ML20206Q170

Rev 0 to Steady-State & Quasi-Steady-State Methods Used in Analysis of Accident & Transients

ML20206Q170
Person / Time
Site:	Oyster Creek
Issue date:	02/18/1987
From:	Bujtas E, Fu H, Furia R GENERAL PUBLIC UTILITIES CORP.
To:
Shared Package
ML20206Q111	List: ML20206Q106 ML20206Q125 ML20206Q170
References
TR-040, TR-040-R00, TR-40, TR-40-R, NUDOCS 8704210356
Download: ML20206Q170 (26)

Text

TR-040 Rev. O l

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I STEADY-STATE AND I QUASI-STEADY-STATE METHODS USED IN THE ANALYSIS OF ACCIDENT AND TRANSIENTS I TR-040 REV. O I

I BA N0.: 335430 E. R. BUJTAS I H. FU R. V. FURIA AUTHORS C. B. MEHTA DATE: February 18, 1987 APPROVALS:

0 d? -/9-87 muclear Analfsts and Fuels Director Date I

GPU NUCLEAR l 100 INTERPACE PARKWAY PARSIPPANY, NEW JERSEY 07054

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TR-040 Rev. 0 ABSTRACT Steady-state and quast-steady-state methods are used in the analysis of certain accidents and transients. The events include the fuel assembly misortentation, fuel assembly mislocation, control rod withdrawal error and the loss of feedwater heating. This report oliscusses the bases and procedures used by GPUN to analyze these events. Results of these analyses are presented for the Oyster Creek Cycle 10 reload.

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TABLE OF CONTENTS  !

SECTION Page 1.0 Introduction and Sumary 6 j 2.0 Fuel Assembly Misloading Analyses 8 I

2.1 Fuel Assembly Misortentation Error 8 2.2 Fuel Assembly Mislocation Error 11 3.0 Control Rod Withdrawal Error 17 4.0 Loss of Feedwater Heating 24 5.0 References 26 TOTAL PAGES 2r l l

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LIST OF TABLES i

TABLE TITLE PAGE 2-1 CPR for a Misortentated Fuel Assembly 14 1

I 23 l 3-1 aCPR and MLLHGR for Control Rod H1thdrawal Error (Maximum allowed failure combination)

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I LIST OF FIGURES I

I FIGURE TITLE PAGE 2-1 Four Bundle Control Cell with Misorientated Fuel 15 As'sembly (180* Misorientation) l 2-2 Axtal View of Misortentated Fuel Assembly 16 3-1 APRM Response to Control Rod Withdrawal 20 3-2 Mininum CPR to Control Rod Withdrawal 21 I

3-3 Control Rod Location for Worst APRM Response 22

! to a Control Rod Hithdrawal I

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1.0 INTRODUCTION

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SUMMARY

This report describes the steady-state and quast-steady-state methods used by GPUN in the analysis of certain transients and accidents. These methods are used because they: (1) provide more core detail than is available in the core model of the system transient model; (2) provide a i conservative alternative to the system transient model; or (3) the nature of the transient or accident is a steady-state event.

The transtents and accidents that are covered in this report are the fuel assembly misorientation, fuel assembly mislocation, control rod withdrawal error and the loss of feedwater heating. Except for the fuel I bundle misortentation, these events are analyzed with the 3-0 core  ;

simulator, N00E-B" ' . The fuel assembly misorientation is analyzed using the lattice physics code CPM"'. l I The parameters of interest for these events are the critical power ratto i (CPR) and the local linear heat generation rate (LLHGR). The safety limit MCPR is determined using the General Electric Company Thermal Analysis Basis, GETAB"' with the GE critical quality (X) bolling length (L), GEXL, correlation.

Each transtent and accident is discussed in a separate section of this report. A brief description of the event and the analysis is provided l I along with the procedure for determining the limiting parameters.

l Results of these analyses are presented for the Oyster Creek Cycle 10 reload.

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TR-040 Rev. O The verification of these methods !s difficult since test data 1 not generally available. As such the nature of Jhese analyses is to use a I very conservative approach for each ana'nysts to account for i

uncertainties. This is done by combining conservative assumptions for I the time of occurrence, reactor cc.'ditions ard in the computer code modeling. Finally, no credit is taken for procedures that are designed to prevent the occurrence of such events.

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2.0 FUEL ASSEMBLY MISLOADING ANALYSIS Two separate events are analyzed as part of the fuel assembly misloading analysis. The first event is the fuel bundle misortentation error where a fuel assembly is loaded in its correct core location but misortentated I (rotated) from its correct orientation. The second 9 vent, the fuel assembly mislocation error, is where a fuel assembly is loaded in the incorrect location. For both events, it is postulatec that the misloading error goes undetected and the reactor is operated with the fuel in the misloaded position for the entire cycle. l i

l The limiting parameter of interest for these events is the critical power ,

ratio. The fuel assembly misorientation analysis calculates a ACPR between the correctly and incorrectly orientated assembly. The fuel bundle mislocation analysis calculates the ACPR between the CPR for the l mislocated fuel assembly and the MCPR for the assembly in a symmetric location that is correctly loaded. The procedures for analyzing these events are designed to find the highest ACPR that would result from a fuel misloading of a limiting (high power) assembly.

l 2.1 Fuel Assembly Misorientation Error The fuel assembly misortentation error is defined as an event in which a fuel assembly loaded in its correct location, but is rotated 90' or 180' from its proper orientation. Oyster Creek is a 0 lattice plant having non-uniform water gaps between the fuel assemblies. The fuel pin enrichment distribution accounts for the variation in the water gap size by having lower enriched pins next l

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i I TR-040 Rev. 0 I-to the wider water gap and higher enriched pins next to the narrow water gap. Thus, a rotated assembly will result in having fuel pins normally adjacent to the narrow water gap adjacent to the wide water gap. The reverse is true for the pins normally adjacent to the wide l

water gap.

I The misorientaiton results in an increase in local pin power peaking and an increase in assembly power. A mitigating effect of the misorientation is the channel spacer pads, normally facing the wide water gap, are against the upper core support grid (Figure 2-1).

This causes the fuel assembly to sit in a tilted position varying I the size of the water gap along the axial length (Figure 2-2). The local power peaking along the wide water gap is reduced as the wide water gap narrows.

I The fuel assembly inisorientation analysis uses the CPM code. The l CPM code performs 2-0 lattice physics calculations and the code is used to generate lattice physics constants for Oyster Creek reload I methods. The code is limited to analyzing a single fuel lattice having diagonal symmetry and thus c&n only be used for the 180*

misortentation. Four bundle PDQ calculations have shown the 180*

misorientation to be mora severe than the 90* misorientation. A single assembly analysis is used since it is assu e d that the correctly orientated assembly would be on limits, thereby maximizing the consequences of the event.

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TR-040 Rev. O The procedure for calculating the ACPR due to a fuel assembly misorientation error is as follows.

I a. CPM calculations of a fuel design are performed for both the correct and 180* rotated position. These runs are done at different assembly exposures to determine the time in assembly burnup which yields the largest change in peaking.

I b. At the assembly exposure determined in (a), the axial average R factor is calculated for the coerect and rotated positions. The change in axial average R factor due to rotation is calculated.

Effects of tilting are included in the calculation of the ax,lal average R factor,

c. A CPR calculation is performed with assembly power, flow, peaking factors and maximum R factor for a correctly orientated bundle to be at the CPR operating limit. This is the initial CPR.(ICPR) for the transient.

d. The CPR calculation is now performed for the misorientated bundle applying the change in R determined in (b) to the bundle maximum R factor and increasing bundle power by 3.2%, while holding all other parameters constant. This is the MCPR for the assembly being analyzed. The 3.2% increase in power is an upper bounding value for increased power in a misorientated bundle determined in sensitivity studies.

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e. The ACPR for the fuel design analyzed is the change in ICPR and MCPR, steps (c) and (d).

I f. The fuel design having the largest ACPR is the ACPR fo, this transient.

I The ACPR for Oyster Creek Cycle 10 fuel designs are shown.in Table 2-1, along with the ICPR and MCPR for each design.

2.2 Fuel Assembly Mislocation Error The fuel assembly mislocation error is when a fuel assembly is loaded in an incorrect core location. The fuel assembly mislocation can result in a significant change in local bundle power between the region of the misloca.ted assembly and a correctly loaded symmetrical region. The limiting case is having a high reactivity bundle in place of a low reactivity bundle. Sensitivity studies have shown loading new fuel assemblies in place of highly exposed fuel assemblies will produce the largest reactivity mismatch and the potential for the largest power mismatch.

I The GPUN 3-D core simulator code, NODE-8, is used to analyze this event. The procedure for analyzing this event is as follows.

a. Each control cell (the four fuel assemblies that surround a control rod) in the core, except for those on the periphery, is reviewed tg determine the candidate cells for mislocation. A candidate control cell is determined as follows.

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I TR-040 Rev. 0 (1) The cell average exposure is calculated using the average exposure for each of the four fuel assemblies in each of the control cells.

I (11) The highest exposure fuel assembly is removed from each of the above cells and the cell average exposure is recalculated using zero exposure for the assembly removed. The delta exposure is calculated between Step (1) and this step.

(111) The cells having a combination of higher A exposure (11) and the higher cell average exposure (1) are the l potential candidate control cells.

(iv) A minimum of ntno cells are analyzed for determining the ACPR.

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b. A candidate control cell has the highest exposed fuel assembly replaced by a fresh fuel assembly. The core is burned in 1.0 GWD/MT intervals to EOC using projected control rod patterns for full power operation. The change in CPR between the mislocated I assembly and the minimum CPR in a symmetric cell is determined at each burnup step. Each ACPR is adjusted as if the assembly had been at its CPR operating limit. The process is repeated '

for each candidate cell.

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c. The cell having the maximum ACPR is the ACPR for this transient.

I The ACPR for this event using the above procedure is 0.29 for the O.C.

Cycle 10 core. In this particular example a fresh GE 2.65H fuel design replaced a fuel design having an exposure of 18.! GWD/HT. Minimum CPR occurred at a cycle exposure of 5.0 GWO/HT.

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I I TABLE 2-1 CPR FOR A MISORIENTATED FUEL ASSEMBLY I FUEL ASSEMBLY ICPR MCPR ACPR I ENC Type VB 1.400 1.172 0.228*

I GE P8DRB239 1.400 1.220 0.180 GE P8DRB265H 1.400 1.239 0.161 I

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• The ACPR for this transient.

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I FOUR BUNDLE CONTROL CELL WITH MISORIENTATED FUEL ASSEMBLY

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(1800 MISORIENTATION) j I l l

I Channel Misorientated Assembly U

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I TR-040 Rw. O Figure 2-2 I AXIAL VIEW OF MISORIENTATED FUEL ASSEMBLY I Normal Fuel Misorientated (Orientation { Fuel Assembly g Channel Spacer 5 8attaa p:

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3.0 CONTROL R00 WITH0RAWAL ERROR The control rod withdrawal error (RHE) event is initiated by continuously withdrawing a control rod at its maximum withdrawal rate. During the event the core average power increases with the control rod withdrawal.

The withdrawal is stopped by the rod block signal from the average power range monitoring (APRM) system. The limiting parameters of interest for this transient are the CPR and fuel design local linear heat generation rate (LLHGR). The ACPR is determined from the initial CPR and the CPR at the time the rod block is initiated.

I The APRM response, and hence rod block effectiveness, versus transient I rod position will vary based upon the number of available LPRMs feeding the APRM (Figure 3-1). Certain LPRM detectors in each APRM channel may be bypassed or failed without making the APRM channel inoperable. The MCPR for the RHE is determined (Figure 3-2) for each of the 3 status conditions for the APRM system, which are: status 3, no LPRM failures and status I and 2, allowable combinations of APRM bypass and LPRM failure. (See Oyster Creek Technical Specification section 3.10-C).

I The most 11miting case is a control rod withdrawal error at full power.

To maximize the worth of the control rod, the analysis is done with a Xenon free core at peak cycle reactivity, and the control rod of interest is fully inserted. The control rod pattern is adjusted to put a fuel assembly near the control rod to be withdrawn on thermal limits.

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TR-040 Rev. O GPUN performs two withdrawal error analyses to determine which control rod will give the largest change in thermal limits. The first analysis looks at the highest worth control rod. The second analysis uses a core location (Figure 3-3) that will result in a poor response of the APRM system with the maximum permissible combination of failed LPRMs and bypassed APRMs (Status 1) and also has a high worth control rod.

The control rod withdrawal error is analyzed as a series of steady-state calculations. Use of the quasi-steady-state approach is justified on the basis that the reactivity insertion rate is relatively slow and the core has time to equilibriate. The 3-0 core simulator code N00E-B is used to analyze this event.

l The procedure for analyzing the control rod withdrawal error is as follows:

a. Identify the highest worth control rod for the core loading at peak cycle reactivity, I

b. Set control rod pattern with control rod identified in (a) fully inserted and a nearby fuel assembly on thermal limits.

I c. Determine power level for each step of control rod withdrawal.

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d. Calculate the CPR, LLHGR, APRM and LPRM responses for each control rod step using power levels determined in (c).

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e. Determine rod block point for each APRM status and resulting ACPR and LLHGR. The ACPR is adjusted to reflect the ICPR as if it were on the operating limit for CPR. The maximum LLHGR (MLLHGR) is calculated assuming the fuel assembly had been at its operating LLHGR limit.

I f. Repeat steps b to e for control rod located in core location 42-31. j I g. The highest ACPR and LLHGP. for each APRM status condition between the two analyses is the ACPR and MLLHGR for this transient.

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en au m m em m as e e en as e e e as ee em FIGURE 3-1 APRM RESONSE TO CONTFDL FDD WITHDRANAL APRM READING 1.16 1.14 -

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TR-040 Rev. O TABLE 3-1 ACPR AND MLLHGR FOR CONTROL R00 WITHDRAWAL ERROR

( (MAXIMUM ALLOWED FAILURE COMBINATION) l APRM POWER MLLHGR ROD WITHDRAHAL (PERCENT) ACPR (KW/FT) (FT) 100.0 0.0 13.4 0.0 101.0 0.04 14.1 1.5 102.0 0.11 14.4 2.5 103.0 0.14 14.5 3.0 104.0 0.23 14.6 4.0 105.0 0.31 14.9 5.5

( 106.0* 0.33 14.9 6.0 107.0 0.33 14.9 6.0 108.0 0.35 15.0 6.5

• APRM R00 BLOCK t

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I TR-040 Rev. 0 4.0 LOSS OF FEEDWATER HEATING The loss of feedwater heating results in a core power increase and power b distribution shift d w to an increase in core inlet subcooling.

Feedwater heating can be lost when the stean extraction line to a feedwater heater is closed. The maximum number of feedwater heaters which can be tripped by a single event represents the most severe transient for analysis considerations. For 0.C. this represents a 100*F decrease in feedwater inlet temperature.

The limiting parameters of interest in this transient is the CPR and LLHGR. The ACPR is determined from the initial CPR and MCPR occurring during the transient.

The limiting case for this event is the power level increasing to the scram point but not initiating a scram. This allows the heat flux to equilibriate with the neutron flux at the increased power level. Since the loss of feedwater heating is a relatively slow event and no scram occurs, it is analyzed using the 3-D simulator model, NODE-B. It requires two steady-state calculations, one at the initial core conditions and the second at the increased power level.

The procedure for this analysis is to set up a N00E-B base case at full power conditions. A dependent case follows with the power level increased to 2233.0 MWth (scram point), feedwater temperature is decreased by 100*F and feedwater fleeir increased to run out. Xenon F

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Rev. 0 levels are maintained at full power level for increased peaking. The ACPR is calculated between the base case and the dependent case. The ACPR is then adjusted to reflect the ICPR as if it were at the operating limit for CPR. The maximum LLHGR is calculated assuming the fuel assembly had been at its operating LLHGR limit at the start of the transient.

The maximum ACPR for this transient during 0.C. Cycle 10 is 0.13 occurring at the cycle energy of 5.0 GWD/MT. The maximum LLHGR is 17.11 KW/ft, which is well below the transient limits for this event.

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5.0 REFERENCES

c t 1. " Methods for the Analysis of Bolling Water Reactors Steady State Physics,"

GPUN TR-021, Rev. O, R. V. Furia, January 31, 1986.

2. " Methods for the Analysis of Bolling Water Reactors Lattice Physics," GPUN

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TR-020, Rev. O, H. Fu, R. V. Furla, July 25, 1985.

3. " General Electric BWR Thermal Analysis Basis (GETAB) Data, Correlation and i Design Application," January, 1977, (NEDE-10958-PA and NEDO-10958-A).

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