ML20009F658
| ML20009F658 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 06/30/1981 |
| From: | Feuerbacher R SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER |
| To: | |
| Shared Package | |
| ML20009F651 | List: |
| References | |
| XN-NF-78-51(NP), XN-NF-78-51[NP], NUDOCS 8107310386 | |
| Download: ML20009F658 (41) | |
Text
,
XN NR8-51[NP) e-f t
EXXON NUCLEAR CONTROL ROD DROP ACCIDENT ANALYSIS FOR BIG ROCK POINT JUNE 1981
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XN-NF-78-51(NP)
Issue Date: 7/06/81 EXXON NUCLEAR CONTROL R00 DROP ACCIDENT ANALYSIS FOR BIG ROCK POINT Prepared By R. L. Feuerbacher fl Y
/
APPROVED:
0 Jo Td NP/
R. B. Stout, Manager Date Neutronics and Fuel Management m
APPRPUED: /
d
-G. F.80isisy, Manager
/ Orte Reload Fuel Licensing APPROVED:
d-YSI/
G. A. Sofer,-
ngings/
ger Date Nuclear Fu ring 9f ERON NUCLEAR COMPANY,Inc.
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-i-XN-NF-78-51(NP)
Table of Contents Section Page
1.0 INTRODUCTION
AND
SUMMARY
1 2.0 CONTROL R00 DROP ACCIDENT ANALYSIS.................
3 2.1 CONTROL h ) DROP ACCIDENT DESCRIPTION.............
3 2.2 GENERAL DESCRIPTION OF ANALYSIS................
4 2.3 TRANSIENT ANALYSIS METHOD...................
5 2.3.1 T ra n s i en t Comp u te r Mo del................
5 2.3.2 Control Rod Drop Analysis Method............
7 3.0 RESULTS 11
)
3.1 GENERAL DESCRIPTION......................
11 3.2 PARAMETRIC RESULTS 13 I
3.3 APPLICATION OF PARAMETRIC RESULTS...............
14
4.0 REFERENCES
31 APPENDIX A 32
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-ii-XN-NF-78-51 (NP)
List of Tables t
Table n
Page 2.1 BIG ROCK POINT CORE CONDITION FOR CONTROL R00 DROP ANALYS 17 t
2.2 BIG ROCK POINT AXIAL EXPOSURE DISTRIBUTION FOR CONTROL R00 DROP ACCIDENT ANALYSIS la 2.3 BIG ROCK POINT KEY KINETICS PARAMETERS F
)
CONTROL R00 DROP ANALYSIS....... OR 4
19 2.4 BIG ROCK POINT CONTROL R00 OROP ANALYSIS INPUT DELAYED NEUTRON CONSTANTS................
1 3.1 BIG ROCK POINT CONTROL ROD DROP ACCIDENT ANALYSIS NOMINAL RESULTS......................
21 s
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-lii-XN-NF-78-51 (NP) o c'.=
List of Figures Figure Page r
1.1 BIG ROCK P0 INT R00 GR0P ACCIDENT PARAMETRIC DEPOSITED ENTHALPY VS CONTROL ROD WORTH...............
22 l.2 BIG ROCK POINT ROD DROP ACCIDENT PARAMETRIC DEPOSITED ENTHALPY VS D0PPLER COEFFICIENT..............
23 i
1.3 BIG ROCK P0 INT R0D DROP PARAMETRIC DEPOSITED ENTHALPY VS DELAYED NEUTRON FRACTION 24 l
2.1 BIG ROCK P0 INT SCRAM BANK INSERTION VS TIME FROM RECEIPT OF SCRAM SIGNAL..................
25 i
2.2 BIG ROCK POINT CYLINDRICAL R0D DROP ACCIDENT MODEL 26
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2.3 CONTROL FRACTIONS a VS a FOR CENTER CONTROL R0D FULL IN OR OUT27 j
2 2.4 BIG ROCK POINT CALCULATED SCRAM BANK FRACTIONAL WORTH........ 28 3.1 BIG ROCK POINT PEAK DEPOSITED ENTHALPY VS CONTROL R00 WORTH....
i 3.2 BIG ROCK P0 INT CORE AVERAGE POWER VS TIME DURING CONTROL ROD DROP ACCIDENT......................
30 A.1 COMPARISON OF CORE AVERAGE POWER DUI,:N1 CONTROL R0D DROP ACCIDENT FOR TYPICAL AND UNIFORM AXIAL EXPOSURE DISTRIBUTION.........35 l
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XN-NF-78-Si (NP)
1.0 INTRODUCTION
AND
SUMMARY
Exxon Nuclear Company (ENC) hcs performed a reference control rod drop accident analysis for Consumers Power Company's Big Rock Point Boiling Water Reactor.
A control rod drop accident was simulated for an exposed core con-I' s:iting entirely of ENC manufactured G-3 reload fuel assemblies.
The second reload batch of the G-3 fuel type will be loadc:1 for Big Rock Point's next e
{
operational period (Cycle 16). This reference control rod drop accident analysis applies directly to Cycle 16 operation and also generically to future t
operating cycles loaded with fuel similar to the G-3 design.
J A transient, two dimensional (r-z cylindrical geometry) computer madei
(
with fuel temperature reactivity feedback is utilizea for this analysis.
The model simulates the reactivity insertion caused by a control rod being rapidly removed from the reactor core followed by the subsequent shutdown due to Doppler feedback and the scram rod bank entering the core. Prior to the start of the control rod drop accident, the initial core condition assumed i.s a hot standby, near critical state.
For the analysis, a bounding minimum scram worth curve is employed to ensure that the reference rod drop accident results will apply for future cycles.
Finally, the transient model computes the limiting conse-I quences of the control rod drop accident in terms of the resultant peak energy j
deposition in the fuel.
I
~
In order to apply this reference rod drop analysis specifically to the
')
Big Rock Point Cycle 16 operation as well as to future cycles, all important fuel assembly and core neutronic parameters are enveloped.
The core variables that significantly affect the control rod drop accident corwequences are dear-mined to be:
I b
~
XN-NF-78-51(NP)
The subject control rod reactivity worth e
e The Doppler coefficient The power peaking (after the control rod has been completely e
removed from the core)
The delayed neutron fraction.
e These parameters encompass the effects of potential variations in core loading patterns, fuel assembly enrichments, core exposures, and exposure distributions.
The results of this reference rod drop accident analysis are comprehen-sively parameterized with respect to maximum control rod worth, the Doppler reactivity coefficient, inaximum power peaking, and the delayed neutron fraction.
The parametric results of the Big Rock Point rod drop accident analysis are summarized in Figures 1.1 through 1.3 in terms of peak energy deposition in the fuel.
These parametric results are presented in a format that facilitates application of this reference red drop accident analysis to Big Rock Point ope rating evcles.
Consequently, tne results of these analyses may be applied to not only the core loading for Cycle 16 but to future Big Rock Point core conditions or configurations.
A_ sensitivity analysis was performed to determine the effects of the axial exposure distribution upon the outcome of the control rod drop accident analysis.
This sensitivity analysis indicates that a typical axial exposur e distribution aith the higher exposed fuel peaked in the lower half of the core is conserva-tive as compared to a uniform axial exoosure distribution.
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-)- XN-NF-78-51 (NP) eM l
l 2.0 CONTROL R00 DROP ACCIDENT ANALYSIS 2.1 CONTROL R00 DROP ACCIDENT DESCRIPTION The sequence of events postulated during a control rod drop accident are described as follows:
}
1.
A fully inserted control rod becomes decoupled from its drive mechanism.
I 2.
The drive mechanism is completely removed during the n1rmal rcd withdrawal sequence with the control rod remaining stuck in the reactor core.
3.
Normal withdrawal sequence continues in the approach to criti-cality, despite the disconnected and stuck control rod.
4.
At the worst time during the rod withdrawal sequence, the stuck rod suddenly becomes freed and falls out of the core.
5.
The reactor subsequently becomes prompt critical resulting in t
a rapid power increase that eventually initiates a scram signal.
6.
The core power reaches a maximum and then decreases rapidly due to Doppler reactivity feedback.
7.
Subsequent power behavior depends upon the dropped rod velocity and worth, the scram delay time, and the scram bank velocity and reactivity worth.
(The dropped rod.is not included in the J
scram bank).
8.
The reactor becomes and remains subcritical due to the - mbina-g tion of the scram bank insertion and Doppler feedback.
.1
_t GR m
. XN-NF-78-51(NP) 2.2 GENERAL DESCRIPTION OF ANALYSIS The limiting consequence due to a control rod drop accident is col-culated in terms of peak energy deposition in the fuel. Guideline values of i
stcred energy content, corresponding to various degrees of fuel and/or cladding j
failure, have been established based on experimental studies (References 1 and 2).
l Thus, the objective of the control rod drop calculations is 'to detemine if any fuel will exceed these guideline values during the unlikely occurrence of
~
a control rod drop accident.
The general core conditions (hot standby, near critical state) assumed
}
[
in this rod drop analysis for Big Rock Point are outlined in Table 2.1.
For tnis analysis, the reactor core is assumed to be at essentially equilibrium cycle conditions loaded entirely with G-3 reload fuel assemblies. The core i
axial exposure distribution (presented in Table 2.2) input into this transient analysis is approximately the same as that actually calculated for the start of Cycle 15 operation.
'\\
The scram bank insertion velocity and delay time are obtained from
.I Big Rock Point Technical Specifications that state.
10% of stroke 0.6 seconds (maximum time after receipt of scram signal) g 90% of stroke 2.5 seconds (maximum time after receipt of scram signal)
The reference scram bank insertion as a function of the time from the receipt of the scram signal is plotted in Figure 2.1.
The statically computed scram bank worth is conservatively set at 89 x 10-3 ok/k in order to envelope future cycles.
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' XN-NF-78-51(NP)
Since there are no control rod velocity limiters in Big Rock Point, the control rod dropped is assumed to be a free falling object accelerated by gravity.
In the analysis, the postulated stuck rod begins to fall at the
)
start of the transient calculation after a static power distribution has been solved.
To enable the appl. :ation of this reference control rod drop analysis to the Big Rock Point Cycle 16 operation as well as to -future cycles, all impoi: ant fuel assembly and core neutronic parameters are conservatively repre-t sented. The significant kinetics parameters used in the reference rod drop analysis are presented in Table 2.3.
The Big Rock Point Cycle 16 core will consist of a mixture of G fuel types except for four F type assemblies which will be placed in low power locations not adjacent to control rods. These F types will therefore have no i
significant effects upon the control rod crop accident consequences. As stated previously, the reference control rod drop analysis is performed for a full I
core of all reload fuel. The G-3 fuel type has neutronic characteristics similar to and compatible witn the previous G type designs (Reference 4).
)
Therefore, the results of th'e reference control rod drop analysis can be applied to Cycle 16 by conservatively calculating the appropriate kinetics parameters discussed in Section 3.0.
3 2.3 TRANSIENT ANALYSIS METHOD 2.3.1 Transient Computer Model i
A version of the XTRAN computer code (Reference 5) applicable y
to BWR cores is utilized for the Big Rock Point control rod drop accident I
oneos t
a
. XN-NF-78-51(NP)
(
analysis. The XTRAN code has been specifically developed to analyze the con-4 I
trol rod drop accident.
"TRAN is a two limensional (r-z cylindrical geometry) g computer program which solves the space and time dependent neutron diffusion equation with fuel temperature and moderator density reactivity feedbacks.
i' XTRAN employs a nodal method based directly on a one energy group finite dif-i i
ference technique for the solution of the time dependent neutron diffusion equation. The one-group cross sections used in the iterative flux solution
- ?
g are determined from input two-group values and modified at each time step by thermal feedback.
c' The space and time dependent neutronic model incorporated in XTRAN is capable of computing a rapid reactor transient initiated by the reactivity insertion due to a control rod being removed from the core. Since the model utilizes two-dimensional (r-z) geometry, the code can calculate the rapidly changing flux distribution as a control rod leaves the reactor core t
and the scram rod bank subsequently enters the reactor core.
XTRAN initially determines the static flux and power distri-bution corresponding to the problem input. The initial time step for the transient analysis is 0.0001 seconds. The code then automatically determines the time step interval based on the number of iterations necessary to achieve cunvergence. This method permits small time steps during times of large changes in power level, and conversely, large time steps during periods of i
slow perturbations. Therefore, the code efficiently solves the transient problems without the user choosing time step sizes.
For the Big Rock Point control rod drop analyses, calculations are performed for a total time interval of six seconds.
.)
i
XN-NF-78-51 (NP) 1 I
Six groups of delayed neutron precursors are employed in the transient analyses. The decay constants and delayed neutron fractions utilized in the Big Rock Point rod drop analysis are presented in Table 2.4.
These decay constants are obtained from Reference 6.
2.3.2 Co'ntrol Rod Droo Analysis Method The following is a step-by-step description of the procedure employed to perform the reference control rod drop accident analysis for Big Rock Point.
1.
All input two-group cross sections, both uncontrolled and controlled, are calculated using standard ENC design methods for the hot standby conditions outlined in Section 2.2. ~
2.
The reactor core is subdivided into three major radial zones that represent the total volume of the core (84 fuel assemblies). The core is subdivided into nine nodes axially.
The center radial zone is the dropped rod zone.
Its e
volume is equivalent to a module of four-fuel assemblies.
~
The next inner zone is a partially controlled zone e
l with a variable control fraction referred to as a).
8 This radial zcne is further divided into two radial subzones of equal widths approximately equivalent to g
(
an assembly pitch. The scram rod bank enters this zone.
The outer radial zone is also a partially controlled e
zone with a variab? ? control fraction, a This 2
i radial zone is subdivided into two radial subzones of l
onene 1
XN-NF-78-51 (NP) equal widths approximately equivalent to an assembly pitch. The scram rod bank also enters this zone.
The cylindrical core geometry modeled in XTRAN for the rod drop accident simulation is illustrated in Figure 2.2.
The control fractions of the outer two radial zones, 3.
and a, are varied to obtain the desired reference aj 2
control rod worths of 10, 20, and 30 mk. Static XTRAN solutions are made to calculate the control fractions.
With the central zone controlled, an iterative search is made for the amount of control necessary in the outer zones to achieve Keff = 1.000.
Figure 2.3 represents a plot of a2 versus a) for K
= 1.000 with the center etf zone fully controlled. With the center zone uncontrolled, a series of calculations are performed to determine the amount of cor. trol necessary in the outer zones to yield j
K* ff = 1.000 + aK e
CH: where K* ff denotes the reactivity e
--of the core with the center rod removed and AK is the CR i
desired rod worth. The calculated control fractions, 2 versus a), are graphically depicted in Figure 2.3 for a
control rod worths of 10, 20, and 30 mk.
The intersection points o'n the generated curves are the required pairs of aj and a that will result in a Keff = 1.000 when the 2
center control rod is inserted and a K* ff = 1.000 + AK e
CR I
GR4les
. XN-NF-78-51 (NP) l b
with the center control rod removed. These sets of control fractions are utilized to produce the desired I
control rod worths in the transient analyses.
1 4.
From the static XTRAN calculations in Step 3, the rel-I ative axial x radial power peaking factor is determined with the center control rod fully out for the three characterized control rod worths. Doppler and moderator i
I density feedbacks are conservatively not included. For the parametric rod drop studies, the relative power g
peaking factors are varied for each control rod worth by changing the released energy per fission constants
}
in the center control rod zone.
5.
The input transient scram bank reactivity worth function
}
employed in the rod drop accident is presented in Figure 2.4.
Since for actual cycle design. analysis the scram bank worth curve is calculated statically, a series of
}
XTRAN static eig: nyalue calculations are also performed to generate the input static scram worth curve presented 1
in Figure 2.4.
Doppler and moderator density feedbacks are conservatively not included. The input static scram bank reactivity worth is calculated to be 89 x 10-3 ak/k j
which should be definitely bounding for future cycles.
For the actual cycle design calculations, the static scram bank should be demonstrated to be worth at least
-3 89 x 10 ak/k in order to apply this reference rod drop analysis.
anans
.i,
XN-NF-78-51 (NP) l 6.
For the Big Rock Point reference control rod drop accident analysis, an idiabatic boundary condition is assumed at the fuel pellet-gap interface.
In other words, no heat transfer to the coolant is allowed. This
{
is a conservative assumption since in this way the peak deposited enthalpy in the fuel is determined from the sum of all the energy absorbed by the fuel during the transient period with no credit for heat escape from the fuel.
The calculational method is very automated since XTRAN was developed specifically for the control rod drop accident analysis. At the start of the transient solution (time = 0), the center control rod immediately begins to fall from the core. The rod is removed at the rate of input acceleration.
When the core peak reaches scram level magnitude, the power trip is signaled.
The scram rod bank then begins entering the outer two zones (see Figure 2.2) after an input delay time. The total time analyzed for this transient in the rod drop studies is six seconds.
e
.)
1
. XN-NF-78-51 (NP) 3.0 RESULTS 3.1 GENERAL DISCUSSION The Big Rock Point reference control rod drop accident analysis has I
been completed for the core conditions outlined in Tables 2.1 and 2.2 and the key neutronit carameters presented in Table 2.3.
The nominally calculated i
results, in terms of peak deposited enthalpy in the fuel, are summarized in
~
Table 3.1 and plotted in Figure 3.1.
't.e peak enthalpy results in Table 3.1 are presented as axial x s
radial assembly depositions. Thus, in computing the peak deposited enthalpy in the assembly's " hottest" rod, the axial x radial peaks are to be multi-plied by the assembly local pin power factor. The local pin power is the 9
peak to average power in the assembly as determined by auxiliary assembly dif-fusion theory calculations. For example, the total peak deposited enthalpy 1
is calculated using Table 3.1 for the following parameters:
Control rod worth (mk) 10
=
3 0
-6 Doppler coefficient (ak/k/ F)
= -9.52 x 10 I
Axial x radial assembly peaking factor 2.31
=
Delayed neutron fraction, s
.00525
=
i Assembly local peaking factor 1.20
=
.1 J
.1 J
i J
~
XN-NF-78-51 (NP) l i
In order to illustrate the mechanics of the control drop accident, the relative core power experienced during th simulated transient is presented
'i I,
as a functici of time in Figure 3.2.
The three characterized rod worths (10, j
20, and 30 mk) for a -9.52 x 10-6 0
ak/k/ F Doppler coefficient and a.00525 a are depicted.
As the postulated stuck control rod is initially removed from
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t5e core, Figure 3.2 shows that the power begins to increase rapidly after approximately 0.4 seconds.
The scram power trip set at 122% of rated reactor power occurs after 0.37 seconds for the 30 mk rod, 0.42 seconds for the 20 mk r'cd, and 0.56 seconds for the 10 mk rod.
Due to the rapidly increasing reactor power, the fuel temperature also rises quickly causing the Dcppler feedback to compensate the reactivity i,
insertion produced by the falling rod. The primary power peak, shown in Fig-ure 3.2 foi the three rod worths, occurs when the Doppler feedback exactly balances the dropped rod reactivity insertion. After the primary peak, the Doppler feedback becomes the doofnating factor, and the core power is reduced.
For the higher e.atrol rod worths and power peaking factors, the Doppler feedback quickly arrests the reactivity insertion before the control rod is completely removed.
Consecnently, there is a smaller, secondary power increase due to the additinnd reactivity added by the remainder of the con-trol rod.
In other words, the reactivity insertion by the falling control
'. )l rod once again becomes the dominating factor. No secondary power increase occurs for the 10 mk rod case since the control rod is completely out after 0.60 seconds.
Hence, there is no additional reactivity to be inserted af ter the Dopper feedback begins to reduce the reactivity.
onese t
. XN-NF-78-51 (NP)
The scram rod bank begins to enter the core 0.375 seconds af ter receipt of the scram signal. Therefore, the scram bank begins to enter af ter 0.74 seconds of elapsed time for the 30 mk rod. 0.79 seconds for the 20 mk rod, and 0.93 seconds for the 10 mk rod.
Figure 3.2 shcws when the power is reduced by the scram rod bank.
Based on these transient results, it is evident th'at the Doppler feedback is the primary mechanism for shutting the reactor down during a control rod drop accident. The scram reactivity worth function is only of
-secondary importance.
3.2 PARAMET9IQ_RE3ULTS 6
In order to apply the reference Big Rock Point control rod drop accident analysis to Cycle 16 cperation and future cycles, the calculated 3
results have been comprehensively parameterized. The important parameters
~
J that significantly affect the results of the control rod drop analysis are control rod reactivity wcrths, Doppler coefficients, power peaking -factors, 5
and delayed neutron fractions. These variables encompass the effects of core 3
loading patterns, fuel assembly enrichments, core exposures and exposure distributions. Thc results of these parametric studies may therefore be i
applied to the specific core loading for Cycle 16 and also to future Big Rock Point core conditions or configurations.
1 The axial x radial peak deposited enthalpy in the fuel is presented as a function of control rod reactivity worth and axial x radial power peaking factors (statically calculated with the subject control rod entirely withdrawn 1
and no Doppler feedback) in Figure 1.1. These peak results are given for a
-6
-9.52 x 10 ak/k/ F Doppler reactivity coefficient and a.00525 8.
.)
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GAete i
. XN-NF-78-51 (NP)
Figure 1.2 is to be employed to convert the peak deposited enthalpy obtained from Figure 1.1 for a -9.52 x 10-6 ak/k/ F Doppler coefficient to that for the desired Doppler coefficient.
In Figure
'.2, the relative depos-ited enthalpy factor is presented as a function of Doppler coefficient.
The relative peak deposited enthalpy is presented in Figure 1.3 as a function of the deiayed neutron fraction, 8.
This figure is to be employed to include the effect of 8 on 1.he peak enthalpy obtained from Figures 1.1 and 1.2.
As demonstrated by Figure 1.3, the delayed neutron fraction is of second-acy importance in the control rod drop analysis.
As stated in Section 3.1, the assembly local pin power factor must be applied to these axial x radial peak enthalpy results to determine the peak deposited enthalpy. The initial fuel enthalpy also must be added to compute the total peak enthalpy accrued during the rod drop accident.
Section 3.3 outlines a. sample case for applying these parametric rod drop accident results to determine the peak deposited enthalpy.
i 3.3 APPLICATION OF PARAMETRIC RESULTS For a sample calculation, the peak deposited enthalpy resulting from hypothetical conditions is evaluated using the parametric results presented in Section 3.2.
The conditions prescribed for this sample case are as follows:
Maximum insequence control rod worth (mk) 12
=
U Doppler coefficient (ak/k/ F)
-6
-10.1 x 10
=
Axial x radial assembly peaking factor 3.90
=
Delayed neutron fraction, s
.0058
~
=
Assembly local peaking factor 1.20
=
. XN-NF-78-51 (NP) 1 J
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1 I
1 1
I The key neutronics parameters used for the actual control rod drop accident evaluation are to be calculated for each cycle with a core simulator model and other peripheral assembly design calculations. The most severe con-
)
trol rod to be dropped is normally the maximum insequence rod worth at hot standby conditions occuring at the peak core reactivity point in the cycle.
The peak axial x radial power peaking factor is usually located in the same module as the highest worth rod. This axial x radial power factor (also referred to as the " nodal" power peaking) is calculated using a core simulator model.
The calculation is to be performed at a hot standby condit4ans with the dropped rod completely removed and the scram bank not inserted. Nc Doppler feedback is included.
ondos
- XN-NF-78-51(NP)
The Doppler reactivity coefficient as presented in Figure 1.2 is the differential coefficient evaluated at hot standby conditions (583 F) for U
uncontrolled assemblies.
In the reference transient analysis, the XTRAN model spatially treats the controlled and uncontrolled nodes with appropriate Doppler coefficients. However, to facilit. ate application of the parametric results for similar fuel types, only the uncontrolled Doppler coefficient needs to be calculated for each cycle in order to be consistent kith the reference control rod drop accident.
The assembly local power peaking factor is to be calculated with peripheral fine mesh diffusion theory calculations.
If justified, additional engineering factors can be applied to the control rod drop analysis by including the conservative factors with the local peaking factor.
The delayed neutron fraction, B, is to be evaluated at the appro-priate c: 'e exposure for each cycle.
For fuel designs with similar enrich-ments, s is primarily exposure dependent.
The same procedure, as applied here for a sample case, can be employed to compute the peak anthalpy resulting from a rod drop accident for Big Rock Point Cycle 16 operation as well as future cycles loaded with fuel similar to the G-3 design.
I t
==.
=
' XN-NF-78-51 (NP)
Table 2.1 Big Rock Point Core Condition for Control Rod Drop Analysis 1
'l Core loading: G-3 type reload fuel.
Beginning of cycle (B0C) core average exposure: 9.9 GWD/MTM.
Fuel exposure distributions: distribution axially; uniform radially.
Fuel temperature at beginning of transient: 583 F.
Moderator conditions (saturated):
583 F; O void fraction.
U I
Reactor power level at beginning of transient: 240 x 10 MWt.
-6 Scram bank insertion velocity: 2.46 ft/sec.
Scram delay time: 0.375 sec.
Dropped rod acceleration:
-32.2 ft/sec.2,
Scram power (122% rated): 292.8 MWt.
Static scram bank worth: 89. x 10-3 ak/k.
Initial fuel stored enthalpy:
- 18. cal /gm.
O J
. XN-NF-78-51(NP)
Table 2.2 Big Rock Point Axial Exposure Distribution For Control Rod Drop Accident Analysis i
Axial Fuel Exposure Node (GWD/MTM) l 9 (top)
I 8
I
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6 5
4 3
2 1
1 (bottom)
Core Average
.)
Exposure il 1
la 1
al e
. XN-NF-78-51 (NP)
Table 2.3 Big Rock Point Key Kinetics Parameters For Control Rod Drop Analysis Control Rod Worth (mk):
10.0 20.0 30.0 Power Peaking * (Relative axial x radial factor):
2.31 3.12
}j, 3.70 5.00 I
I Dopoler' Coefficient **
(ak/k/0F):
-10.71 x 10~6
- 9.52 x 10-6
- 8.33 x 10-6 I
'l Delayed Neutron Fraction (i):
.00400
.00525 i
.00650 l
j Static, hot standby core conditions with the control rod to be dropped fully removed.
i
- Differential Doppler coefficient evaluated at hot standby conditions (5830F) for uncontrolled assembly 3
configuration.
J J
J
.. =
- XN-NF-78-51(NP) 1 Table 2.4 Big Rock Point Control Rod Drop Analysis Input Delayed Neutron Constants Delayed Neutron Fractional Group Decay Constant Group j e
6M A
3 j (sac'I) 1 h
1 0.038 0.0127 i
}
2 0.213 0.0317 a
3 0.188 0.115
'i 4
0.407 0.311 k
5 0.128 1.40
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6 0.026 3.87 1.000 I
2 1
u i
e 0
.w 1
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1 9
--m--p-g
~~w--e+---
,e_,w,,..--.ewyyw.
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- W-f ti
- -e"'
r t-e' T
+ ='
--M-
-'-r*-v
--ww* e v' r w
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u u~
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9 Table 3.1 Big Rock Point Control Rod Drop Accident Analysis Nominal Results Control
- Doppler Power Peaking
- Delayed Peak Deposited **
Rod Worth Coefficient (Axial x Radial Neutron Enthalpy 6
Ak (mk)
(10 k/k/ F)
Relative Factor)
Fraction,ii (cal /gm) 10
-10.71 2.31
.00525 20
-10.71 3.12
.00525 30
-10.71 3.70
.00525 10
- 9.52 2.31
.00525 20
- 9.52 3.12
.00525 S
30
- 9.52 3.70
.00525 i
9 10
- 8.33 2.31
.00525 20
- 8.33 3.12
.00525 30
- 8.33 3.70
.00525 Control rod rc::tisity worth and relative power peaking factor. determined by static calculations.
- The peak deposited enthalpy results are axial x radial assembly values. To obtain E
the total peak deposited enthalpy, apply the local peaking factor for the assembly.
q Finally, add the initial fuel enthalpy to compute the total peak enthalpy accumulated in the fuel.
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Time from Receipt of Scram Signal 9
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Rod Drop Accident i
f XN-NF-78-51 (NP) i 4.0 ILEFERT.NCES 1.
C. J. Paone, et. al., Rod Drop Accident Analysis for large Boiling Water Reactors, NED0-10527, 72 NED 18, Class 1, March, 1972.
2.
J. E. Grund, et. al., Subassembly Test Program Outline for FY 1969 and FY 1970, IN-1313, August, 1969 (100-17277).
3.
iechnical Specifications for Big Rock Point Plant, Consumers Power Company, Docket No. 50-155, As amended through October 17, 1977.
4.
Design Report Big Rock Point Reactor Reload G Fuel, JN-72-17, July 31, 1972.
(Addenda 1 through 4) 5.
J. N. Morgan, XTRAN-PWR: A Comouter Code for the Calculation of Rac.id Transients in Pressurized Water Reactors with Moderator and Tuel Temperature Feedback, XN ",C-32, September, 1975.
6.
R. J. Tuttle, Delayed-Neutron Data for Reactor-Physics Analysis, Nuclear Science and Engineering:
56, 1975.
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-32 XN-NF-78-51(NP) 1-1 APPENDIX A R0D DROP ACCIDENT ANALYSIS ADDITIONAL STUDIES A.1 AXIAL EXPOSURE DISTRIBUTION SENSITIVITY A sensitivity analysis was perfomed to determine the effects of the axial exposure distribution upon the outcome of the control rod drop acci-j dent analysis. The reference rod drop results presente<' 'n Section 3.0 are based on a typical axial exposure distribution (obtained at BOC 15). For this sensitivity analysis, a rod drop accident case was simulated assuming a uni-l form axial exposure distribution with the same average exposure as the refer-J ence analysis The particular case reanalyzed was the 20 mk control rod with an axial i
-6 x radial peaking of 3.70, a -9.52 x 10 ak/k/ F Doppler coefficient, and a J'
.00525 5.
The transient results indicate that the unifonn axial exposure distribution is not as conservative as the nonunifonn distribution. The peak I
axial x radial deposited enthalpy calculated for the uniform case is The primary reason for the calculated difference in deposited enthalpies j,
is the effect of the axial exposure distribution upon the axial power shape.
The reference case with the higher exposure peaked towards the core bottom 3
produces an axial power shape peaked in the upper half of the core. On the other hand, the uniform axial exposure distribution produces an axial power J,
J J
. XN-NF-78-51(NP) shape that is symetric about the core centerline. As compared to the uniform case, the power rise for the reference case begins sooner because the falling rod leaves the nodes with higher power earlier. The reactivity insertion rate for the reference case is thus higher during the initial part of the f
transient, and consequently, the reference case reaches its primary power burst before the uniform exposure case. The peak cc e power is approximately.equiv-i' alent for both cases, but the primary power peak occurs at 0.45 seconds for the reference case and 0.50 seconds for the unifonn case.
Since the drop-ped control rod requires 0.6 seconds to fall completely out of the reactor i
core, there is a longer period of time between the primary peak and the rod l
out point for the reference case. Therefore, the reference case produces a higher secondary power increase than the unifonn case. This increased second-ary peak is the principal cause of the peak enthalpy difference between the reference and uniform axial exposura case.
A secondary factor contributing to the difference between the uniform i
and nonuniform axial exposure transient results is the scram. Since the scram rods enter from the bottom of the core, the scram rods reach the nodes with higher powers (and thus higher deposited enthalplies) earlier for the uniform case than the reference nonuniform case. Therefore, the scram bank tenninates l
power generation in the paak enthalpy nodes earlier in the uniform exposure case than in the reference case. Since the impact of the rod drop transient is mainly constrained by Doppler feedback (as shown in Section 3.1), this effect due to the scram bank upon the power distribution is minimal.
A com-parison of the core power histories for the nonuniform and uniform axial exposure transients is presented in Figure A.1.
XN-NF-78 ~1(Np)
This sensitivity study demonstrates that a typical axial exposure dis-tribution with the higher exposed fuel in the lower half of the core is con-serva tive. The least conservative case would be a hot stanoby power profile with the peak near the bottom of the core.
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Drop Accident for Typical and Uniform Axial Exposure Distribution I
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XN-NF-78-51(NP)
Issue Date: 7/06/81 EXXON NUCLEAR CONTROL R0D DROP ACCIDENT ANALYSIS FOR BIG ROCK POINT Distribution J. L. Maryott G. F. Owsley H. G. Shaw (5)
G. A. Sofer R. B. Stout Document Control (5) l 4
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