). Three test were conducted to determine Spent Fuel Pool Cooling (KF) System heat exchanger UAF, two of which are documented in Reference 1. The two tests were conducted with different Component Cooling (KC) System flow rates. The UAF for the KF System heat exchangers during high heat load conditions was calculated to be 1.3562E6 BTU /Hr-T. The UAF for the heat exchangers during reduced heat load conditions was calculated to be 1.1703E6 BTU /Hr-T. The higher UAF was calculated from test data obtained with a KC System flow rate of 3500 GPM, and the lower UAF value was calculated from test data obtained with a KC System flow rate of 2450 GPM. For both the normal and maximum heat load cases, a UAF value of IE6 BTU /Hr-T was assumed to apply. For the normal decay heat load case, it was assumed that the KC System flow rate was 3000 GPM to the heat exchanger in operation.
f For the maximum decay heat load case, it is noted that the preoperational flow balance of the KC System adjusted the KC System flow control valve (to the KF System heat exchanger) for only one KF train operating at a time, not two heat exchangers operating simultaneously and in parallel (each KF System heat exchanger has a KC System flow control valve operated by a manual loader). The preoperational flow balance also utilized only one KC System pump in operation, while two KC System pumps would be capable of delivering 3000 GPM of flow to each KF heat exchanger simultaneously. However, to ensure that the most conservative case was considered, for the maximum heat load evaluation, the assumption of 3000 GPM KC System flow delivered to the KF System heat exchangers is maintained by dividing KC flow rate to each KF heat exchanger by two (i.e.,3000 GPM of KC System flow through the common non-essential header, split into 1500 GPM per KF System heat exchanger). To maintain consistency in the analysis, the heat exchanger UAF was linearly extrapolated down to the KC flow assumed to be delivered to each heat exchanger (1500 GPM), resulting in a UAF of IE6 BTU /Hr-T. The results of this analysis are presented in Attachment 1.
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C It should be noted that the UAF values for higher versus lower KC System flow r'ates do not vary significantly, and the UAF values should not change significantly -
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- with operational time since both of these systems are clean and chemically treated.7
.l More significant to the results of the calculation is the assumption for KC flow rate -
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(shell side) to the KF System heat exchangers, as well as the assumed KC supply:
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' temperature to the shell side of the heat exchangers. Test data demonstrates that j
the KC System header to the KF System heat exchangers is capable of dehvermg a -
value greater than the assumed KC System flow rate of 3000 GPM (i.e.,
approximately 3500 GPM for a single train in' service). Additionally, the assumed value for KC supply temperature to the heat exchangers is based on maximum-predicted temperature during summer months.
l Operations monitors the SFP temperature and adjusts KC System flow with a ~
1
' manual loader on the Main Control Board to achieve the required heat removal.
- Additionally, the Catawba Nuclear Engineering Section monitors SFP temperature by procedure during the refueling evolution. During some procedures this monitoring takes the form oflogging the data per procedure, while in other -
procedures the SFP temperature requirement takes the form of a Limits and -
Precaution or a Prerequisite System Condition (i.e., no procedural actions may be taken if the SFP temperature is > 150 T). There is also an annunciator window which alarms at a SFP temperature of 135 T, which would cause Operations to enter the tanu'nciator response procedure.,
It is concluded that the assumptions and inputs are adequate for an analysis of SFP equilibrium temperature, and that there are adequate controls in place to arrest an increase in SFP temperature during either maximum or normal heat load conditions.
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. Q4) - At this time, do you have any plans for storage of any spent fuel from other Duke -
plants in either Catawba SFP which is noted as a possibility in the UFSAR of October 1993 (P.9-11)? What are your plans, if any, to deal with this?
A4). _No, we do not anticpate shipping any spent fuel from either Oconee or McGuire to Catawba. Current plans are to continue with dry stroage at Oconee, and to begin dry storage at McGuire in 2001. However, given the current uncertainty in the High Level Waste Program, it is possible that conditions could change such that transfer of spent fuel from one, or both of these facilities to CNS could become the -
more desirable option. Therefore, it is imperative that Duke Power Company keep.
all possible spent fuel storage options open.
Q5)
You state that " PANTHER" is a certified code. Explain how and by whom it has been certified.
A5)
" PANTHER" (Programmed Analysis of Nuclear Thermal Energy) is a computer code written by Catawba Nuclear Engineering and certified in a calculation file per the Duke Power Company Nuclear Generation Department Engineering Documents Manual process. It is cenified in Reference 6. PANTHER uses the ANSI /ANS-5.1 methodology for performing decay heat calculations. The ANSI /ANS-5.1 methodology has been deemed to be acceptable for decay heat calculations in Reference 7.
The ANSI /ANS-5.1 methodology provides guidance on decay heat calculations for a burnup period by the use of a series of exponential fit equations and constants for decay heat produced from each actinide, but provides no guidance on determination of the fraction of power produced by each actinide during burnup periods, since this information is specific to core design. The PANTHER computer code performs an evaluation of decay heat produced by nuclear fuel by dividing the total burnup period into burnup blocks, in which the fraction of power produced by each actinide is automatically updated for the each burnup block. The power fractions for each actinide are programmed into the PANTHER computer code based on ORIGEN2 results for a typical PWR core. ORIGEN2 runs were evaluated at different initial enrichment values to obtain a table of values of actinide power fractions versus burnup based on initial enrichment. The initial enrichment entered into the PANTHER computer code will determine which set of actinide power fractions are used. In this manner, the PANTHER computer code is capable of evaluating the effects ofincreasing the initial enrichment of discharge fuelin the spent fuel pool.
It should be noted that the Branch Technical Position 9-2 methodology has no such capability. Initial enrichment is not a variable in the BTP 9-2 methodology.
The BTP 9-2 methodology assumes that the decay heat power from fission products produced by different actinides would be the same as for "U (see Reference 8). Hence, the PANTHER computer code was written to address the issue of differences in initial enrichment of discharge fuel to the spent fuel pool.
The PANTHER computer code has been verified in Reference 6 to produce the same results as the ANSI /ANS-5.1 example problems. Additionally, the PANTHER computer code results have been verified to be similar but conservative as compared to actual tested heat load data and ORIGEN2 results.
No direct comparison can be made between the decay heat load results for this j
submittal and ;he FS AR results. The decay heat load results for this submittal were evaluated with a different methodology, as well as different inputs (i.e., the decay heat results for this submittal were evaluated with the ANSI /ANS-5.1 methodology which accounts for fission products produced from different actinides, versus the BTP 9-2 methodology, and ANSI /ANS-5.1 accounts for a
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more accurate, albeit conservative, method to account for fission produst neutron _
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capture versus the BTP 9-2 methodology; additionally, higher burnups were used ~
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i QS)
The layout of racks and assemblies in Figure 5.1 of your submittal of September 19,1994, shows a total of 1421 fuel assemblies in lieu of 1418. Explain these differences.
A6)
The current storage module layout provides 1421 fuel storage locations in each of the Catawba spent fuel pools. Though all 1421 storage locations can be accessed wiht the fuel handling mast, and all analyses supporting this amendment request consisered a minimum of 1421 fuel assemblies, actual storage ofirradiated and new fuel has been, and will continue to be governed by the 1418 limitation specified in the technical specifications.
The licensing basis established for the Catawba spent fuel storage pools was based on a projected capacity and a specific storage module assortment. Storage module installation, which was not completed until well afler initial startup of the plant, actually allowed for an additional 3 storage cells above the projected storage capacity. This explains the difference between the licensing limit of 1418 assemblies and the actual capacity of the pools being 1421 assemblies.
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. Q8).Section VIII.2, " Proposed FSAR Modifications" (page 8-12) of your September 19,1994 submittal contains a discussion of spent fuel pool cooling and states: "...
3 the results of an analyses (sic) based on these higher heat loads will be reflected in -
this section." Explain how this analysis differs from that reported in Section VI.8,
'" Spent Fuel Pool Cooling Considerations," ofyour submittal.
A8)
Section VIII 2 refers to the same analysis as is reported in Section VI.8.
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. Attachment 1 l
SFP Temperature Calculation Inputs and Results l
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f The Spent Fuel Pool Cooling (KF) System heat exchangers are counterflow shell L
. and tube heat exchangers, with one shell pass and two tube passes.~ Component
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Cooling is on the shell side and Spent Fuel Pool Cooling is one the tube side.
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. The equation below is used to calculate the temperature difference between -
j entrance and exit for both sides of the spent fuel pool cooling heat exchangers (the -
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Cooling (KF) System is the tube side):
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Component Cooling (KC) System is the shell side, and the Spent FuelPool G = [nc,(AT) i
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-i Where:
Q
= Heat Load, BTU /Hr.
m
= Mass Flow Rate, lbm/Hr j
c,,
= Specific Heat,1 BTU /lbm 0F 1
AT
= Temperature Difference Between Supply and Exit,0F -
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.The equation below is used to obtain a value for UAF from test data (heat load is 1
calculated from flow rate and AT with the first equation, and LMTD is calculated j
- from temperature test data). A derivation of this equation is also used below to.
t determine spent fuel pool temperature based on KC System and KF System -
j conditions:
l Q = UAF(IkfTD)
.l Where:
Q
= Heat Load, BTU /Hr U
= Heat Transfer Coefficient, BTU /ft2-Hr 0F
-l' A
= Heat Transfer Area, fl2 F
= Correction Factor (for heat exchangers other than double-pipe)
LMTD = Log Mean Temperature Difference ~
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IMID =
In(07h7 For the calculation ofLMTD:
A7; = KF,, - KC,,,
1 AT, = KF,,, - KF - KC,,,
g G _ (KF.u, - E.,)- k, - F - E.o,)
u UAF ~
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KF,,, - KC,,
ln g KF,,, - KF - KC,,,,
g s
KF,,, - E,,
KF,,, - KF - KC,,,
a Where:
,, _ (KF + KC,,, - K,,fF g
o The solution for KF,,,(supply to the KF System heat exchangers, or spent fuel pool temperature) yields the following:
KC., -(KF + KC,,,)e#
g KF
=
sup
,x Where the term "X" is as defined above.
The equation above is used to solve for equilibrium spent fuel pool temperatures.
Each variable on the right side of the equation is known, and the equation can be solved for Spent Fuel Pool Cooling (KF) System supply temperature (i.e., SFP temperature), based on known and/or assumed conditions.
Inputs:
a)
KC System Supply Temperature = 95 T b)
KC System Flow Rate = 3000 GPM Single Train (Normal Heat Load),1500 GPM Two Train Operation (Maximum Heat Load) c)
KC System AT = Calculation based on Mass Flow (equation 1) d)
KF System Flow Rate = 2300 GPM e)
KF System Heat Exchanger UAF = 1.267651E6 BTU /Hr-T (for assumption of 3000 GPM KC System flow to each KF Heat Exchanger), IE6 BTU /Hr-T (for assumption of 1500 GPM KC System flow to each KF Heat Exchanger) f)
KF System AT = Calculation based on Mass Flow (equation 1)
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c With 'two trains of the KF System in service for the maximum heat load case, the i
decay heat load used is half the calculated value, since each train is assumed to remove half of the heat load. Additionally, each KF System train is assumed to receive 1500 GPM ofKC System flow.
i The table below provides results assuming that 3000 GPM of KC System flow is.
delivered to each KF System heat exchanger, both normal and maximum decay =
heat load cases.
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" Maximum" Heat Load Case -
" Normal" Heat Imad Case
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0 0
SER Limit -
150 F SER Limit 140 F i
ANSI-57.2 Limit 140 F ANSI-57.2 Limit 140 0F 0
SRP Limit '
< 212 F SRP Limit 140 0F l
0 SFP Temperature Results SFP Temperature Results -
PANTHER 131.943 F PANTHER 127.%2 0F 0
ORIGEN2 129.026 0F ORIGEN2 The results tabulated above consider the heat load data presented in Attachment 2.
Inclusion of decay heat from the additional 12 cells filled with discharged fuel-(
increases the SFP equilibrium temperature from 131.943 'F to 131.9995 'F.
j (maximum decay heat load, PANTHER results).
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.The table below provides results assuming that 1500 GPM of KC System flow is j
delivered to each KF System heat exchanger for the maximum decay heat load i
case, and uses the decay heat load calculated with every SFP cell filled with j
discharged fuel. The results presented below assume a UAF of IE6 BTU /Hr 'F as l'
previously discussed.
" Maximum" Heat Load Case 0
SER Limit 150 F 0
ANSI-57.2 Limit 140 F 0
SRP Limit
< 212 F l
SFP Temperature Results f
PANTHER 145.1210F 0
ORIGEN2 141.115 F 6
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I Maximum Decay Heat Results Decay Time Enrichment Bumup '
Ouanitity Decay Heat -
(Days)
(U235 w/o)
(MWD /t)
(BTU /Hr)
-7 4.15 22326 44 7.76E + 06 7
4.4 19105 32 5.56E+06 7
4.15 38586 44 8.14E + 06 c,
7 4.4 36877 32 5.88E + 06 L
7 4.15 51817 20 3.81 E + 06 7
4.4 43552 12 2.24E + 06 7
4.4 44407 8
1.49E + 06 7
4.4 58239 1
1.94 E + 05
.l 25 4.15 47277 44 4.71 E + 06 25 4.4 42501 32 3.39E + 06 568 4.15 47277 44 7.02E + 05 j
568 4.4 42501 32 4.80E + 05 1110 4.15 47277 44
- 3.41 E + 05 1110 4.4 42501 32 2.29E + 05 1653 4.15 47277 44 2.00E + 05 1653 4.4 42501 32 1.33E + 05 2196 4.15 47277 44 1.43E + 05 2196 4.4 42501 32 9.48E + 04 2738 4.15 47277 44 1.21 E + 05 2738 4.4 42501 32 8.04E + 04 3281 4.15 47277 44 1.07 E + 05 3281 4.4 42501 32 7.11 E + 04 3823 4.15 47277 44 9.59E + 04 3823 4.4 -
42501 32 6.3 8E + 04 4366 4.15 47277 44 8.62E + 04 4366 4.4 42501 32 5.73E + 04 4909 4.15 47277 44 8.11 E + 04 4909 4.4 42501 32 5.39E + 04 5451 4.15 47277 44 7.80E + 04 5451 4.4 42501 32 5.19E + 04 5994 4.15 47277 44 7.35E + 04 5994 4.4 42501 32 4.89E + 04 6537 4.15 47277 44 7.08E + 04 6537 4.4 42501 32 4.71 E + 04 7079 4.15 47277 44-6.77E + 04 7079 4.4 42501 32 4.50E + 04 7622 4.15 47277 44 6.52E + 04 7622 4.4 42501 32 4.34 E + 04 8164 4.15 47277 44 6.29E + 04 8164 4.4 42501 32 4.18E + 04 Total Assemblies and Decay Heat =
1409 4.70E + 07 The above data (maximum decay heat load) includes the decay heat load data for the maximum decay heat load case. The exact result is 47,013,892 BTU /Hr. Inclusion of heat from the additional 12 cells (using the last decay heat data point) yields: 47,013,892 BTU /Hr + ((32+12)/32)
- 4.I8E4 BTU /Hr = 4.70713665E7 BTU /Hr.
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Normal Decay Heat Results Decay Time Enrichment Burnup Quanitity Decay Heat.
(Days)
(U235 w/o)
(MWD /tl
~ (BTU /Hr) 7 4.4 58000 76 1.47E + 07 568 4.15 47277 44 7.02E + 05 568
- 4.4 42501 32 4.80E + 05 1110 4.15 47277 44 3.41 E + 05 1110 4.4 42501 32 2.29E + 05 1653 4.15 47277 44 2.00E + 05 1653 4.4 42501 32 1.33E + 05 2196 4.15 47277 44 1.43E + 05 2196 4.4 42501 32 9.48E + 04 2738 4.15 47277 44 1.21 E + 05 2738 4.4 42501 32 8.04 E + 04 3281 4.15 47277 44 1.07E + 05 3281-4.4 42501 32 7.11 E + 04 3823 4.15 47277 44 9.59E + 04 3823 4.4 42501 32 6.38E + 04 4366 4.15 47277 44 8.62 E + 04 4366 4.4 42501 32 5.73E + 04 4909 4.15 47277 44 8.11 E + 04 4909 4.4 42501 32 5.39E + 04 5451 4.15 47277 44 7.80E + 04 5451 4.4 42501 32
- 5.19E + 04 5994 4.15 47277 44 7.35E + 04 5994 4.4 42501 32 4.89E + 04 6537 4.15 47277 44 7.08E + 04 6537 4.4 42501 32 4.71 E + 04 7079 4.15 47277 44 6.77E + 04 7079 4.4 42501 32 4.50E + 04 7622 4.15 47277 44 6.52 E + 04 7622 4.4 42501 32 4.34E + 04 8164 4.15 47277 44 6.29E + 04 8164 4.4 42501 32 4.18E + 04 Total Assemblies and Decay Heat =
1216
- 1.85E + 07 The above data is for the normal heat load case.
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References -
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'1)
H. P. Smith, Duke Power Company Calculation CNC-1201.30-00-0014, l
" Decay Heat and Spent Fuel Pool Temperature Calculation for Fuel j
Enrichment Upgrade."
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2)~
C. V. Parks (ORNL),'" Overview of ORIGEN2 and ORIGEN-S:
1 Capabilities and Limitations," High Level Radioactive Waste Management,-
-l Proceedings of the Third International Conference. See also Allen G. Croff (ORNL), "ORIGEN2: A Versatile Computer Code for Calculating the j
Nuclide Compositions and Characteristics of Nuclear Materials," Nuclear l
Technology,62, September,1983.
l 3)
Regulatory Guide 3.54, " Spent Fuel Heat Generation in an Independent
-l
. Spent Fuel Storage Installation."
4)
Kanji Tasaka, " Effects ofNeutron Capture Transformations on the Decay j
Power of Fission Products,"_ Nuclear Science and Engineering, 62,1977.
1
-l 5)
NUREG/CR-2397, " Fuel Inventory and Afterheat Power Studies of l
Uranium-Fueled Pressurized Water Reactor Fuel Assemblies Using the -
l SAS2 & ORIGEN-S Modules of SCALE with an ENDF/B-V Updated.
.I Cross Section Library."
i 6)
H. P. Smith, Duke Power Company Calculation CNC-1201.30-00-0013,
" Certification of the PANTHERI Computer Code for Performing Decay I
Heat Calculations."
7)
Regulatory Guide 1.157, "Best-Estimate Calculations of Emergency Core Cooling System Performance."
.i t
8)
Virgil E. Schrock, " Evaluation of Decay Heating in Shutdown Reactors,"
Progress in Nuclear Energy, Vol. 3,1979.
4 e
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