ML20040G368
| ML20040G368 | |
| Person / Time | |
|---|---|
| Site: | Fort Saint Vrain  |
| Issue date: | 02/28/1982 |
| From: | GENERAL ATOMICS (FORMERLY GA TECHNOLOGIES, INC./GENER |
| To: | |
| Shared Package | |
| ML20040G366 | List: |
| References | |
| NUDOCS 8202120218 | |
| Download: ML20040G368 (69) | |
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TESTING AND OPERATION OF FORT ST. VRAIN UP TO 100% POWER DRAFT General Atomic Company February 1982 l i 8202120218 820209 PDR ADOCK 05000267 > P PDR
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CONTENTS P, age
1.0 INTRODUCTION
1-1 2.0
SUMMARY
AND CONCLUSIONS 2-1 30 TESTING 3-1 3.1 Objectives 3-1 32 Test Description 3-1 3.3 Test Sequence 3-2 4.0 DATA AND INTERPRETATION 4-1 4.1 Observations on Data 4-1 4.1.1 Measured Region Outlet Temperature Redistributions 4-1 4.1.2 Other Effects of the Redistribution 4-4 4.2 Analysis and Interpretations 4-16 4.2.1 Expected Versus Measured Region Outlet Temperatures 4-16
,,, 4.2.2 Gap Temperature Changes 4-20 4.2 3 Core Resistance Changes 4-23 4.2.4 Region 35 Crossflow Calculation 4-23 4.2.5 Type II Flow Effects 4-24 4.2.6 Location of Type II Flow and Crossflow 4-25 4.2 7 Nuclear Channel Deviations 4-34 4.2.8 Reactivity Perturbations 4-36 4.3 Scenario of Events 4-37 4.3.1 Introduction 4-37 4.3 2 Temperature Redistribution Scenario 4-37 t
11
Y s Pagg 5.0 SAFETY CONSIDERATIONS 5-1 5.1 Introduction 5-1 5.2 Safety Evaluation of the Outlet Temperature Redistribution 5-1 5.2.1 Wide Range Linear Channel Flux Signals 5-2 5.2.2 control Rod Insertability 5-2 5.2 3 Structural considerations 5-3 5.2.4 Secondary System 5-4 5.2.5 Bypass Flow Increase After the Outlet Temperature 5-5 Redistribution 5.2.6 Accident Analyses 5-6 53 conclusions 5-6 6.0 LONG TERM OPERATION 6-1
7.0 REFERENCES
7-1 l l \ l l l l 111
3) . s
1.0 INTRODUCTION
During the initial rise-to-power program of the Fort St. Vrain reactor in October 1977, while approaching 60% power, temperature fluctuations were observed in the primary coolant circuit at the outlets of individual core regions and the inlets to steam generator modules. A comprehensive program of investigation into the nature and cause of the temperature fluctuations was initiated immediately. The fluctuation investigations led to the design and fabrication of region constraint devices (RCDs) as a solution to the problem. These mechani-cal links were installed on the top of the core in November 1979. They were installed at locations where three regions intersect and are designed to provide inter-region linking to stabilize the gaps between regions at the top of the core to near nominal values. Steady-state tasting performed during initial operation following installation of the RCDs verified that the overall core performance was unaffected by the presence of the RCDs. Testing to evaluate the success of RCDs as a solution to the temperature fluctuations was first performed in November and December of 1980. These tests confi"med that the RCDs were successful at preventing fluctuations up to 70% power and a core pressure drop of 4.2 paid. However, once in November and again in December, at a transient peak core pressure drop of 3.8 psid, following an increase in reactor power, a region outlet temperature redistribution was observed. These redistributions resulted in several boundary region outlet temperatures, particularly in the NW section of the core, decreasing while inner core region outlet temperatures generally increased somewhat more than would be expected from the power increase. In March 1981, NRC approved testing of Fort St. Vrain above 70% power. Testing to confirm the success of the RCDs as a solution to the temperature fluctuations and to investigate the region outlet tempera-ture redistribution above 70% power was conducted in Cycle 2 during March, April, and May of 1981, and in Cycle 3 during October and November of 1981. Again, no fluctuations occurred, but redistributions 1-1
.g. . 6 similar to those experienced in November and December of 1980 were observed. Typical results of these tests are summarized in this report, an evaluation of safety consequences is presented, and plans for long term operation of FSV above 70% power are described.
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. s 2.0
SUMMARY
AND CONCLUSIONS Testing wherein attempts were made to induce fluctuations, after installation of region constraint devices, was conducted during Cycle 2 in November and December of 1980 and in March, April, and May of 1981. During Cycle 3, testing was conducted in October and November of 1981. Reactor power levels from 40% to 100% of rated thermal power have been surveyed with a maximum core pressure drop of 5.0 paid. No fluctuations have been observed, even in operating regimes where fluctuations occurred prior to installation of the region constraint devices. However, as reactor power was increased rapidly ($35 per minute) in small (s3%) steps from 40% to 100% power, ene or more region outlet temperature redistributions were observed. These redistributiens were generally observed at core pressure drops between N3.5 and N4.0 psid. They esulted in several boundary region outlet temperatures, particularly in the NW sector of the core, decreasing while inner core region outlet temperatures generally increased somewhat more than would be expected from the power change. The region outlet temperature redistributions are all similar in character. While the same regions do not always participate to the same degree, the phenomena which take place in the core have been determined to be essentially the same in all cases. The region outlet temperature redistributions are the result of small in-core displacements. These displacements are similar in nature to the initial motion which occurred during fluctuations. However, these displacements are not cyclic. These small (on the order of 0.10 in or less) displacements cause changes in gap distribution, changes in crossflow, and (for the seven NW boundary regions, Regions 20 and 32-37) changes in the amount of cool transverse helium flow (Type II flow) along the sleeve (s) surrounding the region outlet temperature thermocouples. These observations are consistent 2-1 i
e a with a general (although asymmetric) tightening of the inner regions of the core, wherein the gaps around the outer regions genertily are increased and gaps between inner regions are generally decreased. Calculations done during and in support of various testing performed from 40% to 100% power and analyses of data from these tests indicate that, even before a redistribution, significant region outlet temperature measurement discrepancies exist in the seven NW boundary regions (Regions 20 and 32-37). Extensive evaluations of observed differences between calculated and measured (inferred) region peaking factors and steam generator module helium inlet temperatures have provided further evidence of region outlet temperature measurement discrepancies in the NW boundary regions. Thus, the region outlet temperature redistributions amount to additional perturbations of pre-existing measurement discrepancies in the NW boundary regions as well as real changes in the outlet temperatures of the remaining regions. Evaluations of the region outlet temperature redistributions indicate that these events involve no unreviewed safety questions. A method for operating the reactor which accounts for region outlet temperature measurement discrepancies both before and after a redistri-bution has been developed. Under this operating method, the seven NW boundary regions, which are susceptible to outlet temperature measure-ment errors, will be operated by comparison regions in a manner similar to that employed in test procedure RT-500K. For the other 30 regions in the core, indicated changes in the region outlet temperature which occur during a region outlet temperature redistribution are real. These temperature changes can be accommodated and corrected as desired by orifice valve adjustments as are made routinely following load changes. In support of this operating method, appropriate revisions to the Technical Specifications will be submitted to the NRC for review and approval. 2-2
Fluctuation testing up to 100% power has been completed. The testing has demonstrated that region constraint devices are successful at preventing fluctuations for power levels up to 100% and core pres-sure drops up to 5.0 paid. The data and experience gained operating at full power indicate that with proper operating procedures the plant can be operated in a stable manner at 100% power without increased risk to the health and safety of the public. I s 8 o 0 2-3
30 TESTING 3.1 Objectives Testing above 70% reactor power was conducted per the procedures of RT-500K (Ref. 1). The main objectives of this test were 1) to con-firm that the region constraint devices (RCDs) had eliminated fluctua-tions at all power levels up to 100% of rated thermal power and 2) to obcain data and operating experience during region outlet temperature redistributions for use in developing procedures for normal long-term operation. 3.2 Test Description With the plant as close to normal operating conditions as possible within the limits of RT-500K, reactor power was increased in steps of
$3% at $3%/ min. from 40% to 70% power. After each power increase data were recorded for 1 hour. Above 70% power the power was increased N3%
at NO.5%/ min., and data were recorded for 1 hour. The power was then reduced s3% and the power increase repeated at s3%/ min., and again data were recorded for si hour. This procedure was repeated until the reactor reached 100% of rated thermal power. Special procedures were provided for operating those regions with significant region outlet temperature measurement discrepancies in such a manner that their operation would be in compliance with both the letter and the intent of the limits of Technical Specification LCO 4.1.7. These procedures involved the operation of such regions based on the operation of selected " comparison regions."1 A region was 1 0peration of a discrepant region by a " comparison region" is based upon knowing the relative power of the two regions (from core physics calculations) and adjusting the region inlet orifice valves such that
- 1) the outlet temperature of the comparison region is within Technical (Footnote continued) 3-1 1
- , r
O 4 operated per a comparison region if the calculated region outlet temperature was more than 700F higher than the measured temperature. Procedures were also provided for continued operation after a region outlet temperature redistribution occurred. In RT-500K a redistribution was considered to have occurred, by definition, if the measured outlet temperature of one or more regions dropped by 200F or more following an $35 increase in reactor power. This procedure required an adjustment to the allowable region outlet temperature mismatch 2 for those regions (other than regions being operated via comparison regions) where the outlet gas temperature decreased by
>200F.
33 Test sequence RT-500K was performed in March, April and May 1981, during Cycle 2 operation, and in October and November 1981, during Cycle 3 operation. The testing sequence is summarized below: March 1981 (Cycle 2) Testing was performed at power levels from 40 - 70% power. At 70% power a turbine trip occurred, after which excessive leakage of purified helium from the PCRV penetration interspace of steam generator module B-2-3 to the cold reheat steam was detected and testing was terminated. No redistributions were observed. l(continued) Specification limits, and 2) the discrepant region has a power-to-flow ratio equal to or less than that of the comparison region. 2 Measured region outlet temperature minus core average outlet tempera-ture. 3-2
April 1981 (Cycle 2) Testing was continued at power levels from 70 - 915 power. At 91% power a loop trip and subsequent reactor shutdown due to high reheat steam temperature following a flow upset occurred, and testing was terminated. Two redistributions were observed. May 1981 (Cycle 2) Testing was performed at power levels from 70 - 78% power. At 78% power a turbine trip occurred due to turbine vibrations. Testing was terminated, and the plant was shutdown for turbine inspection and subsequently for the second refueling. Two redistributions were observed. The sequence of events for these Cycle 2 tests is given in Table 3-1 and Fig. 3-1. October 1981 (Cycle 3) Testing was performed at power levels from 40 - 75% power. At 75% power the purified helium leak was again detected from steam generator module B-2-3 interspace, and testing was terminated. No redistributions were observed. November 1981 (Cycle 3) Testing was continued at power levels from 68 - 100% power, completing the requirements of RT-500K. The procedure for this testing period was altered to include only the fast power rises (i.e., N3%/ min.) to 90% power, since that power level had previously been achieved in April 1981. Three redistributions were observed. l The sequence of events for Cycle 3 testing is given in Table 3-2 and Fig. 3-2. 3-3
TABLE 3-1 Sequence of Events-Cycle 2 Testing Steady State Conditions ' Date (time) Prior to Power Increase Comments Powerl31 Flow Core AP-(%) (%) (psid) March 1981 3/19/81 (0415) 42.5 57.8 1.63 Initial conditions 3/19/81 (0608) 45.3 59.9 1.76 3/19/81 (0856) 48.0 60.8 1.83 3/19/81 (1340) 51.1 61.2 1.86 3/19/81 (1703) 53.6 64.2 2.04 3/19/81 (1843) 56.9 65.5 2.16 3/20/81 (1837) 62.1 70.2 2 38 3/20/81 (2037) 65.9 73.5 2 59 3/20/81 (2231) 69.1 74.7 2.69 3/20/81 (2315) 71 3 76.0 2.80 After Final power increase 3/21/81 (0230) Turbine Trip - Testing Terminated April 1981 4/16/81 (1434) 70.7 77.1 2.80 Initial conditions 4/16/81 (2307) 71 3(1) 77.6 2.87 4/17/81 (0503) 71.5 78.7 2.89 4/17/81 (0624) 76.7(1) 82.8 3.16 4/17/81 (1057) 76.4 83 3 3 21 4/17/81 (1328) 79.5(1) 85.5 3.40 4/17/81 (1751) 79.1 86.6 3.38 4/17/81 (1830) 82.1(1) 89.2 3.56 4/19/81 (0130) Reactor Shutdown
-4/23/81 (1601) 72.2 78.8 3 06 Initial conditions 4/23/81 (1932) 72.4 78.6 3 07 4/23/81 (2147) 76.2(2) 83.5 3 38 Redistribution 4/24/81 (0745) 80.5 87.4 3.54 4/24/81 (1111) 82 9(1) 90.2 3.80 3-4
TABLE.3-1 (Cont'd.) Steady State Conditions Date (time) Prior to Power Increase Comments Powerl31 Flow Core AP (%) (5) (psid) 4/24/81-(1412) 81.9 89.2 3 72 4/24/81 (1543) 86.(11)(2)92 7 4.03 Fedistribution 4/24/81 (1620) 91.2 96.5 4.35 After final power increase 4/25/81 (0140) Reactor Shutdown - Testing Terminated May 1981 5/13/81 (0721) 70.4(2) 79.7 3 13 Redistribution 5/13/81 (0928) 74.7(2) 84.2 3.44 Redistribution 5/13/81 (0958) 79.4 89.4 3.81 After final power increase 5/13/81 (1335) Turbine Trip - Reactor Shutdown - Testing Terminated and Second Refueling Commenced 1 (1) Conditions prior to 1/2%/ min. power increase - all other increases were 3%/ min. (2) Temperature redistribution occurred during increase from this power. (3) Determined by secondary side heat balance. l i 3-5
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TABLE 3-2 Seouence of Events-Cycle 3 Testing Steady State Conditions Date (time) Prior to Power Increase Comments Powerl3) Flow Core AP (5) (%) (psid) October 1981 10/21/81 (1036) 39.0 55.1 1.33 Initial conditions 10/21/81 (1255) 43 7 58.2 1.49 10/22/81 (0105) 46.1 59.5 1.56 10/22/81 (0210) 50.2 63 0 1.76 10/22/81 (0332) 53 0 67.1 2.00 10/22/81 (0655) 57.6 66.3 1.94 10/22/81 (1020) 60.8 70.9 2.20 10/22/81 (1050) 66.8 77.0 2.59 After final power increase 10/22/81 (1310) Loop Trip 10/24/81 (0204) 70.7 76.2 2.80 10/24/81 (0235) 74.9 79.2 3.04 After final power increase B-2-3 He Interspace Leak November 1981 11/5/81 (1115) 68.6 75.7 2.53 Initial conditions 11/5/81 (1327) 73.8(2) 81.6 2 96 Redistribution 11/5/81 (1744) 78.1 85.6 3 26 11/5/81 (2016) 82.2(2) 88.2 3.53 Redistribution 11/6/81 (0447) 85.7 91.3 3.63 11/6/81 (0930) 88.0(1) 96.5 3.95 11/6/81 (1106) 87.9 96.5 3.96 11/6/81 (1225) 90.6(1) 98.6 4.14 11/6/81 (1406) 92.2(2) 99,9 4,14 Redistribution 11/6/81 (1534) 96.4(1) 105.4 4.55 11/6/81 (1547) 100.1 109.2 4.92 Initial full power operation 11/7/81 (0030) 98.0 103.6 4.65 11/7/81 (0102) 101.1 107.0 4.93 After final power increase (1) Conditions prior to 1/2%/ min. power increase - all other increases at 3%/ min. (2) Temperature redistribution occurred during increase from this power. (3) Determined by secondary side heat balance. 3-7
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4.0 DATA AND INTERPRETATION 4.1 Observations on Data As discussed in Section 3 0, a total of seven RT-500K test sequences have been performed - four in Cycle 2 and three in Cycle 3 In this section emphasis will be given to data obtained during the current fuel cycle, Cycle 3, because it is enst pertinent to the current core configuration. In Cycle 3 testing, region outlet tempera-ture redistributions were observed during the third test sequence, which was conducted in November 1981 (Fig. 3-2). During this sequence power was increased, without interruption, from 70% to 100% power. The reactor was then operated in a stable manner for approximately 60 hours at 100% power, followed by a shutdown for planned plant maintenance and modification of the circulator auxiliary systems. In RT-500K, a redistribution was considered to have occurred, by definition, if a measured decrease in exit temperature of a region of 0 20 F or more was observed following an N3% increase in reactor power. As shown in Table 3-2, three events occurred during the November 1981 testing sequence which met the definition of a redistribution. 1 In this subsection, representative data from the November 5, 1981 redistribution between 82% and 86% power are presented. Interpreta-tions of these data are provided in Section 4.2 of this report. 4.1.1 Measured Region Outlet Temperature Redistributions f The region outlet temperatures were continuously measured prior to, during, and after all temperature redistributions. The tempera-tures before, during, and after the November 5, 1981 redistribution between 82% and 86% power are shown in Figs. 4-1 (interior regions) and 4-2 (boundary regions). These are typical behavior of region outlet temperatures for redistributions during Cycle 3 Note that the inner 4-1
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o Channel 3 increases less than the average in all events except the November 5, 1981 event shown in Fig. 4-5, when it increased slightly more than the average. o Channel 7 sometimes increases more than the average and sometimes less than the average. From these nuclear channel deviation data it may be concluded that, in general, the responses were similar for all temperature redistribution events. The nuclear channel deviations were never cyclic as observed during fluctuation events. 4.1.2.2 Core Reactivity Perturbations Careful examination of the nuclear channel signals during the initiation of the region outlet temperature redistribution again reveals the existence of a small reactivity change of less than +1d (0.00005 40) not due to control rod motion. Similar effects were observed for redistributions in November and December of 1980 (Ref. 2). 4.1.2 3 Gap Temperature Change Twenty-six special test thermocouples are installed in the calibration tubes at the core outlet as shown in Fig. 4-6. Most of these thermocouples indicate temperature changes at the time of a temperature redistribution. During temperature redistribution events, most of the gap thermo-couples located near the N-NW core boundary have in the past indicated a temperature decrease (typically < 400F) while most other thermocouples indicated a temperature increase. Tne same behavior has been observed for the most recent redistributions. For instance, at the time of the temperature redistribution between 82% and 86% power on November 5, 1981, thermocouples 4, 5, 8, 9, 11, 12, 13, 14, 22 and 26 showed temperature decreases, although with somewhat different magnitudes than 4-8
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Figure 4-6. Calibration tube thermocouple locations 4-9
in previous cases. All other thermocouples showed temperature increases. Typical plots of gap temperatures for this redistribution as a function of time during which'the redistribution occurred are shown in Figs. 4-7 through 4-9. 4.1.2.4 Regions 5 and 35 Temperatures Regions 5 and 35 contain instrumented control rod drives (ICRDs). This instrumentation includes thermocouples located axially near the top, middle and bottom of the region. The response of these thermo-couples at the time of the November 5, 1981 temperature redistribution
~
between 82% and 86% power is shown in Figs. 4-10 and 4-11 for Regions 5 and 35, respectively. Also shown in Figs. 4-10 and 4-11 are the corresponding region outlet temperatures and the temperatures of the calibration tube thermocouples located under these regions. (The outlet temperature for Region 5 is not shown on Fig. 4-10 since it is essentially the same as the temperatures 'hown s by the gap thermo-couples.) Region 5 (interior region) temperatures show a normal behavior (increase) during the power increase, whereas 15e temperatures in Region 35 (NW boundary regi'n) o generally show a decrease. 4.1.2.5 Core Resistance The core resistance (R)3 has typically shown an abrupt decrease of up to 5% at the time of a temperature redistribution (Ref. 2). 3R = 2 x 1013 Ap p T(5) 1.85 where R = core resistance Ap = core pressure d' rop, paid P = PCRV pressure, paia T = average circulator inlet temperature, OR m = total circulator flow rate, lbm/hr 4-10
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Figure 4-12 shows a plot of the resistance as a function of time, which includes the time of the redistribution which occurred between 82% and 86% power on November 5, 1981. Note the small decrease of sl%. This small decrease in resistance is evidence that this temperature redistribution is of smaller magnitude and/or of less extent than redistributions previously observed during Cycle 2 operation. - 4.2 Analysis and Interpretations The previous discussions were concerned primarily with the presen-tation of some significant data and some comparisons to data of pre-vious redistributions. In this section the observed measurements are explained and compared to calculations and mechanisms which tend to reproduce the observed data. 4.2.1 Expected Versus Measured Region Outlet Temperatures During a normal power rise (taken on the regulating rod) of 3-4% the region outlet temperature is expected to increase s400F in Region 1, s20-250F in the ring 2 regions, s10-150F in ring 3 regions, and s5-100F in ring 4 regions. Comparisons of these expected (calculated) temperature changes with those measured are useful in detecting which regions participate in temperature redistributions. These changes in temperature, as a result of the 82-86% power rise that initiated a temperature redistribution on November 5, 1981, are shown in Figs. 4-13 and 4-14 for each of the 37 regions. Note that the measured outlet temperature changes for all interior regions (1 through
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B D 4.2.2 Gap Temperature Changes The gap temperatures are observed to change during the region outlet temperature redistribution. The temperature change can be explained by gap size changes which affect the heat transfer and flow (Ref. 3). A thermal-flow model of the gap between regions had been previously constructed (Ref. 3) and used to make calculations of gap temperature and flow responses. In the previous redistribution analysis (Ref. 2), a series of steady state analyses were performed for the reactor conditions before and after a temperature redistribution as a function of inter-region gap size. The gap temperatures before and after the temperature redistribution were then tabulated from the measured data. The gap temperature before the redistribution indicates the approximate initial size of the gap, and from the change in gap temperature the approximate final gap size may be determined as well. Similar results for the November 5, 1981 redistribution between 82% and 86% power are shown in Table 4-1 and again in Fig. 4-15 on the core map. Thers were only five gap thermocouples in the interior of the core which showed any significant changes for this redistribution event. All gap thermocouples listed La Table 4-1 had significant changes for most redistributions in Cycle 2. There are clearly fewer indicated motions for the November 5,1981 event although there was one difference for indicated movement above gap T/C 11. This gap appears to have opened for this case, whereas in November of 1980 the gap appeared to close. The transient calculations which were performed for the November 1980 redistribution (Ref. 2) were also used to determine the approxi-mate size of the gap redistributions of this event. The present observations of gap temperature change are consistent with the arguments of gaps opening or closing and is, therefore, deduced to be l 4-20 l
TABLE 4-1 ESTIMATED GAP CHANGES (November 5, 1981 At About 20:20) T/C T1 T2 G1 G2 G2-G1 3 1415 1465 0.02 0 -0.02 4 1345 1300 0.12 0.18 4 .06 7 1470 1500 0 0 0 8 1400 1400 0.02 0.08 +0.06 11 1330 1250 0.20 0.30 +0.10 15 1450 1480 0 0 0 18 1440 1470 0 0 0 19 1440 1460 0 0 0 20 1450 1470 0 0 0 21 1470 1490 0 0 0 22 1295 1260 0.25 0 31 +0.06 23 1395 1395 0.02 0.02 0 24 1420 1420 0.02 0.02 0 26 1130 1125 0.35 0.35 0 T1 - Calculated gap temperature before redist:*ibutien (OF) T2 - Calculated gap temperature after redistribution (OF) G1 - Gap size before redistribution (inches) G2 - Gap size after redistribution (inches) 4-21
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CALCUIATED G 12JS 36 37 20-CHANGE IN O 8 4 GAP (INCHES) 34 18 8 21
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the cause. The Cycle 3 data, however, indicate far fewer gap width changes. Nonetheless, the indicated magnitude of the width changes is about the same as indicated for the redistributions in November and December 1980. 4.2 3 Core Resistance Changes Calculations were previously performed to estimate the gap redistribution necessary to cause the observed 5% reduction in resistance at the conditions of November 14, 1980 and December 13, 1980 (Ref. 2). Assuming a uniform initial gap distribution, the change in hydraulic diameter which is necessary to cause the observed 5% resistance drop was calculated. Constant total gap area was assumed, i.o., gaps are simultaneously opened and closed. Assuming that half of the gaps close and the other half open, it was shown that the reaistance reduction can be accomplished by displacements of 0.060 in. The test data of November 5, 1981 indicate that the core resis-tance drop at the time of the redistribution between 82% and 86% power was only sl% compared to the s5% of previous redistributions (Fig. 4-12). This smaller change could be explained either by smaller motions or by fewer gaps which are changing. The latter argument is consistent with the observed behavior. That is, fewer gap thermocouples sensed any temperature changes, and those that did change indicated gap changes of 0.06 to 0.10 in., which is nominally the same as experienced in November and December 1980. 4.2.4 R_egion 35 Crossflow Calculation l l l The measured outlet temperature of Region 35 usually decreases at the time a temperature redistribution occurs. This can be attributed to several possible causes, one of which is a change in the influx of cold crossflow into the region (e.g., gas through a slight jawing of the stacked fuel blocks). (Also, a change in Type II flow over the outlet thermocouple can contribute to the change in measured outlet temperature. See Section 4.2.5.) A change in the crossflow (jaw) gas l l l 4-23
entering (or leaving) the control channel within the region is also an explanation for the temperature response of the middle and bottom ICRD thermocouples. The changes in the outlet temperature of Region 35 during the November 5, 1981 redistribution between 82% and 86% power were about the same size as in previous redistributions, indicating approximately the same size jaw openings. Calculations indicate that a jaw size of 0.1 in. existing halfway around a region (a total jaw opening about 6 ft in length) can cause the observed change in region 35 outlet temperature. A jaw at two levels all the way around a region would reduce the necessary jaw size to 0.025 in. Displacing one end of a single block by 0.060 in. can produce a jaw of about 0.03 in. at both the top and bottom of the block. This displacement is again consistent with all other indicators of displacements being on the order of about 0.060 in. to 0.10 in. The ICRD thermocouples at the middle and bottom of the control rod hole exhibit a small decrease in temperature st the time of the temperature redistribution. This decrease can be explained by a change in colder crossflow into the control rod channel. That is, colder crossflow suddenly enters through a crossflow gap (jaw) opening when the temperature redistribution event occurs. This conclusion is based upon calculations performed with a control rod channel thermal / flow model of estimated and actual temperature rises to the middle thermocouple before and after the outlet temperature redistribution. Figure 4-11 shows the observed behavior during the November 1981 event. l 4.2.5 Type II Flow Effects Previous analyses and data (Refs. 4 and 5) have consistently shown the existence of a cool transverse flow along the thermocouple sleeve. l This flow can cause the measured outlet temperature to be different i from the actual region outlet temperature for the boundary regions on I 4-24
the end of a thermocouple string. The analysis to date indicates large differences (up to 300 - 4000F) between expected and measured region outlet temperature can exist if a source of cool gas flow is available. Analyses have shown that pressure differences of 0.010 psid across the sleeves extending through individual core support blocks can cause transverse flow of about 10 lbm/hr and a temperature discrepancy of about 500F. Changing the pressure gradient to 0.08 paid increases the Type II flow to 80 lbm/hr and the temperature discrepancy to 4000F. Inter-region gap size changes on the order of 0.10 in, are capable of producing changes of this magnitude in the lateral pressure drop across the thermocouple sleeves, as well as changing the temperature of the gas traversing the sleeve. Therefore, changes in the gap flow through the core support block gaps can cause changes in pressure differences capable of causing changes in cool Type II flow, which induce changes in measured outlet temperature of 4000F even though the actual tempera-ture of the region outlet gas remains constant. In the analyses described above, the temperature of the Type II flow was assumed to be 8000F. This value is based upon measurements of temperature at the inlet of the thermocouple tubes at the NW core boundary which were obtained during steady state traverses of the T/C tubes (Ref. 4). The Type II flow effects are summarized in Fig. 4-16 where the effects of inter-region changes are shown with the present estimated gap sizes. j 4.2.6 Location of Type II Flow and Crossflow l I 4.2.6.1 Type II Flow Location As the previous sections indicate, a region may have both Type II flow and jaws flow; however, in actuality significant Type II flow errors are confined to the seven NW boundary regions. l The difference between the measured and calculated outlet tempera-ture of a region at any point in time is referred to as its temperature discrepancy. It is expected that there could be as much as 70cp I l 4-25
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difference between the calculated and measured outlet temperature due to uncertainties in the physics calculation of region power. However, the discrepancies indicated for the seven boundary regions at the entrance to the thermoccuple strings (Regions 20 and 32-37) are as large as 4000F. These differences must be attributed to Type II flow alone or in combination with jaws flow, with Type II flow being the predominate influence on the discrepancy. It should be emphasized at this point that Type II flow-induced temperature discrepancies are in fact differences between the actual gas temperature and the indicated (measured) gas temperature. On the other hand, jaws flow discrepancies are not differences between actual gas temperatures and indicated temperatures, but rather a discrepancy in what is assumed as the flow rate in the calculation of the tempera-tures. It is important then to isolate, if possible, those regions which have Type II flow-induced discrepancies from those that do not. Only regions with such discrepancies have significant errors in the sensing of the outlet gas temperature. There are three important parameters in Type II flow-induced discrepancies. They are:
- 1. The geometry of the T/C over which the Type II flow occurs.
t
- 2. The magnitude of the Type II flow.
3 The inlet temperature of the Type II flow. With regard to the first item, it is noteworthy that all 37 outlet T/Cs are similar in geometry (see Fig. 4-16) except for the seven T/Cs in the SE sector of the core at the end of the T/C strings i.e. Regions 23, 24, 25, 26, 27, 28, and 29. These seven thermocouples have an additional nose section about 8 inches long on one end of the spacer e 4-27
block and a 4 inch long additional section on the other end (see Fig. 4-17). As discussed below, this difference is'important in eliminating these regions from consideration as having significant Type II flow-induced discrepancies. With regard to the second item, it can be shown that the magnitude of the Type II flow is different for regions that have different gaps on either side. Note that the permanent side reflector support block (PSR) has openings in the NW (to accommodate the thermocouple strings) which extend to the core barrel (Fig. 4-16). On the ether hand, the PSR in the SE is solid (Fig. 4-17). This geometry allows the T/Cs in the seven NW boundary regions to be exposed to the colder gas from the core barrel-PSR gap. Because of the differenca in size between the core barrel-PSR gap and the PSR-region gap, a larger lateral pressure gradient exists and a higher Type II flow rate is possible for the NW boundary regions. Next we note the location where there is a temperature in the gap that is able to cause a temperature discrepancy (Item 3 above). Both calculations and measurements (thermocouple traverses, Ref. 4) confirm that only the gap flow between regions and the side reflector can be of sufficiently low temperature to cause significant discrepancies. These observations then restrict the possibility of effects of Type II flow to only the boundary regions and only the boundary regions at the ends of the T/C strings. Recall that the T/C spacer is different in the SE compared to the NW (Fig. 4-18). In fact, the 8 inch and 4 inch long sections on either end of the T/C spacer block serve both to reduce the Type II flow and to bring this flow into thermal equilibrium with the region outlet te=perature. In addition, due to differences in PSR geometry, a higher Type II flow rate is possible for the seven NW boundary regions. Ihis leaves only the seven NW boundary regions whose temperature sensors may not be giving a reliable indication of the region outlet temperature. The existence of Type II flow in these regions is further 4-28
? I' block and a 4 inch long additional section on the other end (see Fig. 4-17). As discussed below, this difference is important in eliminating these regions from consideration as having significant Type II flow-induced discrepancies. With regard to the second item, it can be shown that the magnitude of the Type II flow is different for regions that have different gaps on either side. Note that the permanent side reflector support block (PSR) has openings in the NW (to accommodate passage of the thermo-couple strings) which extend to the core barrel (Fig. 4-16). On the other hand, the PSR in the SE is solid (Fig. 4-17). This geometry allows the T/Cs in the sev(n NW boundary regions to be exposed to the colder gas from the core barrel-PSR gap. Because of the difference in size between the core barrel-PSR gap and the PSR-region gap, a larger lateral pressure gradient exists and a higher Type II flew rate is possible for the NW boundary regions. Next we note the location where there is a temperature in the gap that is able to cause a temperature discrepancy (Item 3 above). Both calculations and measurements (thermocouple traverses, Ref. 4) confirm that only the gap flow between regions and the side reflector can be of sufficiently low temperature to cause significant discrepancies. These observations then restrict the possibility of effects of Type II flow to only the boundary regions and only the boundary regions at the ends of the T/C strings. Recall that the T/C spacer is different in the SE compared to the NW (Fig. 4-18). In fact, the 8 inch and 4 inch long sections on either end of the T/C spacer block serve both to reduce the Type II flow and to bring this flow into thermal equilibrium with the region outlet temperature. In addition, due to differences in PSR geometry, a higher Type II flow rate is possible for the seven NW boundary regions. This leaves only the seven NW boundary regions whose temperature sensors may not be giving a reliable indication of the region outlet temperature. The existence of Type II flow in these regions is further 4-29
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supported by the behavior of the discrepancy. Type II flow errors are pressure drop (flow) dependent; therefore, they increase magnitude as tha reactor is increased in power. A fixed jaws type flow discrepancy, on the other hand, causes a difference between the calculated and measured temperature which is a constant proportion of the region axial temperature rise. The jaws-induced discrepancy is, therefore, relatively constant (for nominally constant power to flow ratio and therefore region temperature rise) as the reactor power is raised, except for geometry changes (as in redistributions) that cause changes in the jaws flow rate. This phenomenon is demonstrated by Figure 4-18, which shows, as a function of reactor thermal power, outlet temperature discrepancies for a typical inner region (Region 1), a typical boundary region (Region 25 - a SE boundary region)', and a typical NW boundary region (Region 32). Note the significantly different behavior of the discrep-ancy in Region 32, which steadily increases in magnitude with increas-ing reactor power. Note also that the discrepancy in Region 25 is essentially constant as reactor power is increased. Additional supporting evidence is provided by the steady state temperature traverses through the T/C penetration tube, which show the seven NW boundary regions as being affected (Ref. 4). Likewise the observed transient behavior of the boundary region outlet temperatures during redistributions supports this conclusion. Type II flow changes (in the seven NW regions) are sensed almost immediately, whereas jaws i type flow changes are sensed much more slowly due to the heat capacity f of the core. In addition, Type II flow induced temperature measurement error in the seven NW boundary regions is supported by calculations which were made to compute the steam generator module hellun inlet temperatures. The steam generator helium inlet temperatures were calculated in three ways: 1.) using only measured region outlet temperatures; 2.) using only calculated region outlet temperatures (calculated based upon core physics analyses of region peaking factors); and 3.) using the measured 4-31
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outlet temperatures of all regions except the seven NW boundary regions, for which the calculated value was used. The best agreement with the measured helium inlet temperatures for the 12 steam generator modules was obtained using the third method. This result further supports the argument that only Regions 20 and 32-37 have significant temperature measurement errors. It is concluded that cool Type II flow only significantly affects the outlet temperature measurement for those seven regions of the NW sector that are at the entrance to the T/C strings. 4.2.6.2 Location of Crossflow The jaws type crossflow changes occur in the top 1/3 of the core (including the top and bottom reflectors). This is concluded to be the location of crossflow changes based on the following evidence:
- 1. Region 35 ICRD thermocouples show the temperature changes during redistribution to occur above the core midplane.
- 2. 1/2 scale model tests show that the maximum lateral pressure forces to cause block movement occur at about 1/3 the distance into the core.
3 1/2 scale model tests show that the largest pressure differences (and therefore the largest driving force to sustain crossflow) occur about 1/3 the distance into the core.
- 4. 1/5 scale model tests of stacked blocks (side reflector) show that the minimum required force to open gaps occurs at about 1/3 the distance into the core.
- 5. Computer almulation of stacked fuel elements confirms that the minimum force to move the blocks occurs at about 1/3 the distance into the core.
4-33
o .
- 6. The transient temperature changes of regions during changes in crossflow are characteristic of flow changes entering near the top of the core.
4.2.7 Nuclear Channel Deviations A nuclear channel deviation is defined as the response of an individual channel minus the average response of all channels. Nuclear channel daviations have proven to be highly reliable and sensitive indicators of fluctuations (Refs. 3 and 4). The nuclear channel responses during the 82% to 86% power increase which initiated the region outlet temperature redistribution of November 5, 1981 are ehown in Figs. 4-4 and 4-5. The corresponding nuclear channel deviations are also shown in Figs. 4-4 and 4-5. The deviations during the outlet temperature redistribution are characterized by small initial offsets followed by a gradual (10 to 15 min.) approach to a new stable value. These abrupt offset deviation responses are smaller than those typically observed during fluctuations (prior to installation of RCDs), and they do not exhibit any cyclic behavior. For comparison, the deviation response during a Cycle 2 fluctuation is shown in Fig. 4-19 The initial offset behavior (1,0.9%) of the deviations is caused by small changes in the neutron streaming through gaps in the side reflec-tor. The remaining gradual changes (1,0.9%) are responses to thermal effects, e.g., changes in core and reflector temperatures and/or configuration resulting from the redistribution of gaps. Comparing this redistribution event of November 5, 1981 with the redistribution of November 14, 1980 (Ref. 2), it is noted that the maximum nuclear channel deviation in November 1981 is s1.4 times larger than that in November 1980. This is es expected for the same size gap change since the ratio of the power (neutron flux) levels at which these two events occurred is also sl.4. As a result of the difference in power (neutron flux) levels when the two redistributions occurred, a gap size change 4-34
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of the same magnitude would cause an N1.4 times larger nuclear channel deviation in the November 1981 event than in the event of November 1980. This corroborates the conclusion that the gap size changes are of nominally the same magnitude for the two events. While the deviation responses during the region exit temperature redistribution are clearly not fluctuations, a comparison of the data in Figs. 4-4, 4-5, and 4-19 does indicate some similarities. Note for example channel 5 in Fig. 4-5, which shows an initial small offset followed by a slow response. This is quite similar to an extension of the circled areas in Fig. 4-19 This indicates that the two responses are the result of similar phenomena. During fluctuations, gap changes caused the initial offset responses, and thermal effects were then evident for a few minutes until a second gap change occurred and caused a second offset response, etc. throughout the fluctuation. In contrast, during the exit temperature redistribution a single, small gap change causes a small initial offset response, followed by thermal effects which are evident for a longer period of time (10 to 15 min) as a new, stable value is achieved. 4.2.8 Reactivity Perturbations Further evidence of small core displacement is the small positive reactivity change (N16) which occurred at the time of the region outlet temperature redistribution. Perturbations to core reactivity of similar magnitude were observed during fluctuations prior to installa-l tion of RCDs (Ref. 3). However, during fluctuations these perturba-l tions were cyclic in nature with a 5 to 20 min period. Analysis has indicated that reactivity changes of this order of magnitude can be caused by a displacement of core components so as to reduce the effective diameter of the core, i.e., a compression or tightening of the core so as to close the gaps between regions. The reactivity perturbations occurring during the initiation of the region outlet temperature redistributions correlate with the onset of the changes in the region outlet and gap temperatures. l l 4-36 t
4.3 Scenario of Events 4.3 1 Introduction Region outlet temperature redistributions have been observed during five out of seven different test sequences during Cycles 2 and 3 The region outlet temperature redistributions have all been similar in character to those observed in November and December of 1980 (Ref. 2). While the same regions do not always participate or partici-pate to the same degree, it is always boundary region outlet tempera-tures (particularly in the NW sector of the core) which decrease while inner core region outlet temperatures generally increase somewhat more than expected from an %3% power increase. The scenario of events involved in region outlet temperature redistributions is given below. The scenario is based upon data and analysis from all redistribution events (Cycles 2 and 3). This scenario is basically the same as that presented previously (Ref. 2) but there are some details which are now better understood, in particular, the conclusion that significant region outlet temperature measurement errors induced by Type II flow occur only in the seven NW boundary regions i.e., Regions 20 plus 32-37. 4.3.2 Temperature Redistribution Scenario The data and analyses from each of the region outlet temperature redistributions consistently corroborates small physical displacement i of fuel elements as being the phenomenon causing temperature redistri-butions. This displacement results in a general tightening of the j inner regions of the core in somewhat of an hourglassing manner, 1 l wherein the gaps around outer regions are often increased and gaps between inner regions are generally decreased. Given the many degrees of freedom for movement, the varying j potential combinations of core component dimensional tolerances, the differing lateral pressure fields, etc., this core tightening is, of l 4-37
O s course, not expected to be symmetrical or uniform. In addition, while the redistributions are very similar in nature, they are not expected to be precisely identical. The decreases in core flow resistance, the gap temperature data, and supporting calculations are consistent with the opening and closing of gaps between regions by amounts of nominally 0.10 in. The nuclear channel deviation behavior and the very small (Sld) reactivity changes during the initiation of the temperature redistri-bution are responses to the redistribution of gaps (i.e. , core geometry) and the corresponding redistribution of temperatures. The slight displacement of the fuel elements is consistent with the opening and closing of jaws in the boundary regions of the core of up to approximately the same magnitude as the fuel block motion displacements themselves, i.e., NO.10 in. Regions 1 through 19 are not adversely affected by changes in jaws type crossflow. The temperature redistribution in these interior regions can be explained by decreased gap cooling and a small redistri-bution of flow due to increased bypass ficw. For the seven NW boundary regions through which the region outlet thermocouple assemblies enter the core -(Regions 20 and '32-37), the discrepancy between calculated and measured region outlet temperature increases as the core pressure drop increases due to increasing cool Type II flow effects (References 4 and 5). Changes in cool Type II flow can cause significant changes (error) in measured region outlet temperatures only in the seven NW boundary regions (20 and 32-37). The redistribution of the inter-region gaps can cause changes in Type II flow sufficient to cause changes in measured region outlet temperatures which explain the differences between the expected and measured results for these seven boundary regions. 4-38
n The measured outlet temperatures of the NW boundary regions, Regions 20 and 32-37, are generally significantly cooler than expected aftera}temperatureredistribution. These effects can be explained by a combination of changes in crossflow (jaw flow), increased bypass
.jcooling, and Type II flow effects. To the extent that Type II flow changes are the cause of the decrease in measured outlet temperature, the indicated change is not representative of a real change in the
< ., actual outlet temperature.
The measured outlet temperatures of some of the boundary Regions 21-31 are cooler than expected after a temperature redistribution due to a combination of changes in crossflow (jaw flow) and increased bypass cooling. These effects manifest themselves as real changes in the actual region outlet temperature. The core tightening scenario is summarized in Fig. 4-20, which shows the correlation between various predictions and observations. l W 4-39
~, ,. .
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9
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{ , { [ CORE CORE [
; T (ASYMMETRIC) T
- o o T. R. g s
' 'BEFORE AFTER
+ REDISTRIBUTION REDISTRIBUTION f PREDICTION OBSERVATION REACTIVITY INSERTION OF N14 REACTIVITY INSERTION OF mld DUE TO CORE GEOMETRY CHANGE DECREASE CORE R BY UP TO 5% FOR 1% to 5% DECREASE IN CORE R CORE REDISTRIBUTION OF UP TO 0.10 IN. I INNER REGION T-EXITS INCREASE INNER REGION T-EXITS INCREASED DUE TO DECREASED FLOW AND DE-CREASED GAP. COOLING T-EXITS OF BOUNDARY REGIONS 21-31 BOUNDARY REGION T-EXITS (21-31) DECREASE DUE TO JAWS FLOW AND IN- EXPERIENCED REAL DECREASE CREASED GAP COOLING MEASURED T-EXITS OF REGIONS 20 MEASURED NW BOUNDARY REGION AND 32-37 INDICATED DECREASE DUE T-EXITS INDICATED DECREASE TO JAWS FLOW, INCREASED GAP COOLING AND/OR TYPE II FLOW
" JAWS" FLOW PATHS OPENED IN " JAWS" FLOW EVIDENT IN BOUNDARY REGIONS DUE TO FUEL REGION 35 (ICRD)
BLOCK DISPLACEMENT
, INTERIOR REGION GAPS CLOSE, GAP CHANGES DEDUCED FROM TEMP BOUNDARY REGION GAPS OPEN CHANGES CONSISTENT WITH PRE-(HOURGLASS) DICTION CHANGE IN TRANSVERSE FLOW RATE DEDUCED GAP CHANGES ARE ALONG T/C SLEEVE SUFFICIENT TO CAUSE CHANGE (i.e., TYPE II FLOW) IN TYPE II FLOW NOTE: "T-EXITS" are the temperatures indicated by the y core region outlet thermocouples.
s Figure 4-20. Temperature redistribution scenario
, 4-40
5.0 SAFETY CONSIDERATIONS 5.1 Introduction In Reference 2 a safety evaluation of the region outlet tempera-ture redistribution was submitted to the Nuclear Regulatory Commission. In this report it was noted that, because the outlet temperature redistribution is caused by a mechanism similar to that which previous-ly produced fluctuations, the safety evaluation for a fluctuation event (Ref. 6) remains valid for the outlet temperature redistribution event. It was concluded tnut the reactor could be operated following a redia-tribution, up to 100% power, without increasing the risk to public health and safety. The redistribution safety evaluation in Reference 2 was based upon data obtained during testing below 70% power in November and December 1980. In this section that safety evaluation is reviewed in light of data obtained above 70% power. It is again concluded that the reactor may be operated following region outlet temperature redistributions, up to 100% power, without increasing the risk to public health and safety. 5.2 Safety Evaluation of the Outlet Temperature Redistribution In the presentation and evaluation of data in Section 4 of this report, the results of testing during Cycle 3 were given primary emphasis. For the region outlet temperature redistributions observed during those tests, the magnitude of the changes in most parameters was less than or equal to that observed in the November-December 1980 testing. For the redistributions observed during Cycle 2 testing above 70% power, parameter changes are bracketed by the Cycle 3 redistribu-tions and the November-December 1980 redistributions. In this section, the safety evaluation of Reference 2 is reviewed in light of data obtained above 70% power. 5-1
5.2.1 Wide Range Linear Channel Flux Signals During tne region outlet temperature redistributions, both above and below 70% power, the linear nuclear channel signals have displayed both a small rapid change on some of the detectors and a very small (Nld) reactivity change. Similar behavior was also observed during fluctuations, and the changes that were observed during fluctuations were of about the same magnitude, or larger, than those observed during any of the outlet temperature redistributions. The rapid changes are explained by small displacements in the permanent side reflector resulting in changes in small neutron streaming paths. Not all detectors displayed this behavior, and temperature feedback normally following real reactivity changes was not observed. The small reactivity changes observed continue to be explained by core displacement causing an effectively smaller coro diameter. As discussed in the fluctuation safety analysis report (Ref. 6), the most
- extreme change in reactivity which could result from core motion is to compress the core so as to close all available gaps. The resultant reduction in neutron leakage amounts to only 0.00015 Ak. In FSAR Section 14.2.1 3, the effects of a reactivity charge of 0.006 Ak were evaluated and no damaging effects were found. Therefore, no significant effects from these small reactivity changes due to core displacement are possible. This conclusion is the same as that reached in previous evaluations of the region outlet temperature redistributions.
5.2.2 Control Rod Insertabilicy As discussed in the fluctuation safety analysis report, the maximum misalignment of a control rod channel available if all gaps across the core are combined is 1.5 in. As discussed in Section 3.8.1.2 of the FSAR, rod insertion tests were conducted using 1.6 in. misalignment at the insertion location and 2.5 in. misalignment in the 5-2
rod channel; no appreciable increase in scram times was noted when compared to similar tescs with the core aligned. With the RCDs installed only negligible displacement can occur at the top of the core, and the maximum displacement of the middle of a region relative to its ends is less than 1.5 in. As discussed in Section 4.2.2 of this report, all displacements which have been evaluated for the various redistribution events have been much smaller (by at least an order of magnitude) than 1.5 in. It is again concluded that no predictable misalignment in the core will interfere with the ability of the control rods to be inserted or withdrawn. 5.2 3 Structural considerations 5.2.3.1 Possible Impact Velocities In Section 5.3.1.5.2 of Ref. 6, a 3 in/see maximum fuel element impact velocity associated with fluctuations was calculated. In Section 5 3 3 1 of Ref. 2, using a conservative method and essuming 100% core power conditions, a 2 3 in/see maximum impact velocity associated with the redistribution event was calculated. Data collected during RT-500K testing above 701 power indicate that the 2 3 in/see maximum redistribution impact velocity remains valid for all redistributions which have been experienced. Therefore, it is again concluded that the maximum fuel element impact velocity associated with a region outlet temperature redistribution is less than that associated with fluctuations. 5.2 3 2 core structural Loads In the section above, it was shown that the maximum possible impact velocity, if core motion occurs with the top restrained by the RCDs, is less than that experienced during fluctuations. Moreover, the mode of impact is also similar from the standpoint of causing impact loads. It can, therefore, be concluded that the structural loads 5-3
associated with the redistribution are less than the loads associated with fluctuations, which were shown in Ref. 6 to be small compared to the load capacity of the fuel elements. It should also be noted that the results of the Fuel Surveillance Program for Segment 1 (Ref. 7) showed that no damage had occurred to the fuel elements during Cycle 1. These results further confirm that the fuel elements have the capacity to accommodate the loads associated with fluctuations, which occurred during Cycle 1, and, therefore, to , accommodate the smaller loads associated with occasional redistribu-tions. , 5.2 3 3 RCD Structural Loads In Section 5.3 of Ref. 8, the most highly stressed part of the RCD was found to be the Inconel pin, which is subjected to a maximum load of 1,167 lb. during normal operation. For core motion occurring during a redistribution, as discussed in Section 5 3 3 1 of Raf. 2, the impact load in the pins would be negligible because of their remoteness from the impact zone. Subsequent to the impact, however, about 50% of the pressure force on the displaced column is transferred to the column it leans on, and its associated RCD pin would experience a higher load. If it is assumed that the maximum pin load of 1,167 lb. were to increase by 50%, which is very conservative (since most of the original load comes from a postulated leaning of seven columns in the same direction due to uneven irradiation shrinkage), the maximum stress in the pin would increase from 36,000 psi to 54,000 psi, still below the yield limit of 134,000 psi by a large margin. This conclusion is the same as that reached in Ref. 2 and is not altered by the results of testing above 70% power. 5.2.4 Secondary System l As discussed in Ref. 2, the secondary systems have been eliminated as the cause of fluctuations or the outlet temperature redistribution because steam temperature perturbations lag the primary side (helium) 5-4
temperature perturbations. Further, the helium temperature perturba-l tions are damped in the steam generator, i.e., the steam temperature perturbations are smaller. The principal concern in the secondary ' system during fluctuations was the effect of varying steam temperatures on the fatigue stress limits of the steam generator modules, and for this reason the duration for fluctuations during testing was strictly limited. During some of the temperature redistribution events some steam generator modules experience a single inlet helium temperature decrease of less than 100F, and thereafter respond normally to the power rise. The single temperature decrease, which is believed due to increasing cold bypass flow, is of no consequence to the fatigue stress limits in the steam generator modules. 5.2.5 Bypass Flow Increase After the Outlet Temperature Redistribution For the region outlet temperature redistributions of November and December 1980, the bypass flow fraction was calculated to be N12% before the outlet temperature redistritation events, increasing to S14% after the temperature redistribution. The effect of this 2% increase was to increase maximum core fuel temperatures by about 200F and average core fuel temperatures by about SOF. These small increases in fuel temperature have a negligible impact upon fuel performance. As indicated in Section 4 of this report, the change in core resistance associated with the redistribution events above 70% power in Cycle 3 has been less than that experienced in November and December 1980. In addition, the magnitude of the decrease in boundary region outlet temperatures following an N3% power increase is less, and fewer gap thermocouples indicate any significant changes in gap conditions. Therefore, it is concluded that the increase in core bypass flow
' Core bypass flow fraction is the portion of core cavity flow which does not pass over the region exit thermocouples. It includes the flow in the vertical gaps between fuel element columns, between side reflector columns, and between the side reflector and the core barrel.
5-5
fraction experienced during the November and December 1980 redistribu-tions bounds any increases experienced during subsequent redistribution events. 5.2.6 Accident Analyses FSV accident analyses (FSAR Chapter 14) have been reviewed to determine if any reevaluation is required for plant operation after the temperature redistribution. It was determined that the localized initial conditions created by the outlet temperature 'edistribution 4 would not affect the accident consequences since the accidents are initiated at Technical Specification limits and operation before, during, and after the temperature redistribution is within those limits. Safety systems such as the plant protective systems, the reserve shutdown system, and the liner cooling system which may be required to effcet a safe shutdown are neither associated with nor impaired by the temperature redistribution. It is concluded that no reevaluation of the FSAR accident analyses is required as a result of the observed outlet temperature redistribu-tions. 5.3 conclusions Evaluations of the region outlet temperature redistributions have indicated that these events present no unreviewed safety considera-tions. The small reactivity effects of the redistribution are much less than those evaluated in the FSAR. Control rod insertibility is not affected by the redistribution. Structural loads on core components, including the RCDs, are small compared to the capacity of the components. The impact of the redistributions on secondary system components is insignificant. Fuel temperature changes associated with changes in the core bypass flow fraction are small. As was shown in Section 5.4 of Ref. 2, changes in Type II flow during a redistribution, while producing apparent changes in the region outlet helium 5-6
temperature, produce no change in peak fuel temperature for the affected region. Changes in jaws crossflow during a redistribution, on the other hand, produce an actual decrease in region outlet helium temperature and in peak fuel temperature for the affected region. Based upon evaluations of the region outlet temperature redistri-bution events, it is concluded that, while the redistribution is a curious phenomenon, it in itself entails no unreviewed safety questions and is at most a perturbation on the discrepancies between measured and calculated region outlet temperature which exist prior to the redistribution. It is concluded that with proper operating procedures, the reactor may be operated following region outlet temperature redistributions, up to full rated power, without increasing risk to public health and safety. A method for operating the reactor which accounts for the effects of the redistributions and the discrepancies between measured and calculated region outlet temperature has been developed. This operating method is discussed briefly in the following section. 5-7
6.0 LONG TERM OPERATION Extensive analyses of measured and computed region peaking factor (RPF) distributions during cycles 1, 2 and 3 have shown discrepancies (>10%) between the measured and calculated RPFs for the NW boundary regions (Ref. 5). The discrepancies in the NW boundary regions typically indicate a measured region outlet temperature lower than calculated, with the discrepancies increasing with core pressure drop. Other regions exhibit smaller discrepancies which are essentially independent of core pressure drop. Based upon extensive investigation, the core physics calculated RPFs are discounted as a major source of the discrepancy. The major cause of the RPF discrepancies in the NW boundary regions is cool Type II flow which results in an incorrect measurement of the region outlet temperature. The driving potential for Type II flow increases with increasing core pressure drop. This is consistent with the observed increase in discrepancy with core pressure drop. Only the seven NW boundary regions are susceptible to significant cool Type II flow induced outlet temperature measurement errors. Crossflow (jaws flow) may also contribute to the RPF discrepancies in the NW boundary regions, but does not contribute to a
- egion outlet temperature measurement error.
Analyses performed in support of the investigation of the core temperature fluctuations (Refs. 3 and 4), and the calculations performed during and in support of testing from 40 to 100% power per RT-500K, as well as analyses of the data from these tests, corroborate E.e above observations. The measured temperature changes which occur as a result of region outlet temperature redistribution events are real changes in the actual outlet temperatures with the exception of those which occur in the seven NW boundary regions (Regions 20 and 32-37). For regions other than the seven NW boundary regions, the outlet temperature changes observed during redistribution events are due to changes in gap cooling, increased bypass cooling, a slight redistribution of flow due to increased bypass flow, and changes ia crossflow (jaw flow). All of these phenomena cause real changes in the actual region outlet temperatures, which can easily be accommodated by 6- 1
r o , orifice valve adjustments as are made routinely following load changes. Thus, for regions other than the seven NW boundary regions, outlet temperature redistributions are of no unusual significance. On the other hand, the observed changes in the neasured region , outlet temperatures of the seven NW boundary regions only partially reflect real changes in the actual outlet temperatures. The changes in measured outlet temperature are partially due to changes in crossflow and/or increased bypass flow cooling. To this extent, the observed changes are real. However, these seven regions are also susceptible to cool Type II flow effects. The observed changes in outlet temperature for these regions can be totally or in part due to changes in cool Type II flow, which cause a temperature measurement error. In fact, the measured exit temperatures for these regions are potentially in error due to cool Type II flow even prior to a redistribution event. In this sense, the redistribution is only a perturbation on the temperature measurement error. Thus, special operating guidelines or procedures are to be provided for these regions to insure compliance with both the letter and intent of the technical specifications related to region outlet temperatures. Operation of the reactor with errors in measured region exit temperatures for selected regions was addressed in test procedure RT-500K (Ref. 1). Throughout the various periods of testing from 40% to 100% of rated power per RT-500K, the use of comparison regions was demonstrated to be an effective manner in which to deal with potential region outlet temperature measurement errors. Based upon the operating experience obtained during RT-500K testing, it is planned for normal operation up to 100% power to operate each of the seven regions susceptible to significant outlet temperature measurement error, i.e. Regions 20 and 32-37, via comparison regions starting at a relatively low power level. This eliminates the need for frequent cocouter calculations, such as were done in RT-500K, to identify which of the seven regions requires the use of a comparison region or when such operation is required. 6-2
Comparison regions typically have calculated power densities of magnitude and shape as a function of control rod configuration similar to those of the corresponding region having an outlet temperature measurement error. Knowing the relative power densities (calculated) of the region susceptible to outlet temperature measurement error and that of the corresponding comparison region, the region susceptible to measurement error can be orificed to have the same (or lower) power-to-flow ratio as the comparison region. Thus, the core inlet orifice valves of Regions 20 and 32-37 can be adjusted based upon the orifice valve positions of their respective comparison regions such that their actual outlet gas temperatures (and power to flow ratios) are less than or equal to those of the corresponding comparison regions. Regions affected similarly by changes in the regulating rod position (Region 1 control rod pair) are preferred as comparison regions. The regions to be selected for use as comparison regions are not susceptible to significant cool Type II flow effects. Accordingly, the indicated region outlet temperatures for these regions are considered to be reliable. Operational flexibility is provided by the fact that more than one region can be used as a comparison region for each of Regions 20 and 32-37. Any one of the comparison regions may be selected for use within the range of shim bank configurations for which it is best suited. Furthermore, the relative power densities of the various core regions involved can be pre-calculated, and the reactor can be operated based upon these values for extended periods of time (e.g. several weeks or months). Periodically (e.g. monthly) core physics computer calculations can be performed to revise or update the relative power densities used during this mode of operation. Inose regions which are not susceptible to significant cool Type II flow-induced outlet temperature measurement errors will continue to be operated in a normal manner based upon their measured outlet temperatures. However, to provide margin in addition to what is 6-3
currently provided in the Technical Specifications 5 and therefore allow for the uncertainty associated with operation based in part on physics computer calculations, it is intended that the region outlet temperature mismatch limits will be restricted to those of Figure 6-1. This figure is based upon " Figure C" of RT-500K. Figure C was an optional, more restrictive margin than the mismatch limits required by either LCO 4.1 7 or by the test procedure and was used as a guideline during testing and not as a restriction. However, experience gained during testing showed that the reactor can be operated within the limits of this figure. Therefore, it is now planned that the reactor be operated based upon the more restrictive limits of this figure with corrective action requirements similar to those currently in Technical Specification LCO 4.1.7. The above procedures for plant operation are consistent with the operating experience gained during testing per RT-500K and are consistent with meeting the intent of the long term thermal design analysis reported in Section 3.6 of the FSAR. On this basis, Technical Specification LCO 4.1 7 will be revised, and new Surveillance Require-ments will be developed for long-term reactor operation above 70% power. l I r l 5 Namely, LCO 4.1.7 " Core Inlet Orifice Valves, Limiting Conditions for Operation" l 6-4
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7.0 REFERENCES
- 1. Request for Test RT-500K, March 14, 1981.
- 2. K. E. Asmussen, et. al. , " Testing at Fort St. Vrain After Installation of Region Constraint Devices," GA-C16277, February 1981 - PSC Submittal to NRC P-81312, December 10, 1981.
3 R. Hackney and J. Saeger, " Investigations of t'he Fort St. Vrain Cycle 2 Fluctuations Through October 20, 1979," GA-C15767, March 1980 - PSC Submittal to NRC P-80417, December 2, 1980.
- 4. G. J. Malek, et. al. , " Investigations of the Fort St. Vrain Reactor Fluctuations," GA-C15469, September 1979 - PSC Submittal to NRC P-79287, November 28, 1979.
- 5. "FSV RECA Verification," PSC Submittal to NRC P-81303, December 1, 1981.
- 6. " Safety Evaluation - Reactor Outlet Temperature Fluctuations," PSC Submittal to NRC P-78137, August 11, 1978.
- 7. C. M. Miller and J. J. Saurwein, " Nondestructive Examination of 51 Fuel and Reflector Elements from Fort St. Vrain Core Segment 1,"
GA-A16000, December 1980 - PSC Submittal to NRC P-81254, November 16, 1981. 6, "SAR for Core Region Constraint Devices," PSC Submittal to NRC P-79068, March 23, 1979. 7-1}}