ML20039A597
| ML20039A597 | |
| Person / Time | |
|---|---|
| Site: | Fort Saint Vrain  |
| Issue date: | 12/01/1981 |
| From: | Lee O PUBLIC SERVICE CO. OF COLORADO |
| To: | Novak T Office of Nuclear Reactor Regulation |
| References | |
| FSV-313, P-81303, NUDOCS 8112180493 | |
| Download: ML20039A597 (90) | |
Text
,
PUBLIC SERVICF
)MPANY OF COLORADO NVER.
COLORADO 80201 P.
O.80X 840
+
December 1,1981
- osc, Fort St. Vrain in g
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Unit No. I a
P-81303 4
gfREct.ygg fsv-313 n
5-\\ neNcDEC.z 7
L 4
Mr. Thomas M. Novak kft= S Assistant Director for Operating Reactors N
N Division of Licensing U. S. Nuclear Regulatory Commission Washington, D.C.
20555
Subject:
FSV RECA Verification
Reference:
PSC letter from Mr F.E. Swart to Mr. J.P. Miller, P-81083, dtd 2/5/81
Dear Mr. Novak:
Amendment 22 to the Fort St. Vrain Facility Operating License required Public Service Company to perform a RECA3 code verification analysis of plant transient response with region constraint devices installed prior to operating the reactor above 70% power.
A draft of the results of this analysis, "RECA3 Code Verification Analysis with Region Constraint Devices Installed," was submitted to the NRC via the above referer;ed letter. At a meeting held on February 10, 1981 between the NRC, PSC and General - Atomic Company, the RECA3 code verification results were discussed.
l The final report of the RECA3 analysis effort required by Amendment 22 is enclosed for your information and records.
Very truly yours,
$llbv
- 0. R. Lee, Vice President Electric Production ORL/JCS:pa Enclosures 6\\
8112180493 811201' PDR ADOCK 05000267 P
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FORT ST. VRAIN ACCIDENT REANALYSIS CODE (RECA3)
VERIFICATION WITH REGION CONSTRAINT DEVICES INSTALLED
References:
(1) P-78138, Response to NRC Question 222.001 dated August 11, 1978.
(2) Safety Evaluation Report by the Office of Nuclear Reactor Regulation Supporting Amendment 22. to
- Facility Operating License No. DPR-34 of Public Service Company of Colorado, Fort St. Vrain Nuclear Generating Station, Docket No. 50-267, dated August 19, 1980.
The Reactor Emergency Cooling Analysis (RECA3) code has been utilized in the teanalysis of several of the bounding accident cases postulated in the Fort St. Vrain Final Safety Analysis Report (FSAR).
In support of the use of RECA3 and in response to NRC questions, certain verification infor-mation has been submitted to the Commission (Reference 1). As a part of this code verification package, comparisons were made between measured and calculated core refueling region outlet. temperatures under four different sets of plant scram conditions.
In the Safety Evaluation Report supporting Amendment 22 to the Fort St. Vr,ain Operating License (Reference 2), the Staff expressed concern over the dis-r crepancy between some predicted and measured core region outlet temperatures presented in the comparison of RECA3 predictions and scram data. These dis-crepancies occurred in the seven regions (numbered 32 through 37 and 20) located in the northwest quadrant of the outside ring of the core. In these regions, the code underpredicted the measured region outlet temperature by as much as 50 F to 100 F in the 40 to 70 minute time frame of the cooldown following the scram. Agreement between RECA3 predictions and all~other measured region outlet temperatures was good.
As a result 6f the expressed concern, the Staff rer. aired an additional com-parison of predicted and measured core region outlet temperatures during a scram transient subsequent to the installation of the region constraint devices (RCDs). RCDs were installed to eliminate core temperature fluctuations.
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During the period following the installation of the RCDs in November of 1979, several scram transients have occurred in the course of plant operation. The transient selected for comparative analysis occurred on July 8,1980 due to an upset in the feedwater system. Prior to the upset, the plant had operated steadily at the 35% power level thus providing an unambiguous set of initial conditions for the simulation. In addition, essential data including core inlet and outlet temperatures, loop flow and orifice positions were available throughout the transient.
Just af ter the scram, there was a brief (approximately one minute) interrup-tion of forced circulation. Following this, primary circulation was reestablished on Loop 1 only and the core cooldown proceeded. During the 30 to 35 minute time frame in the cooldown, primary coolant flow was reduced by approximately 25% and the operator began adjusting core refueling region orifice positions in the outer ring for low power operation. The time frame of this cooldown is thus relatively long compared to previous cases selected for RECA3 verification, but these cooldown features have been included in the simulation.
Methods used were identical to those reported in Reference 1 so that a valid comparison between this verification and previous results can be made. As I
i in the previous cases, the RECA3 core model was initialized with " measured" region peaking factors (RPFs), indicated orifice valve positions, estimated l
core inlet temperature based upon measured circulator helium temperatures and measured core power. Primary loop flow was taken as measured at the circulator I
1Predictive uncertainty should not be abnormally excessive when compared to the average (Ref. 2).
RPF is " measured" for each region using the common core inlet helium temper-ature, the sensed region outlet temperature and a region flow inferred from orifice valve position using experimental valve flow characteristics.
Measured RPF is the result of a heat balance using thic data.
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. and adjusted by -2.6% to bring all RECA3 steady-state core outlet temperatures into close agreement with measured values. Active core / side reflector flow and leakage flow not passing through the core barrel were calculated utilizing the RECA3 primary loop flow distribution. Following this simulation of core initial conditions, a scram transient from the 35% power level was initiated.
7 Core inlet temperature was assumed equal to circulator inlet temperature during the transient, and circulator measured flow was corrected as above to b
obtain core flow throughout the transient. Afterheat was calculated based upon plant operating history as applied to the FSAR af terheat curve
~
(FSAR Fig. D.1-9).
Results of the comparison of predicted and measured core region outlet tempera-i tures with RCDs installed are presented in Attachment 1 on a core region-by-region basis. Again the predicted outlet temperatures show good agreement with the measured values in all interior core regions and most regions of the outside i
ring bordering the side reflector. Agreement in the northwest quadrant of the outer ring is markedly improved with regions 32, 33, 35 and 36 now showing good agreement and maximum differences near the end of the cooldown are reduced.
j l
However, some discrepancies in the northwest quadrant remain (see regions 20, 34 and 37).
1 The measured outlet temperature vs. time behavior of region 20 is perhaps the best illustration of the remaining discrepancies. Following the reactor scram, the measured outlet temperature may be seen to rise about 50 F during the l
first 10 minutes of the cooldown. This change (sometimes referred to as
" retrograde" behavior) cannot be attributed to heat generation and must therefore be caused by a flow redistribution. The measured region outlet temperature transient is thus the result of changes in the gas flow / temperature environment at the thermocouple as modified by the thermal response of the I
sensing assembly.
RECA3 is dependent upon the use of orifice position to infer core region flow 4
throughout the transient. Studies have shown that measured core region outlet temperature changes of the magnitude observed in region 20 can be f
induced by variations in the region crossflow through the fuel element (block)'
l interfaces (jaws flow) and through variations in the small amount of cool i
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- t. transverse flow which can potentially move along the inside of the core out-let thermocouple tube in boundary region core support blocks (Type II flow).
Changes in either of these flow conditions have the potential to introduce anomalies into the RECA3/ scram transient comparisons, since " measured" RPFs have been used as a part of the RECA3 initial conditions and since major operational events such as the scram may lead to such changes. Measured region peaking factors differ from the values calculated using FSAR diffusion /
depletion methods by a significant amount in regions 20, 34 and 37. Opera-tional data including measured steam generator module inlet helium tempera-tures confirm the calculated region outlet temperatures based upon calculated RPFs. -(Detailed background and evidence for the discounting of certain measured RPFs is presea*ed in the Appendix to this submittal.)
To evaluate the effect which the use of " measured" RPF values may have had on the RECA3 verification, the calculated RPF values have been utilized in a second comparison of measured and predicted core region outlet temperatures for the scram of July 8, 1980. All other methods and conditions remained 4
as before. presents these results on a core region-by-region basis.- Pre-dicted outlet temperature agreement remains generally good in the interior regions and in many regions of the outside ring. The retrograde temperature behavior of region 20 at the beginning of the transient is explained as a loss of Type II flow such that the measurement can more closely approach (with allowance for sensing assembly response characteristics) the predicted actual value. A similar behavior is exhibited in regions 34 and 37.
It is concluded that the installation of the region constraint devices coupled j
with the use of calculated RPFs in the RECA3 simulation has resolved the discrepancies exhibited in the northwest quadrant.
A region-by-region review of all results presented in Attachment 2 reveals one final anomaly in agreement between predicted and measured temperatures which must be addressed. During the latter part of the transient, starting in the 40 to 70 minute time frame, the predicted region outlet temperature moves noticeably below the measured value in region 20 and more particularly l
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l o in region 23.
It is postulated that this anomaly results from region outlet temperature measurement uncertainties due to local effects at the sensing element. Such a deviation starting beyond the point 30 minutes into the cooldown would be influenced by the change in total coolant flow initiated at this point in time and/or the movement of orifice valves in regions 21, 22 and 23 in the 30 to 45 minute time frame.
In particular, it should be noted that regions 20 and 23 are located in the outside ring of active core bordering on the side reflector. These are five column regions with two side reflector columns to complete the standard region configuration. This is significant because, in a cooldown transient, the reflector and core support elocks are last to reach low temperatures due
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to their low cooling surface to volume ratio. During the first 17] minutes of the cooldown, the side reflector support block average temperature and tne region 23 support block average temperature are both predicted to-be hotter than the outlet gas temperature. Thus, there is a nearby potential source for hotter gas which may influence the sensed outlet temperature.
Also, regions 20 and 23 are at the " ends" of the core outlet thermocouple tube. A Type II flow could enter from the warmer side reflector area and influence the region outlet temperature melasurement.
Based on the foregoing considerations and results, it is concluded that region flow and outlet temperature measurement discrepancies can account for the magnitude of the anomalies evident in these RECA3/scran data comparisons.
Predictive uncertainties are not abnormally excessive and discrepancies in the northwest quadrant are resolved. Hence, the additional RECA3 code verification has been completed as requested by the NRC staff.
APPENDIX MEASURED VERSUS CALCULATED FORT ST. VRAIN POWER DISTRIBUTIONS 1.
REGION PEAKING FACTOR-(RPF)
As part of the Fort St. Vrain (FSV) initial rise-to-power pro ram, a u
startup test (SUT B-4, part 2) was conducted to measure the distribution of power generation among the 37 refueling regions of the core. The power dis-tribution among the core regions is characterized by the region peaking factor (RPF), which'is the ratio of the power density in a region to the core average power density. The purpose of the test was to demonstrate that the power distributions were within the limits stated in the bases of Technical Specification LCO 4.1.3.
The initial measurements were made in April 1975,'
with subsequent measurements in July 1976.. Numerous additional comparisons of measured and calculated core radial power distributions have continued throughout cycle 1 and 2 operation.
1.1 Method of Measurement In part 2 of SUT B-4, the power distribution am'ong the regions is deter-mined from measurements of the coolant temperature rise in each of'the 37 regions of the core during steady-state operation.
Power level, helium flow, l
feedwater flow, coolant pressure, control rod positions, and other parameters.
i are held constant for a period sufficiently long to enable core and individual refueling region outlet temperatures to attain steady-state values, and then data are recorded.
l l
The measured RPFs are determined by first (1) measuring the coolant temperature rise across each region, and (2) -inferring the belium flow through each region based upon the orifice valve positions and circulator helium flow rate. Based upon the measured coolant temperature rise across each region and the inferred flow through each region, the " measured" RPFs are calculated.
d Thus, the " measured" RPF is actually calculated based upon temperature measurements and inferred flows.
l.
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t APPENDIX Any discrepancy between the " measured" RPF and that calculated by the core physics code (GAUGE) may be due to (1) a temperature measurement error, (2) an errer in the inferred flow, (3) an error in the physics code, or (4) some combination of these.
1.2 Methods of Calculation The GAUGE code was used to make predictions of the core power distribu-tion in the design of the FSV core and has been used for comparison with measurements of RPF taken during the rise-to-power tests. GAUGE is a two-dimensional X-Y triangular mesh, one-thermal group, diffusion-theory code specifically designed for HIGR calculations. The code will also deplete the fuel for time periods specified by the user and will prcvide depleted nuclide inventories, reactivity, and power distribution as a function of the core history.
1.3 Comparison of Measured and Computed RPF Distribution The measured and computed RPF distributions are compared, and the difference between the two is commonly called the "RPF discrepancy," i.e.,
RPF discrepancy (%) = 100 x, measured RPF - computed RPF computed RPF The comparisons of these measurements with calculations indicated that at power levels greater than about 30%, only a few boundary regions have sig-nificant (410%) discrepancies.
In general, the regions in the north and northwest have a negative discrepancy (indicating that the " measured" RPF j
is lower than calculated) and the regions in the south and southwest have l
a somewhat smaller positive discrepancy (indicating that the measured RPF is higher than calculated). Both measured and computed values are within the LCO 4.1.3 limits, l
Although not required by SUT B-4, core power distribution measurements have been continued throughout the rise-to-power program as a part of the effort to resolve the RPF discrepancy and to understand better the core l
i-l
APPENDlx -
perfor=ance. Detailed investigations of various operating configuration; and of various calculational =odel effects have been made. These include the effects of:
1.
Different orifice positions resulting in equal region flow.
2.
Different total ficw rates.
3.
Operation with less than four circulators, including one-loop operation.
4.
No partially inserted control rods.
5.
Alternate 2-and 3-dimensional calculational models.
6.
Calculational modeling of control rods.
7.
Uncertainties in core composition, i.e.,
fuel, lumped burnable poison, and xenon.
Each of these studies showed some effect on the RPF discrepancy, but no overall significant improvement in the discrepancy was found.
Further studies were made in an attempt to correlate the RPF discrepancy with various plant parameters. The plant parameters studied included reactor power level, power-to-flow ratio, orifice valve position, core pressure drop, region er.it temperature, core flow rate, and shim control rod bank position.
Because of the scatter in the data and because of the inter-relationship of these parameters, it was not possible to identify any one of these parameters as being uniquely related to the RPF discrepancy.
It was shown, however, that in general the north-boundary region discrepancies tend to increase with increasing core pressure drop while the south-boundary region discrepancies are essentially independent of core pressure drop.
At the end of cycle 1 operation, large RPF discrepancies (>20%) existed in a few boundary regions. RPF distribution measurements were continued into cycle 2 operation both before and af ter installation of the region constraint devices (RCDs). Even though the power distribution in cycle 2 is different from that in cycle 1, due to the reloaded fuel and to a different control rod withdrawal sequence, the same boundary regions consistently show large RPF discrepancies. Figure 1 shows a typical RPF discrepancy distribution for cycle 2.
It is seen that there are only a few regions with an RPF
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APPENDIX discrepancy >10%.
Again, both measured and computed data are within LCO 4.1.3 limits.. The north boundary region negative discrepancies indicate that the measurement is lower than calculated and south boundary region positive discrepancies indicate that the measurement is higher than calculated.
2.
POSSIBLE CAUSES OF RPF DISCREPANCY 2.1 Region Exit Temperature Measurement Error From the early comparisons of measured and computed RPFs it was speculated that the major portion of the large discrepancies was due to Type II flow, i.e.,
cool gas flowing along the inside of the thermocouple string resulting in a measured temperature lower than the actual region exit temperature.
Most of the flow which bypasses the coolant channels and flows through the gaps between fuel elements exits into the lower plenum through the gaps between the core support floor blocks. However, some of this flow can pass through the inside of the thermocouple sleeve if a pressure differential exists to drive it.
This Type II flow path is sheen in Figure 2.
The following factors are given as evidence that Type II flow exists:
1.
Due to a variety of reasons, e.g., unequal inter-region gaps, pressure differences may exist from one side of a core support block to the other. Thus flow inside the thermocouple sleeve may be expected.
2.
In general, the regions with the largest negative RPF discrepan-cies are beundary regions at the inlet of the thermocouple strings.
Temperatures in these regions are more likely to be affected by Type II flow since their thermocouple sleeves are open to possible cool helium flow coming from gaps in the side reflector.
3.
Steam generator helium inlet temperatures in modules near regions 20, 34 and 37 are higher than would be predicted using the mea-sured region exit temperatures. However, they are about equal to predictions which use calculated region exit gas temperatures based upon core physics RPF calculations.
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APPENDIX i 4.
The shape of temperature profiles measured by traversible thermo-couples across the core support floor blocks indicate Type II flow effects.
5.
The RPF discrepancy in some of the boundary regions increases with increasing core pressure dron. This is consistent with Type II flow, i.e.,
the driving potential for Type II flow rates l
would increase with ine.reasing core pressure drop.
6.
A few regions with large RPF discrepancies show an unpredicted response following a loop isolation or reactor scram, i.e., the.
region exit temperature increases prict to decreasing whereas most regions show a normal decrease. This " retrograde" behavior can be explained by a reduction in cool Type II flow resulting i-from-the large decrease in core pressure drop which accompanies.
the transient.
7.
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during fluctuations and lack of correlation of these changes with adjacent steam generator module temperature changes indicate changes l
in thermocouple sleeve flow are occurring.
4 8.
Analyses have shown that small, cool themocouple sleeve flows can cause the observed temperature measurement errors.
Type II flow has also been evident during fluctuations and during the region exit temperature redistribution.
4 2.2 Flow Error The region flow is not directly measured but rather is' inferred based on the orifice position and the orifice coefficient. ' Orifice coefficients.were i
measured during cycle 1 operation and agreed well with prediction. Special tests have shown that the orifice position indicators.are accurate to within 1%. This alone would indicate that region flow is correctly inferred.
j However, there is some evidence.to indicate the existence of region flow i
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APPENDIX leakage which would result in an incorrect inferred flow and hence an incor-rect RPF. This flow leakage is called crossflow (or jaws flow),
i.e.,
flow into a region through very small horizontal gap openings between fuel ele-ments. Jaws flow was evident during some fluctuations tnd is evident in region 35 during the region outlet temperature redistribution. Therefore, the differences between calculated and measured RPF may be due, in part, to errors in inferred flow caused by jaws crossflow.
2.3 Calculational Error The RPF distribution calculated by the GAUGE code has been shown to be essen-tially (within +5%) the same as that calculated by other codes, including three-cimensional models, models using more neutron energy groups, and models using a finer geometrical mesh structure than GAUGE. The GAUGE code has also been successfully used to predict power distrik tions and other core physics parameters in the Peach Bottom HTGR. The code nas also provided accurate predictions of core reactivity, control rod worths, and_ shutdown margins for the FSV core. In addition, the RPF for most of the regions in FSV is calcu-lated to be within 10% of the measured RPF. Based on extensive studies of calculational modeling effects, the calculated RPFs are discounted as the source of the large discrepancies.
2.4 Thermocouple Location Error Thermocouple transverses across the core support floor blocks indicate no error in the position of the permanent region exit thermocouples.
APPENDLX 3.
SUMMARY
Af ter extensive analyses of measured and computed RPF distributions during both cycle 1 and cycle 2 operation, significant (>10~) RPF discrepancies still exist it several core boundary regions. The discrepancies in the north boundary regions are typically negative (indicating a measuremenc lower than calculation) and increase with core pressure drop, while the smaller discrepancies in the south boundary regions are typically positive (indicating a ceasurement higher than calculation) and are essentially inde-pendent of core pressure drop. Based upon extensive investigation, the core physics calculated RPFs are discounted as the source of the discrepancy.
The most probable major cause of the RPF discrepancies is a small amount of cool helium flow inside the core outlet thermocouple sleeve (Type II flow),
which results in an incorrect measurement of the region exit temperature in selected boundary regions. Jaws-type crossflow may also contribute to the discrepancy. Therefore, the "RPF discrepancy" is more properly charac-terized as a region outlet temperature / flow discrepancy.
ATTACHMENT 1 COMPARIS0N OF MEASURED DATA (-) TO RECA (X)
SCRAM FROM 35% POWER ON 07/08/80 MEASURED RPFs
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