ML20134K383
| ML20134K383 | |
| Person / Time | |
|---|---|
| Site: | Cooper  |
| Issue date: | 11/30/1996 |
| From: | Braden D, Sadeghi N GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20134K368 | List: |
| References | |
| GENE-A13-00395, GENE-A13-00395-01, GENE-A13-395, GENE-A13-395-1, NUDOCS 9702130272 | |
| Download: ML20134K383 (36) | |
Text
_ _ _ _ _ _ _ _ _ _ - - _ - _ _ _ _ _ - - - - - - - - - - - - - -
L GENuclear Energy 173 Curtner Ave SanJose, CA 95123 GENE-A13-00395-01 DRF A13-00395 Class I APPLICATION OF THE " REGIONAL EXCLUSION WITH FLOW-BIASED APRM NEUTRON FLUX SCRAM" STABILITY SOLUTION (OPTION I-D)
TO THE COOPER NUCLEAR STATION Licensing Topical Report November 1996 Prepared for Cooper by GE Nuclear Energy 9702130272 970210
^
PDR ADOCK 05000298 P
a GE Nuclear Energy I73 Curtner Ave San Jose, CA 95125 GENE-A13-00395-01 DRF A13-00395 Class I APPLICATION OF THE " REGIONAL EXCLUSION WITH FLOW-BIASED APRM NEUTRON FLUX SCRAM" STABILITY SOLUTION (OPTION I-D)
TO THE COOPER NUCLEAR STATION Licensing Topical Report i
November 1996 1
Prepared for Cooper by GE Nuclear Energy Nader Sadeghi, Q'
/d M b-Approved y:
Approved by:
,D. S. BrQe,
Technical Account Manager Project Manager
GENE-A13-00395-01 IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT Please Read Carefully The only undertakings of GE Nuclear Energy respecting information in this document are contained in the contract between the Customer and GE Nuclear Energy, as identified in the i
purchase order for this report, and nothing contained in this document shall be construed as changing the contract. The use of this information by anyone other than the Customer for any purpose other than that for which it is intended is not authorized; and, with respect to unauthorized use, GE Nuclear Energy makes no representation or warranty and assumes no liability as to the completeness, accuracy, or usefulness of the information contained in this document.
1 i
i i
GENE-A13-00395-01 TABLE OF CONTENTS i
Section lidt East LIST OF TABLES
- iv.
LIST OF FIGURES y
ABSTRACT' vi 1
INTRODUCTION 1-1 r
.l.1 Historical Perspective 1-1
(
l.2 BWR Owners' Group Response 1-2 1.3 Option I-D Solution 1-2 i
1.4 Applicability of Option I-D to Cooper 1-4 2
SUMMARY
AND CONCLUSIONS 2-1 3
APPLICATION OF BWROG STABILITY LONG-TERM SOLUTION REGIONAL EXCLUSION METHODOLOGY 3-1 3.1 Void Coefficient 3-1 3.2 Thermal-hydraulic Data 3-1 3.3 Hot-channel Axial Power Distribution 3-1 3.4 Average-channel Axial Power Distribution 3-2 3.5 Radial Power Distribution 3-3 3.6 Pellet-clad Gap Conductance 3-3 3.7 Miscellaneous Input Values 3 4 REGIONAL EXCLUSION RESULTS 4-1
.5 APPLICATION OF BWROG STABILITY LONG-TERM SOLUTION DETECT AND SUPPRESS METHODOLOGY 5-1 5.1 Licensing Compliance 5-1 5.2 Methodology Overview 5-1 5.2.1 Pre-Oscillation MCPR 5-2 5.2.2 Statistical Calculation of Hot Bundle Oscillation Magnitude 5-3 5.2.3 MCPR Performance of the Hot Bundle 5-5 5.3 Final MCPR Calculation 5-5 ii
GENE-A13-00395-01 TABLE OF CONTENTS (continued)
Section Iltle Eage 6
DETECT AND SUPPRESS RESULTS 6-1 6.1 Statistical Model Calculation 6-1 6.2 Final MCPR Calculation 6-4 7
RELOAD APPLICATION 7-1 8
REFERENCES 8-1 4
iii
GENE-A13-00395-01 i
1 LIST OF TABLES j
i Table
'Iille Eage 4-1.
Probe Points in Operating Map 4-2 i
i 4-2 Coordinates of Exclusion Region Boundary 4-3 I
6 Cooper Cycle 17 Inputs for Hot Bundle Oscillation Magnitude Calculation 6-1 6-2 Results of Cooper Cycle 17 Hot Bundle i
Oscillation Magnitude Calculation 6-2 i
7-1 Parameters for Reload Review Evaluation 7-1 4
}'
l i
i c
l i.
'N 4
f-i i
i i
iv
GENE-A13-00395-01 LIST OF FIGURES Eigutc
.Titic Eage 1-1 Typical Exclusion Region in Operating Map 1-3 2-1 Cooper Exclusion Region (Cycle 17) 2-1 3-1 Hot-channel Axial Power Profiles 3-2 3-2 Average-channel Axial Power Profiles 3-3 4-1 Probe Points in Operating Map 4-1 4-2 Coordinates of Probe Points in Stability Criterion Map 4-3 4-3 Cooper Exclusion Region (Cycle 17) 4-6 5-1 Fixed DIVOM Curve for Core-Wide Mode Oscillations 5-6 y
J i
GENE-A13-00395-01 ABSTRACT
]
1 This report demonstrates the application of the " Regional Exclusion with Flow-Biased APRM Neutron Flux Scram" Stability Solution (Option I-D) of the "BWR Owners' Group. Stability
[
Long-term Solutions Licensing Methodology" to the Cycle 17 as-loaded-core of the Cooper Nuclear Station, in compliance with General Design Criterion 12. An Exclusion Region is presented for the plant which identifies plant conditions that may lead to an instability. The Exclusion Region analysis shows that the Decay Ratios obtained creates a preference for core-wide mode oscillations should the plant maneuver into the conditions susceptible to a reactor
-instability. The Exclusion Region analysis concludes that regional mode oscillations are not anticipated to occur for Cooper. In addition, a statistically based Detect and Suppress analysis is performed to demonstrate protection of the fuel Minimum Critical Power Ratio (MCPR) Safety l
Limit from the flow-biased APRM neutron flux trip. The Detect and Suppress analysis is i
performed for core-wide mode oscillations only, censistent with the Detect and Suppress I
licensing methodology documented to the NRC in NEDO-32465, May 1995.
l l
1 Vi
GENE-A13-00395-01 1 INTRODUCTION This report demonstrates the application of the " Regional Exclusion with Flow-Biased APRM Neutron Flux Scram" Stability Solution (Option I-D) to the Cooper Nuclear Station as prescribed ti by the BWR Owners' Group Long-term Stability Solutions Licensing Methodology.21. This solution creates an " Exclusion Region" in the plant operating map wherein oscillatory power behavior is conservatively predicted to be possible and which is avoided during plant operations.
The Exclusion Region analysis also confirms that core-wide reactor instability is the predominate mode and regional mode oscillations are not expected to occur for Cooper. The protection of the Safety Limit Minimum Critical Power Ratio (SLMCPR) afforded by the flow-biased Average Power Range Monitor (APRM) neutron flux trip is demonstrated for the preferred core-wide mode of coupled thermal-hydraulic /neutronic oscillations for Cooper.
1.1 Historical Perspective Protection against power oscillations that might lead to fuel damage has been required by t31 General Design Criterion 12, which requires that such oscillations either not be possible or be reliably detected and suppressed.
In the past, this requirement was met by showing that oscillations are not possible by calculating core and channel decay ratios as a part of reload licensing analyses. Such results notwithstanding, guidance was provided to BWR operators as t41 early as 1982 in the form of a GE Service Information Letter on the detection and suppression of hypothetical power oscillations at low-flow and high-power conditions.
With the advent of 8X8 fuel designs and more aggressive operating strategies to improve operational flexibility and fuel utilization (e.g., extended load lines, feedwater heaters out-of-service, etc.), itability margins decreased such that instabilities could no longer be demonstrated to be impossible; therefore, in 1982 and after, protection against power oscillations was ensured by providing plant operators with guidance on detecting and suppressing such oscillations!U. In addition, analysis was performed to demonstrate that the occurrence of such oscillations did not challenge fuel thermal-mechanical limitstol, Additional concerns about BWR stability were raised by the March 9,1988, oscillation event at the LaSalle-2 plant, when investigations revealed that power oscillations could occur more rapidly than had been thought probable. Furthermore, new analyses predicted less margin to the t4 IM SLMCPR than was previously shown. This event led NRC to issue Bulletin 88-07, which requires BWR owners to indicate how they would guard against such events in the future.
1-1
GENE-A13-00395-01 1.2 BWR Owners' Group Response In response to NRC Bulletin 88-07, the BWR Owners' Group, in conjunction with GE, implemented a program to develop a long-term solution to the stability issue. The BWROG approach, as well as interim protective guidelines, was accepted by the NRC in Supplement I to
' the aforementioned Bulletin 'I.
The BWROG efforts led to generation of the "BWR Own rs' U
Group Long-term Stability Solutions Licensing Methodology"UI, which outlines several solution options. Some of these involve the introduction of a new Reactor Protection System (RPS) trip function and may be applied to all BWR's, while others demonstrate the adequacy of existing hardware but are applicable to only a limited set of plants.
1.3 ' Option I-D Solution t
One of the solutions which demonstrates the adequacy of existing hardware is Option I-D, entitled, " Regional Exclusion with Flow-Biased APRM Neutron Flux Scram." This solution consists of two parts. The first is the creation of an Exclusion Region in the operating map for the plant (Figure 1-1). This is a region where conservative decay ratio calculations indicate that power oscillations are possible. If the plant should enter this region due to a flow reduction event, such as a recirculation pump trip or runback, or due to a power increase at low flow, the operators are instructed to promptly exit the region and initiate a manual scram if oscillations occur. As a part of the generation of the Exclusion Region, the margin to regional mode DI oscillations is quantified using the methodology identified in Supplement I to NEDO-31960.
The second part of this solution is a demonstration that, even in the unlikely event of a power oscillation, an APRM flow-biased flux trip will detect and suppress the most probable mode power oscillations (core-wide made) before the SLMCPR is reached. This demonstration uses t21 the statistical methodology described in NEDO-32465. It is conservatively applied for core-wide mode oscillations both in terms of the inputs and confidence levels used in the statistical methodology.
While the Exclusion Region and MCPR analysis are the components of the Option I-D solution which are analytically demonstrated, they are not, in and of themselves, the complete solution.
Recognizing that highly skewed axial power shapes reduce margin to the onset reactor instability, an on-line stability predictor and administrative controls are being added to Cooper by the licensee. Therefore, the au
+al demonstrations are part of a hierarchy of barriers that provide a high degree of assurance that fuel thermal limits cannot be approached. The barriers that must be scaled before fuel limits can be approached may be summarized as:
1-2
GENE-A13-00395-01 Occurrence of a transient that brings the plant into the Exclusion Region (e.g., recirculation pump trip, recirculation pump runback, inadvertent control rod withdrawal or loss of feedwater heating during startup).
. Failure to leave the Exclusion Region either by increasing flow or decreasing power (It has been observed that an appreciable time lapse occurs before the system stabilizes at the new operating point and that oscillations require some time to evolve: there is adequate time for the operators to maneuver the plant out of the Exclusion Region or to scram the plant upon recognition of an oscillation.).
. Development of oscillatory power behavior outside of the expected statistical occurrence for which a RPS trip does not occur before fuel thermal limits are exceeded.
Figure 1-1. Typical Exclusion Region in Operating Map 120 7 100 1
,G" Anatytically Determined Region Boundary y
80 1
/
g 60 f Exclusion Region f
p k
f,/
N O
20 +
_j' o
0 20 40 60 80 100 120 Core Flow (% of Rated) 1-3 l
l
1 e
o GENE-A13-00395-01 1.4 Applicability of Option I-D to Cooper Integral to the Option I-D approach is the assertion that regional mode oscillations have a low probability of occurrence. Regional mode oscillations have not been observed for channel hydraulic decay ratios less than 0.6, and core wide mode may be assumed to be the predominant mode as long as channel decay ratio is below 0.56 when core decay ratio is greater than 0.8.
Results of this study show that the channel decay ratios are below 0.56 for all the cases except at the intersection of ELLLA and natural circulation line which is 0.65, but the core decay ratio is 1.46, which makes core wide mode oscillations more likely to occur at this point.
A second feature in the application of Option I-D is that Cooper has an unfiltered APRM Flow Biased Flux Scram instead of a simulated Thermal Power Monitor (STPM). The APRM neutron flux signal provides an instantaneous response to an oscillation rather than the slower fuel thermal response associated with a STPM. The assertion for Cooper is that (1) a core-wide mode oscillation will be excited long before an azimuthal (regional mode) oscillation, and (2) the APRM flow-biased flux trip will suppress the oscillations before a thermal limit is reached (the MCPR limit is the most sensitive thermal limit for oscillatictu).
1-4
?
GENE-A13-00395-01 2
SUMMARY
AND CONCLUSIONS Compliance with General Design Criterion 12 is demonstrated with tne Regional Exclusion with Flow-biased APRM Neutron Flux Scram Stability Solution (Option I-D) for Cycle 17 of the Cooper Nuclear Station.
The Exclusion Region ihr Cycle 17 of Cooper is shown in Figure 2-1. The analysis confirms that core-wide mode osciliations are the preferred mode for Cooper primarily due to the low channel decay ratios.
Figure 2-1. Cooper Exclusion Region (Cycle 17)
j h,
,/
(108/100) l
,/
100 APRM Rod Block
/
/
/
,/
,/
/
I 80
/
/
8
/,
N
/
Rated rod line ff 40
/
I Min pump speed y
v Natural 20 circulation l
i
/
l 0
0 20 40 60 80 100 120 l
l j
Core Flow (% of rated) l t
i 2-1 l
GENE-A13-00395-01 Protection of the SLMCPR is demonstrated for core-wide mode oscillations on the rated licensing procedure flow-control line in accordance with the statistical methodology defined in NEDO-32465[21 Therefore, the flow-biased APRM neutron fiux trip provides protection of the fuel SLMCPR against the preferred mode of oscillation with high statistical confidence for Cycle 17.
Results of this demonstration for Cycle 17 are expected to be applicable to future reload cycles due to the use of conservative inputs and assumptions. However, it is appropriate to confirm the applicability of the specific inputs and conditions identified in Section 7 for subsequent reload designs on a cycle-by-cycle basis.
h l
GENE-A13-00395-01 3 APPLICATION OF BWROG STABILITY LONG-TERM SOLUTION REGIONAL EXCLUSION METHODOLOGY Section 3 describes the application of the BWROG Regional Exclusion Methodology for Cooper.
This application is intended to define the power flow conditions to be avoided during normal operation. Also, the results of this analysis conservatively verify that the core-wide mode is the preferred mode for Cooper.
The analysis inputs described below for the demonstration application were developed for Cycle 17. Future operating cycle reload analysis will confirm the applicability of the power flow map Exclusion Region and preference for core-wide mode oscillations to the particular characteristics of the new fuel cycle.
The algorithm used to define the Exclusion Region is based on the FABLE /BYPSS methodology and the inputs to it are as described in Section 5.2 of the BWROG methodology repo#I. Input parameters that are dependent upon cycle specific parameters, such as fuel loading, are from Cycle 17 for Cooper. As such, the Exclusion Region is specific to Cycle 17 and its validity must be confirmed for each subsequent fuel reload.
3.1 Void CoefHelent Void-feedback parameters (nuclear void coefficient and delayed neutron data) are chosen from the exposure point in Cycle 17 for which void coefficient is most negative. Other inputs to the methodology (e.g., axial power distribution) are not from the same exposure point, but use of the most limiting void-feedback parameter values is conservative.
3.2 Thermal-hydraulie Data Standard design values for Cooper, consistent with the FABLE /BYPSS qualification bases, are used in the analysis.
3.3 Hot-Channel Axial Power Distribution Channel hydraulic stability is known to be strongly affected by the channel's axial power distribution. For the hot channels, the axial power distribution is fixed by the procedure to be peaked near the bottom of the channel, a distribution that is known to be less stable. These axial power distributions for both forced flow and natural circulation are shown in Figure 3-1. These axial profiles are consistent with those shown in Figure 5-5 of the BWROG methodology report 'l. Hot channels are identified for each hydraulic channel design in the Cooper core.
l 3-1
GENE-A13-00395-01 Figure 3-1 Hot-channel Axlal Power Profiles
{
2.00,
}
/N L80j 1.60 g
N,\\
'k, 1.40 1.20,
- '3g
$ 1.00 A
g 0.80 y I05:g
$ 0.60 IQ 0 40
'*Da 0.20
.g 7
g 0.00 '
1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 Axial Location
--s-Natural Orculation Forced Flow 3.4 Average-Channel Axial Power Distribution Core stability is known to be affected by the axial power distribution of th: bulk of the channels in the core (all those other than the " hot channels"). In the absence of other changes, a relatively
" flat" axial power distribution will be less stable than top-peaked or bottom-peaked distributions:
therefore, for forced circulation conditions, the Haling End-of-Cycle 17 (EOC-17) full power and flow core-average axial power distribution is used (see Figure 3-2). For natural circulation conditions, the power distribution moves strongly to the bottom of the core and use of a Haling profile characteristic of full power and flow would be too conservative; therefore, a core-average axial power distribution characteristic of natural circulation flow at the Haling EOC-17 exposure point is used. The axial power profile at the intersection of the rated flow-control line (FCL) and the natural circulation flow line is shown in Figure 3-2.
I 3-2
1 i
GENE-A13-00395-01
~I Figure 3-2 l
l Average channel Axial Power Profiles l
l l
l 1.40 I
, s._,
/
" ' = ~, _
.20 1.w.j
, *-+-+-+-+ ;=,=,_,=,==~s,_ -=
7 e
- ~.
j,-
s N
B "x
M 0.80. /
"N.\\.
m l
?
I N\\
g 0.60.-[
g\\
O 40 0.20 -
0.00 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 Axial Location l-s NaturalCaculaton + Forced Fbw l 3.5 Radial Power Distribution The radial peaking factors for the channel grouping used in the FABLE /BYPSS analyses are l
based on those obtained from the GE 3D BWR Simulator Code "3. The values chosen are from the EOC-17 Haling exposure point.
'3.6 Pellet-Clad Gap Conductance Core average pellet-clad gap conductances were determined for each fuel design using the approved fuel licensing model consistent with the FABLE /BYPSS qualification bases.
3.7 Miscellaneous Input Value; Other input values to the FABLE /BYPSS analyses, such as heat balance data, recirculation loop resistance, fuel physical parameters and material properties are standard design values for the 3-3
_ 7. _._.. _ _ _ _ _ _ _.. - --
GENE-A13-00395-01 i
l Cooper plant. It is assumed that the nominal heat balance assumptions, such as the operation of l.
all feedwater heaters, are valid for this model.
t t
l t
l l
i l
l l
l l
I 3-4
GENE-A13-00395-01 4 REGIONAL EXCLUSION RESULTS Core and channel decay ratios were calculated for several power flow combinations on the operating map (see Figure 4-1) using the inputs described in Section 3. The purpose of analyzing these combinations is to determine the Exclusion Region boundary on the power flow map and, using the generic BWROG Stability Criterion Map, establish the preferred mode of oscillation and the margin to the occurrence of regional mode oscillations for Cooper.
Figure 4-1. Probe Points on Operating Map i
\\g (108/100) 100 APRM Rod Block
/
/
{
80
/
/
f f / /%
/
Rated rod line 60
,- Min pump speed
+
l 40 g
Natural 20 circulation j
/,
0 0
20 40 60 80 100 120 l
Core Flow (% ofrated)
The points calculated are provided in Table 4-1. Points I through 4 are along the ELLLA Rod Line and Points 5 through 8 are along the natural circulation line. Points 1,2,4,7, and 8 are shown on Figure 4-1.
The core and channel decay ratio results of the analyzed points are tabulated in Table 4-1.
4-1
i GENE-A13-00395-01 Table 4-1 Probe Points on Operating Map Point Power Flow Channel Core Decay Symbol on Number
(%)
(%)
Hydraulic Ratio Figure 4-2 Decay Ratio 1
76.1 45.0 0.39 0.74 '
+
2 74.3 42.0 0.45 0.89
+
3 73.2 40.0 0.51 1.01
+
4 50.0 30.0 0.65 1.46
+
5 45.0 30.0 0.52 1.31 5
6 40.0 30.0 0.42 0.94 5
7 37.0 30.0 0.37 0.82 E
8 35.0 30.0 0.34 0.76 E
The points shown in Figure 4-1 and provided in Table 4-1 are plotted on the Stability Criterion Map in Figure 4-2. The plotting symbols have been provided in Table 4-1 for clarification. The lines which connect the appropriate state points in Figure 4-2 are used to determine the power and flow conditions at which the stability map criterion are exactly met. The coordinates of the intersections with the stability map criterion lines are given in Table 4-2. Point A is on the ELLLA rodline and point B is on the natural circulation line.
4-2
GENE-A13-00395-01 Figure 4-2 Coordinates of Probe Points on stability Criterion Map 1.2 1+
hrgin to Regonal 2de Oscillations
/
0.8 8
s' /
ct 06 e
e 04-0.2 0
0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1
Channel Hydraulic Decay Ratio l
Table 4-2 Coordinates of Exclusion Region Boundary Point #
Power (%)
Flow (%)
A 75.4 43.8 B
36.3 30.0 4-3
GENE-A13-00395-01 1
The coordinates of the probe points on the generic BWROG Stability Criterion Map, Figure 4-2, provide further evidence that regional mode oscillations are not probable for Cooper. It was lU shown in the stability solutions licensing methodology report that the probability of regional mode oscillations becomes progressively smaller as the channel hydraulic decay ratio is decreased. Regional mode oscillations have not been observed for channel hydraulic decay ratios less than 0.6, and core wide mode may be assumed to be the predominant mode as long as channel decay ratio is below 0.56 when core decay ratio is greater than 0.8. Results of this study show that the channel decay ratios are below 0.56 for all the cases except at the intersection of ELLLA and natural circulation line which is 0.65, but the core decay ratio is 1.46, which makes core wide moce oscillations more likely to occur at this point.
The points identified in Table 4-2 were then used to determine the location of the Exclusion Region boundary, which is shown in Figure 4-3. The Exclusion Region boundary far Cooper is specified by the boundary shape function equation which has been validated against previous Option I-D plant-specific region boundary calculations. The equation for the boundary is as follows:
~
1 W-W,
' W-W, '
P = Pg P ' 5 W,-W, $ W,-W,s A
s Pa >
- where, P = a core thermal power value on the Exclusion Region boundary (% of rated),
W = the core flo'y rate corresponding to power, P, on the Exclusion Region boundary
(% of rated),
P = Core thermal power at State Point A ("A e f rated),
A Pa = core thermal power at State Point B (% of rated),
W = core flow rate at State Point A (% of rated),
3 We = core flow rate at State Point B (% of rated),
The range of validity of the fit is:
4-4
GENE-A13-00395-01 30.0% < % Flow < 43.8%.
u Note that entry into the exclusion region above the ELLLA line is operation in a non-licensed part of the power-flow map.
t t
4C,
t d
4-5 4.
a
' o
'j; ;;
4 s.
o GENE-A13-00395-01 Figure 4-3. Cooper Exclusion Region (Cycle 17) 140
[
MRM sc=n une 120-
,/,, - (108/100) 100 APRM Rod Block f'
\\p
/
80 Rated rod line 60 t
40 Min pump speed ci75ldn 20
/
0 O
20 40 60 80 100 120 Core Flow (% ofrated) i 4-6
1 GENE-A13-00395-01 5 APPLICATION OF BWROG STABILITY LONG-TERM SOLUTION DETECT AND SUPPRESS METHODOLOGY 5.1 LICENSING COMPLIANCE Section 5 describes the application to Cooper Cycle 17 of the Detect and Suppress portion of stability long-term solution Option 1-D.
This application demonstrates protection of the SLMCPR provided by the flow-biased APRM neutron flux trip for core-wide mode oscillations.
The Detect and Suppress licensing methodology for application to Option 1-D is documented in BWROG Licensing Topical Report NEDO-32465[2j. Consistent with Cooper qualification as an Option I-D solution plant, the Regional Exclusion methodology demonstrates that core-wide is the predominate oscillation mode and, therefore, the Detect and Suppress calculation must only be performed for core-wide mode oscillations.
t2j The Detect and Suppress methodology assumes that a core-wide mode oscillation occurs, and is terminated by automatic reactor scram when the APRM oscillation magnitude reaches the flow-biased APRM flux trip. The methodology applies a statistical method, using a combination of statistical and deterministic inputs, to determine the final MCPR (FMCPR) with a high statistical confidence when control rod insertion disrupts the oscillation. The flow-biased APRM flux trip provides adequate protection as long as the FMCPR is greater than the SLMCPR.
5.2 METHODOLOGY OVERVIEW The Detect and Suppress methodology is used to determine the FMCPR resulting from core-wide mode oscillations which are terminated by the APRM flow-biased scram. The rated flow-control line (RLPFCL) is used to define the plant conditions for application of the methodology. The methodology consists of three major components:
Calculation of the Pre-Oscillation MCPR: A Cooper Cycle 17-specific determination a.
of the MCPR on the RLPFCL captures the margin to the SLMCPR prior to the oscillation. This is known as the initial MCPR (IMCPR). The IMCI'R is calculated conservatively assuming the plant is initially operating at the MCPR operating limit (OLMCPR).
- b. Statistical Calculation of Peak Oscillation Magnitude: A statistical evaluation of the normalized peak oscillation magnitude, Ah (defined as oscillation (peak-minimum)/ average), due to an oscillation initiating on the RLPFCL captures the effect of plant characteristics, trip system definition, and setpoint values on the peak fuel bundle power oscillation magnitude. The statistical methodology considers power distributions, oscillation contours, oscillation growth rates, oscillation 5-1
GENE-A13 00395-01 frequencies, trip overshoot, LPRM failures, and APRM failures. The result of the evaluation is a statistically conservative value of the peak hot bundle oscillation magnitude, Ah s5, at a 95% probability and 95% confidence level for anticipated 95 reactor instability.
c.
MCPR Performance of the Hot Bundle: A relationship between the fractional change in CPR and the hot bundle oscillation magnitude for core-wide mode oscillations captures the effect of fuel design. The relationship has been derived from 3-D TRACO analyses performed over a range of conditions and conservatively represents current Cooper loaded fuel designs.
The IMCPR and oscillation magnitude calculations are both evaluated at the RLPFCL.
Additional conservatisms have been added to the methodology to streamline the reload review process. A relatively simple confirmation of the applicability of each portion of the Cooper Cycle 17 Detect and Suppress calculation is all that will be required for subsequent fuel cycles to assure with a high confidence that the RPS trip setpoints cot.tinue to provide protection of the SLMCPR for anticipated reactor instability. If the applicability of a portion of the calculation cannot be assured, then specific portions of the calculation would need to be re-performed.
c Further information on application of each of the three portions of the methodology to Cooper Cycle 17 is provided in the following.
5.2.1 PRE-OSCILLATION MCPR The IMCPR is the more limiting (lower) of the MCPR from two scenarios on the RLPFCL. The two scenarios evaluated are (1) a two recirculation pump trip from rated flow with the MCPR at the OLMCPR, and (2) steady-state operation at 45% core flow at the applicable flow-dependent OLMCPR.
5.2.1.1 Two Recirculation Pump Trip For Cooper Cycle 17, the OLMCPR is 1.23 'I. Flow runback analysis completed on the 0
RLPFCL with the 3D core simulator determined that the CPR increase due to the flow runback from rated flow to natural circulation is 0.42 Therefore, the IMCPR for Condition 1 is:
IMCPR = 1.23 + 0.42 = 1.65 i
5-2 l
L
l l
GENE-A13-00395-01 5.2.1.2 Steady-State Operation at 45% Core Flow For Cooper Cycle 17, The OLMCPR on the RLPFCL at 45% flow is computed from the flow-dependent MCPR limits for Cooper Cycle 17D33 IMCPR = l 41 2
5.2.1.3 LimitingIMCPR The IMCPR is the more limiting (lower) ofIMCPR and IMCPR :
i 2
IMCPR = Min [IMCPR, IMCPR ] " l 41 i
2 5.2.2 STATISTICAL CALCULATION OF HOT BUNDLE OSCILLATION MAGNITUDE The statistical model is described in BWROG Licensing Topical Report NEDO-32Mst21, 73, model calculates hot bundle oscillation magnitude, Ah, dependent on a combination of statistical inputs and deterministic plant-specific factors. The statistical model results in selection of a conservative value of the hot bundle oscillation magnitude, Ah95M, at the 95% probability with a 95% confidence level.
5.2.2.1 StatisticalInputs Growth Rate: A review of actual instability events indicates that most BWR oscillations would be expected to have a growth rate only slightly above 1.00. For Cooper application, the growth rate is randomly selected from the probability density function 2
with a x distribution shown in Ref. 2.
Overshoot: The trip setpoint overshoot is a measure of how much an oscillation exceeds the trip setpoint. The overshoot is the fraction of the peak-to-peak difference between two consecutive cycles which is above the setpoint, when a trip occurs. Thus,0.0 s 6 s 1.0; and the value of 6 can be consUertti to be essentially random. For Cooper application, the overshoot is rando mly selected from the uniform distribution shown in Ref. 2.
Oscillation Period: The statistical methodology considers a range of oscillation periods.
Studies of actual instability events indicate that the expected value for the period is approximately 18 to 2.0 seconds. However,it is desirable to consider an oscillation frequency range between 0.7 Hz and 0.3 Hz. This corresponds to a desired period range of 1.4 sec $ T $ 3.3 sec. For Cooper application, the oscillation period is randomly selected from the probability density function with a x distribution shown in Ref. 2.
2 LPRM Failures: The statistical model provides options for considering an input LPRM failure probability distribution, a fixed failure percentage, or no LPRM failures in the calculation of hot bundle oscillation magnitude. For Cooper application, a random ntunber of LPRM failures are selected from the distribution specified in Ref. 2 which is 5-3 y
e I
GENE-A13-00395-01 representative of plant data on LPRM failure rates. The specific LPRMs which are defined to fail for a given trial are then randomly selected from the total Cooper LPRM population.
Oscillation Contours: The statistical model randomly selects from the specified set of oscillation contours. Cooper application uses plant-specific contours developed for core-wide mode oscillations.
5.2.2.2 Deterministic Inputs LPRM Assignments: Option 1-D relies on the APRM flow-biased trip. LPRMs are assigned to their respective APRM channels according to the plant configuration. All non-failed LPRM signals in an APRM are used to produce an averaged power signal for comparison to the trip setpoint. Cooper is designed with 124 LPRMs, in 6 APRM channels.
l Trip Setpoint: The nominal APRM trip setpoint is input as a percentage of rated power.
At natural circulation, the flow-biased APRM trip is at 62.0% of rated (2381 Mwt)
IW reactor power Radial Peaking Factor: Since only the fundamental mode from the 3-D BWR simulator is used to calculate the relative LPRM signal averages, A, there is only one hot bundle in the core-wide mode oscillation. This bundle is also the "true" hot bundle with the highest radial peaking factor. Its normalized oscillation magnitude, Ah, is the same as any other 121 location in the core. The radial peaking factor used for Cooper is 1.4753,
RPS Trip Logic: Cooper has a one-out-of-two, taken twice trip logic. Therefore, at least one channel from Division I and at least one channel from Division II must reach the APRM trip setpoint for the trip signal to be generated.
APRM Channel Failure: In addition to the failure ofindividual LPRMs, the failure of one APRM channel is considered. The model provides several options: no APRM channel failure, failure of a specified channel, failure of a randomly selected channel, and failure of the most responsive channel. For conservatism, the failure of the most responsive channel (i.e., the first channel to reach the trip setpoint) is used for Option 1-D analysis.
Delay Time: The delay time for control rod insertion to terminate oscillation growth is input to the model. The time at which the reactor trip criterion is reached plus the delay time sets the time window in which the pHeak hot bundle oscillation magnitude can occur l
The delay time is a plant-specific input consisting of the APRM response time (20 msec), the RPS processing time (50 msec), the control rod drive delay time before rod motion begins and the time for control rods to insert two (2) feet into the core assuming control rods insert at the minimum scram speed allowed by the plant Technical Specifications (784 msec). Even though control rod insertion two feet into the core will not shut the reactor down, it is judged to be adequate to prevent further growth of the hot bundle oscillation. Therefore, the total delay time for Cooper Cycle 17 is 854 msec.
5-4
GENE-A13-00395-01 5.2.3 MCPR PERFORMANCE OF THE HOT BUNDLE The relationship of change in CPR as a function of oscillation magnitude has been designated as the DIVOM curve (Delta CPR over Initial CPR Vs. Oscillation Magnitude). Application to Option I-D uses the generic DIVOM curve for core-wide mode oscillations ), which is the same t2 as the fixed DIVOM curve previously specified for Option 1-D application. The equation of the fixed curve is (ACPR/IMCPR = 0.175
- Ah ss + 0.05]. The specified fixed curve is shown in 95 Figure 5-1.
i a
The generic DIVOM curve for core-wide mode oscillations is reasonably conservative (but not t21 necessarily bounding in all cases) when compared to the TRACG CPR performance data. It is very conservative for application in the licensing methodology since using a nominal value for i
the slope of the generic DIVOM curve with the Ah ss hot bundle oscillation magnitude would 95 produce a FMCPR at approximately the 95/95 level.
5.3 FINAL MCPR CALCULATION The three-parts of the Detect and Suppress methodology provides for a conservative calculation of the minimum MCPR for an anticipated stability-related oscillation. First, the initial MCPR (IMCPR) is determined by a cycle-specific evaluation at the RLPFCL. Next, the hot bundle oscillation magnitude (Ah ss) is calculated at the RLPFCL.
Finally, the MCPR change 95 (ACPR/IMCPR) corresponding to Ah ss is determined. From these three elements, the final 95 MCPR (FMCPR) can be determined:
FMCPR = IMCPR - IMCPR * {ACPR/IMCPR}
where:
{ACPR/IMCPR} =
determined from generic DIVOM curve at the specified (P-M)/A 95ss oscillation magnitude.
The licensing criterion is met when the FMCPR is greater than the SLMCPR. For Cooper Cycle 17, the SLMCPR is 1.07p5),
5-5
l i
GENE-A13-00395-01 l
I Figure 5-1. Fixed DIVOM Curve for Core Wide Mode Oscillations j
l O.4
)
i i
i Fixed DIVOM Curve 1
f 0.3 I
,- l 1
)
,i O.2 i
s'
! Approxrnation to CLtve in Figure A-16 of NEDO-31960 l
j O.1
- s' I
1 o
0 0.5 1
1.5 2
2.5 Oscillation Magnitude, (P-M)/A i
5-6 I
GENE-A13-00395-01 6 DETECT AND SUPPRESS RESULTS 6.1 STATISTICAL MODEL CALCULATION The statistical methodology consists of a 1000-trial Monte Carlo analysis. Based on non-parametric tolerance limits, the methodology rank orders the 1000 trials and selects the 39* trial from the end as the m
95/95 value. A 1000-trial statistical analysis has been calculated for Coop:r Cycle 17. Table 6-1 lists the key inputs. Table 6-2 provides the highest 50 calculated values of hot bundle oscillation magnitude (Ah) and the 95/95 value (Ah ss = 0.847).
95 Table 6-1: Cooper Cycle 17 Inputs for Hot Bundle Oscillation Magnitude Calculation Core Size:
548-bundle core Trip System:
Flow biased APRM Trip Logic:
One-out-of two, taken twice Oscillation Mode:
Core-wide APRM Channel Failure:
Most Responsive APRM Channel (applied to 100.0% of all trials)
LPRM Failures:
Random (x* Distribution)
Oscillation Period, T:
Random (x* Distribution)
Growth Rate, Gr:
Random (X* Distribution)
Overshoot,5:
Random (Uniform Distribution)
Average Reactor Power:
46.0 % rated (100% rod line at natural circulation)
Radial Peaking Factor:
1.4753 APRM Trip Level (nominal): 62.0 % rated (at natural circulation)
Total Delay Time:
854 msec. (measured from time of full trip)
TotalNumber of LPRMs:
124 Oscillation Contour Selection: Random from contours: HTIB12AH1, HTlB12AH2, HT2M09AHl HT2M09AH2, HTlH12AHl HTlH12AH2 6-1
- ~.
- - - - ~.-
GENE-A13-00395-01 Table 6-2: Results of Cooper Cycle 17 Hot Bundle Oscillation Magnitude Calculation Contour Period APRM # LPRM %LPRM Hot Bundle Peak Power Trial #
ID (sec)
GR Overshoot Failure Failures Failures (P M)/A
(% rated) 961 HT2M09AHI 2.10 1.21 0.61 F-11 8.9 0.838 101.49 20 HT2M09AH2 1.88 1.15 0.85 A
26 21.0 0.838 101.51 542 HTIB12AH2 1.67 1.20 0.67 E
16 12.9 0.838 101.51
)
742 HT2M09AH2 1.73 1.17 0.73 F
4 3.2 0.841 101.62 t
45 HT2M09AH2 1.97 1.39 0.35 D
10 8.1 0.842 101.61
[
579 HTlH12AH1 2.09 1.13 0.89 F
9 7.3 0.842 101.66 92 HTlH12AH1 2.02 1.21 0.57 E
16 12.9 0.842 101.66 231 HTlH12AH1 2.52 1.16 0.76 F
5 4.0 0.843 101.72 i
~798' HT2M09AH2 1.48 1.22 0.61 C
8 6.5 0.844 101.78 488 HT2M09AHI 2.02 1.16 0.79 A
17 13.7 0.845 101.79 j
483 HT2M09AH2 2.57 1.14 0.93 F
8 6.5 0.845 101.83
^
179 HTlH12AH1 2.00 1.14 0.89 A
16 12.9 0.847 101.92 39TH TRIAL FROM END: HOT BUNDLE (P-M)/A AT 95/95 LEVEL = 0.847 225 HT2M09AH1 2.81 1.18 0.78 E
12 9.7 0.847 101.92 669 HT2M09AH1 2.21 1.15 0.94 D
7 5.6 0.850 102.02 267 HT2M09AH2 1.89 1.23 0.61 F
5 4.0 0.850 102.04 548 HTlH12AH1 1.73 1.14 0.94 F
11 8.9 0.852 102.16 694 HTlH12AH1 1.75 1.16 0.87 F
6 4.8 0.852 102.15 447 HTlH12AHI 1.92 1.28 0.49 D
8 6.5 0.852 102.14 812 HT2M09AH1 2.49 1.22 0.64 D
11 8.9 0.854 102.22 93 HT2M09AH2 2.40 1.15 0.95 F
7 5.6 0.856 102.33 116 HT2M09AH1 1.97 1.26 0.60 E
10 8.1 0.857 102.36 356 HT2M09AH1 1.45 1.16 0.91 A
7 5.6 0.858 102.40 678 HTlH12AH2 2.65 1.33 0.47 D
29 23.4 OA60 102.47 501 HTlH12AH2 2.20 1.19 0.74 F
3 2.4 0.860 102.49 681 HT2M09AH1 1.47 1.37 0.46 F
9 7.3 0.864 102.65 242 HTlHl2AH1 2.15 1.17 0.91 F
12 9.7 0.865 102.71 656 HTlH12AH2 1.90 1.21 0.74 F
6 4.8 0.866 102.75 733 HT2M09AH2 1.94 1.22 0.72 F
4 3.2 0.866 102.74 t
200 HTlH12AH1 1,91 1.18 0.83 F
6 4.8 0.866 102.77 556 HTIB12AH2 2.52 1.18 0.96 F
11 8.9 0.868 102.86 648 HTlH12AH2 1,71 1.17 0.93 F
3 2.4 0.868 102.87 538 HT2M09AH2 1.68 1.17 0.93 F
16 12.9 - 0.868 102.87 25 HT2M09AH2 1.%
1.19 0.86 F
6 4.8 0.868 102.88 605 HTlHl2AH1 2.40 1.24 0.68 A
7 5.6 0.871 102.98 398 HT2M09AH2 2.03 1.21 0.79 F
12 9.7 0.872 103.04 618 HTlH12AH1 1.47 1.24 0.67 E
9 7.3 0.873 103.06 105 HTlH12AHI 2.02 1.19 0.85 F
14 11.3 0.873 103.08 785 HT2M09AH2 3.50 1.21 0.81 D
13 10.5 0.873 103.09 978 HT2M09AH2 1.50 1.21 0.82 C
15 12.1 0.874 103.13 983 HT2M09AH1 2.34 1.22 0.83 F
13 10.5 0.880 103.39 819 HTlH12AH1 1.91 1.22 0.80 B
22 17.7 0.885 103.66 256 HTIB12AH2 2.08 1.29 0.73 C
5 4.0 0.891 103.93 864 HT2M09AH1 2.20 1.27 0.73 F
16 12.9 0.894 104.04 262 HTIB12AH2 2.45 1.21 0.96 F
11 8.9 0.896 104.13 773 HTlH12AH2 1.88 1.22 0.89 D
23 18.5 0.901 104.37 440 HTlH12AH2 2.00 1.30 0.75 F
5 4.0 0.920 105.26 748 HT2M09AH2 1.91 1.27 0.90 D
13 10.5 0.925 105.48 68 HTlH12AHI 2.08 1.44 0.62 D
21 16.9 0.%3 107.25 936 HT2M09AH2 1.69 1.36 0.88 D
13 10.5 0.991 108.58 551 HT1H12AHI 1.91 1.41 0.80 F
5 4.0 1.002 109.09 6-2
GENE-A13-00195-01 Period
%LPRM Hot Bundle Peak Power (sec)
GR Overshoot Failures (P-M)/A (% rated)
MINIMUM 1.27 1.00 0.00 0.8 0.723 96.35 MAXIMUM 3.67 1.44 1.00 27.4 1.002 109.09 AVERAGE 2.04 1.10 0.49 8.6 0.776 98.69 l
1 1
1 l
i i
l I
J 6-3
e GENE-A13-00395-01 6.2 FINAL MCPR CALCULATION First, the initial MCPR (IMCPR) is determined by a cycle specific evaluation at the RLPFCL.
IMCPR = 1.41 Next, the hot bundle oscillation magnitude (Ah95SS) is Calculated using the statistical methodology at the RLPFCL.
Ah9333 = 0.847 Finally, the MCPR change (ACPR/IMCPR) corresponding to Ah a5 is determined from the 95 generic DIVOM curve for core-wide mode oscillations.
ACPR/IMCPR = 0.175
- Ah ss5 + 0.05 9
ACPR/IMCPR = 0.175 * (0.847) + 0.05 = 0.198 From these three elements, the final MCPR (FMCPR) can be determined:
FMCPR = lMCPR - IMCPR * {ACPR/IMCPR)
FMCPR = 1,41 - 1.41
- 0.198 = 1.13 The licensing criterion is met when the FMCPR is greater than the SLMCPR.
FMCPR > SLMCPR 1.13 > 1.07 Since the FMCPR is greater than the SLMCPR, the APRM flow-biased trip system shows protection for core-wide mode oscillations.
6-4
~ - -
GENE-A13-00395-01 i
7 RELOAD APPLICATION The purpose of the reload review is to determine the applicability of previous plant-specific calculations to the current fuel cycle. The analysis documented in this report constitutes the baseline for future fuel cycle reload reviews. Table 7-1 tabulates the key parameters which must
)
be evaluated to determine the applicability of the analysis documented herein. If some key i
parameters do not meet the specified criteria, the applicable portions of the analysis must be re-j performed.
I Table 7-1. Parameters for Reload Review Evaluation Regional Exclusion Methodology i
Description Criteria Base Value There are no reactor design changes which No reactor changes would affect the thermal-hydraulic stability of the reactor (e.g., recirculation loop performance)
There are no new plant operating modes (e.g.,
No operating region j
power uprated, increased load lines) which change i
would affect the operating region of the reactor The reload fuel design has similar stability Similar to GE9 GE9 i
performance as the Cycle 17 fuel design l
Haling radial peaking factor increases over s 105% of base value 1.47 Cycle 17 by no more than 5%
i Reload batch size changes by no more than 27 Within 27 bundles iS2 bundles l
bundles (5% of core size) from the Cycle 17 from base value j
batch size The Haling cycle exposure changes by no more Withini943 9426 MWD /ST i
than 943 MWD /ST (10% of base value) from MWD /ST from base the Cycle 17 Haling cycle exposure value The actual cycle exposure of the previous cycle Within 932 9317 MWD /ST changes by no more than 932 MWD /ST (10%
MWD /ST from base 4
of base value) from the Cycle ifi actual cycle value exposure 1
7-1
r j
GENE-A13-00395-01 Table 7-1. Parameters for Reload Review Evaluation (continued)
Detect and Suppress Methodology Parameter Description Value OLMCPR (100/100)
MCPR Operating Limit at rated flow on the 2 1.23 I
rated licensing procedure flow-control line i
AMCPR(2RPT)
MCPR increase due to flow runback from a 2 0.3874 l
2RPT l
OLMCPR (100/45)
MCPR Operating Limit at 45% of rated flow on 2 1.41 5
j the rated licensing procedure flow-control line
- LPRMs Number ofinstalled LPRMs 2 124 j
APRM assignment LPRM assignment to APRMs in 6 channels, No APRM design etc.
change APRM trip @ NC Flow-biased APRM trip power level (nominal s 62.0% of rated value) at natural circulation (2381 Mwt) power A @ NC Average power level on the rated licensing 2 46.0% rated power procedure flow-control line at natural circulation To,i,y Total delay time (20 msec APRM response s 854 msec time,50 msec RPS processing time,784 msec delay before start of control rod motion and for 2 feet of control rod insertion)
Fuel Design Fuel Design which is covered by the Generic GE7, GE8, GE9, DIVOM Curve for Core-Wide Mode GE10, gel 1, or Oscillations GE12 7-2
i 4
l GENE-A13-00395-01 l
8 REFERENCES
- 1. "BWR Owners' Group Long-Term Stability Solutions Licensing Methodology," NEDO-31960, June 1991, and NEDO-31960 Supplement 1 March 1992.
- 2. "BWR Owners' Group Reactor Stability Detect and Suppress Solutions Licensing Basis Methodology and Reload Applications," NEDO-32465, May 1995.
- 6. NEDE-22277-P, " Compliance of the General Electric Boiling Water Reactor Fuel Designs to Stability Licensing Criteria," December,1982.
i j
- 7. NEDE-22277-P-1, " Compliance of the General Electric Boiling Water Reactor Fuel Designs l
to Stability Licensing Criteda," October,1984 f
- 8. NEDO-31708, " Fuel Thermal Margin during Core Thermal Hydraulic Oscillations in a Boiling Water Reactor," June,1989.
- 9. NRC Bulletin No. 88-07, " Power Oscillations in Boiling Water Reactors," June 15,1988.
- 10. NRC Bulletin No. 88-07, Supplement 1, " Power Oscillations in Boiling Water Reactors,"
December 30,1988.
1
- 11. NEDO-30130-A, " Steady State Nuclear Methods," April,1985.
l l
l
- 12. NEDE-31917P, " gel 1 Compliance with Amendment 22 of NEDE-24011-P-A (GESTAR l
l II)," April,1991.
- 13. Core Operating Limits Report, Flow Dependent MCPR Limits for Cycle 17.
l l
- 14. NEDC-31892P, Rev.1, " Extended Load Line Limit and ARTS Improvement Program l
Analysis for Cooper Nuclear Station, Cycle 14, May 1991.
- 15. Customer Communication.
8 I
8-1 l
!