ML20039D547
| ML20039D547 | |
| Person / Time | |
|---|---|
| Site: | Fort Saint Vrain  |
| Issue date: | 02/28/1981 |
| From: | Asmussen K, Hackney M, Kapernick R GENERAL ATOMICS (FORMERLY GA TECHNOLOGIES, INC./GENER |
| To: | |
| Shared Package | |
| ML20039D544 | List: |
| References | |
| GA-C16277, NUDOCS 8201050179 | |
| Download: ML20039D547 (62) | |
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- 1 e , *; TESTP'9 AT FORT ST. VRAIN AFTER e ! INSTALLATION OF REGION CONSTRAINT DEVICES s v y (j hy Q K. E. ASMUSSEN, M. R. MACKNEY, R. J. KAPERNICK, 4 and J. C.' SAEGER ?
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f ti GA C16277 TESTING AT FORT ST. VRAIN AFTER INSTALLATION OF REGION CONSTRAINT DEVICES - 1
;) - ; by $ K. E ASMUSSEN, M. R. HACKNEY, R. J. KAPERNICK, ? and J. C. SAEGER I
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,;! Prepared under
. Purchase Order N-3399 for the Public Service Company of Colorado
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? GENERAL ATOMIC PROJECT 1920 ;
i FEBRUARY 1981
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l l i I CONTENTS
- 1. INTRODUCTION . . . .. . . . . . .... ... .... . ... .. 1-1 i 2.
SUMMARY
2-1
- 3. TESTING ..... . . . . . ..... . . . . . . . . .. . . .. '3-1 3.1. Steady-State Testing . . . .. ... .. .. . . ... .. 3-l' 3.2. . Fluctuation Testing . . ... ........ .... .. 3-2 l 3.2.1. Objectives . . . ... ... .. ..- . . . .. .. 3-2 i
l 3.2.2. Test Strategy. . . . . . . ..... . . . .. .. 3-3 3.2.3. Test Procedure . ........... .... . . 3-3
-4. DATA AND INTERPRETATION ... .......... . . .-. .. . 4-1 4.1. Observations on Data . . .. ... ...... . . . .. . 4-1 4.1.1. . Measured Region Exit Temperature Redistributions ... .. . . - . ... .... . . 4-1 4.1.2. Other Effects of'the Redistribution. . . ... . . 4-1 4.1.3. Temperature Redistribution Versus. Fluctuation .. 4-5 4.1.4. Sucmary of Data Observations . .... . . . ... . 4-15 4 . 2 ., Analysis and Interpretations . .. ... .. . . . ... . . 4-15 4.2.1. Expected Versus Measured Temperature Distribution . . ... . ...... ... .. .. 4-15 4.2.2. Gap Temperature Changes .... .. . . ... . . 4-18 4.2.3. Core Resistance Changes .. .... . . ... . . 4-22 4.2.4. Region 35 Jaws calculation . . ... . . . .. . . 4-22 4.2.5. Type II Flow Segments .. .. ... . .. . . . . 4-23 4.2.6. Nuclear Channel Deviations . ... .. ... . .. 4-24
< 4.2.7. Reactivity Perturbations . .. ... . .. .. . . 4-30 ) 4.3. SCENARIO OF EVENTS . . . . .. .. ... . .. . .. . . . 4-30 i i 5. SAFETY CONSIDERATIONS . . . ..... .. .. .. . ... . . 5-1 5.1. Summary . . . . . . . . ... . ... .... . . . . .. 5-1.. l l 5.2. Comparison of Fluctuation and Outlet Temperature l Redistribution Characteristics . . . .. . .. .. . . .. 5-2 e
~
iii
- .3. Safety Evaluation of the Temperature Shift Event . . . .. 5-3 5.3.1. Wide Range Linear Channel Flux Signals . . .. .. 5-3 5.3.2. Control Rod Insertability . . . . .. . .. .. . 5-3 5.3.3. Structural Considerations . . . ... . .. . . . 5-4 5.3.4. Secondary System . .. .............. 5-5 5.3.5. Bypass Flow Increase After the Outlet Temperature Distribution . . . . ..... .. .. 5-6 5.3.6. Accident Analyses ...... . . ... .. . .. 5-6 5.4 Long-Term Operation With the Temperature Redistribution . . . . . . ........... . . . . . 5-6
- 6. REFERENCES . . ... . . . . . . ........... . . ... 6-1 FIGURES 3-1. Fluctuation test strategy . . ..... . . . . ... .. . . 3-4 3-2. Core pressure drop as a function of power (November 1980 testing sequence) . . . . . . ..... . . . .. . . . . . . 3-7 3-3. Core pressure drop as a function of power (December 1980 testing sequence) . . . . . . . . .............. 3-10 4-1. Region outlet temperatures (regions 1 through 19),
November 14, 1980 . . . . . . ............ . ... 4-2 4-2. Region outlet temperatures (regions 20 through 37), November 14, 1980 . . . . . . . . ......... ~ .. . . 4-3 4-3. Nuclear channel deviations: channels 3, 4 and 5 ... . . . 4-4 4-4. Calibration tube thermocouple locations . . . . .. . . . .. 4-6 1 4-5. Cap T/Cs 3, 4, and 5 for November 14, 1980 .. . . . . . .. 4-7 4-6. Gap T/Cs 7, 8, 9, and 10 for November 14, 1980 .. . . . . . 4-8 4-7. Core coolant flow resistance: November 14, 1980 and December 13, 1980 . . . . . . ... .... . . . . . . . . . 4-9 4-8. Region 5 ICRD temperatures and gap T/Cs . . . . .. . . . . . 4-10 4-9. Region 35 ICRD temperatures and gap T/Cs . . . . . . . . . . 4-11 4-10. Representative nuclear channel deviations during fluctuations . . . . . . . . . . .. . . . . .. . . . ... 4-12 4-11. Temperature redistribution, T/Cs 11 and 13 . . . . . . . . . 4-13 4-12. Temperature fluctuation, T/Cs 11 and 13 . . . . . . . . .. . 4-14 4-13. Exr -ted versus measured region exit temperatures (interi r regions) . . . . . . . . ....... . . . . . . 4-16 iv
p . . f' f FIGURES (Continued) 4-14. Expected versus measured region exit temperatures i: (boundary. regions) ... .. . . . . . . . . . . ... . . . . . . . 4-17. i 4-15. -Calculated gap redistributions.for November' 14, 1980 . . . . 4-20 4 4-16. Gap outlet temperature calculated versus measured during temperature redistribution .. . ... . '. . . . . . . . . '. . 4-21
- g 4-17. Type 11 flow calculations . ........ . . .. . . ... . 4-25 '
? 4-18. . Nuclear channel- response during temperature
- redistribution .. . . . . .. ..... . . . . . . . . .. . . 4-26 3
- 4-19. Nuclear channel deviations during temperature redistribution, November 14, 1980 . . . . . . . . . . . . . . 4-27 4-20. Nuclear channel deviations during temperature
- redistribution, November. 14, 1980 . . . . . . . . . . . . . . 4-28
, 4-21. Representative nuclear' channel deviations during cycle 2 fluctuation . . . . ........... . . . . . . 4-29 4-22. Temperature redistribution scenario . . . . . . .. . . . .. 4-32 i
- 5-1. Region 36 peak fuel temperature af ter the region exit temperature redistribution assuming Type Il flow or
. Crossflow . . . .... . . ....... . . . . . . . . . . 5-10
} TABLES l 3-1. . Sequence of events (November 1980) . . . . . . . . .. . .. . 3-5 l 1 3-2. Sequence of events (December 1980) . . . . . . . . . .. . . . 3 ! [ 3-3. Temperature redistribution initiating conditions . . . . . . . '3-11 4- 1. - Calculated gap changes .
. . .. . . . . . . . .. . . . . . .- 4-19 i
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- 1. INTRODUCTION During the initial rise-to-power progran of the Fort St. Vrain reactor in October 1977, while approaching 60% power, tenperature fluctuations were observed in the primary coolant circuit at the outlets of individual core regions and the inlets to steam generator nodules. A comprehensive progran of investigation into the nature and cause of the tenperature fluctuations was initiated inmediately. The fluctuation investigations led to the design and fabrication of region constraint devices (RCDs) as a solution to the problem. These mechanical links were installed on the top of the core in November, 1979. They were installed at locations where three regions inter-sect and are designed to provide inter-region linking to stabilize the gaps between regions at the top of the core to near noninal values.
Steady-state testing was perforced during initial operation following installation of the RCDs to verify that the overall core perfornance was unaffected by the presence of the RCDs. Testing to evaluate the success of RCDs as a solutien to the temperature fluctuations was performed in November and December of 1980. The results of these tests are sucnarized in this report. 1-1
- 2.
SUMMARY
The steady-state testing conducted after the installation of RCDs confirmed that the installation of- RCDs had a ninical effect on the overall core perfornance. Testing wherein attenpts were nade to induce fluctuations, after installation of region constraint devices, was first conducted on Novenber 12 through November 15, 1980. Power levels fron 40% to 70% were surveyed at two sets of core orifice positions (core flow resistances) with a naxinun core pressure drop of 4.1 psid. No fluctuations were observed, even in operating regimes considered unstable prior to the installation of the region constraint devices. However, during the transition fron 55% to 59% power at the higher core resistance, a region exit tenperature redistribu-tion was observed. This redistribution of region outlet tenperatures resulted in several boundary region outlet temperatures, particularly in the NW sector of the core, decreasing while inner core region outlet ten-peratures generally increased somewhat more than expected fron the power change. This redistribution of temperatur*: generally persisted throughout the remainder of the test. Stean generator hellun inlet tenperatures also recorded a change in distribution. Due to the loss of sone of the stean generator data as well as to a desire to confira repeatability of the phenomenon, the test was rerun along the higher core resistance line on Decenber 12 through 14, 1980. Power levels fron 40% to 70% were surveyed up to a naxinum core pressure drop of 4.2 psid. A region outlet tenperature redistribution almost identical to that observed in Novenber was encountered under essentially the sane conditions. Again no fluctuations were observed. 2-1
The region exit temperature redistributions are the result of snall in-core displacements. These displacements are similar in nature to the initial notion which occurred during fluctuations. Houever, these displace-cents are not cyclic. These snall displacenents cause changes in cap dis-tribution (between regions), crossflow, and in the acount of transverse heliun flow along the sleeve (s) surrounding the region exit tenperature thernocouples. These observations are consistent with a " tightening" of the core, wherein the gaps between the outer regions and the pernanent side reflector generally are increased and inner region gaps are generally decreased. Testing below 70% power has been completed. The testing has denonstrated that region constraint devices are successful at preventing fluctuations for power levels up to 70% and core pressure drops up to 4.2 psid. Extrapolation of available data indicates that the plant can be operated in a stable canner above 70% power without increased risk to the health and safety of the pehlic. 2-2
- 3. TESTING After installation o' the region constraint devices (RCDs), both steady-state and fluctuation tests were perforned. The objectives of these tests were to evaluate the effect of the RCDs on steady-state core perfornance and on the fluctuation threshold.
3.1. STEADY-STATE TESTING Three steady-state tests were perforced during initial operati m with RCDs installed to verify that the overall steady-state core perfornance uas unaffected by the presence of the RCDs. These tests provided data on the region peaking factor (RPF) distribution, tenperature. profiles along the region exit thernocouple (T/C) calibration tubes and selected orifice calibrations. RPF distributions were neasured for power levels fron 5% to 69% power. These neasured data were compared to the computed RPF distribution. Sini-larly the measured RPF distributions obtained prior to installation of the RCDs were compared with computed RPFs. (The conputed and neasured RPFs were compared rather than the RPFs themselves since all neasured data did not have the sane orifice configuration and/or control rod positions before and after installation of the RCDs.) These data cere used to evaluate the effect of RCDs on the RPF distribution. Tenperature profiles were obtained by noving a T/C inside the cali-bration tubes of the core outlet thernocouple assenblies in approxinately one-inch increnents and recording the temperature and the distance the T/C was inserted. The core conditions such as power, flow, orifice positions, core pressure drop, core inlat and outlet tenperature were very nearly the sane for neasurenents nade both before and after installation of the RCDs. Temperature traverses uere measured in each of the seven (7) thernocouple 3-1
assemblies at ~70% power. These traverses were compared to those neasured before installation of the RCDs. Orifice calibration data were measured for the orifice valves in regions 10, 28 and 34. These regions represent a refueled interior region (10), a refueled boundary region (28) and a boundary region (34) which had not been refueled. Results of the analyses of these steady state test data showed no significant measurable effects on core performance. Comparisons of these data with data obtaine.1 prior to installation of the RCDs are sunnarized below.
- 1. Core reactivity is unaffected.
?. The tenperature profiles along the calibration tubes of'the thernocouple assemblies are basically the sane, with sone evidence of gap distribution changes.
- 3. Evidence of Type II flow along the calibration tube penetrations still exists.
- 4. There is no indication that the orifice characteristics have changed.
From the results of these studies it was concluded that the core is perforning as expected with no significant neasurable changes as a result of installation of the RCDs. 3.2. FLUCTUATION TESTIf3 3.2.1. Objectives RT-500 is a fluctuation test for evaluating the fluctuation threshold as a function of core pressure drop versus power (or flow). The test was 3-2
originally performed in Novenber 1978 during cycle 1 operation as a part of the fluctuation test progran. While this test has undergone nunerous revi-sions as a result of testing experience and for compatibility with cycle 2 testing, the basic test philosophy has remained unchanged. RT-500H (Revi-sion H) was the test perforned in November and December 1980. The purpose of this test was to deconstrate that the installation of region constraint devices solved the fluctuation problen or to deternine what inpact their installation had on the fluctuation threshold. 3.2.2. Test Strategy RT-500H was a test wherein attenpts were nade to initiate rluctuations, thereby deternining the effect of the installation of RCDs on the fluctua-tion threshold. The core was to be orificed for a core pressure drop of ~1.6 psid at 40% power with a core flow resistance
- of ~45. The power would then be increased in steps of 3% at ~3%/ min (waiting at least 2 hours between steps) to 70% power. This procedure was planned for three values of resistance. A schenatic of this test strategy is shown in Fig. 3-1. The details of the test (RT-500H) are given in Ref. 2.
3.2.3. Test Procedure RT-500H was initially perforned on November 12 to 15, 1980. The sequence of events for that test are given in Table 3-1. A sunnary of the test procedure is given in the following paragraphs.
*The core flow resistance parancter is defined as:
13 AP - P R = 2 x 10 - T - (n)1.85 where AP = neasured core pressure drop (psid), P = PCRV pressure (psia), T = average circulator inlet tenperature (*R), and n = total circulator flow rate (Ihn/hr). 3-3
R = CORE FLOU RESISTANCE R3>R2>Ry 3 R
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{ 1 l 40% N70% CORE POWER Fig. 3-1. Fluctuation test strategy 3 '4
TABLE 3-1 SEQUENCE OF EVENTS (Novenber 1980) 2 Power Flow Core AP Date (Time) % % (psi) Resistance Concents STEADY-STATE CONDITIONS PRIOR TO PORER INCPIASE 11/12/80 (1043) 40 54 1.6 45 Starting conditions 11/12/80 (1607) 43 60 1.8 44 11/12/80 (1848) 46 64 2.1 44 11/12/80 (2250) 50 66 2.2 44 11/13/80 (0145) 53 71 2.5 43 11/13/80 (0505) 56 74 2.7 43 11/13/80 (0712) 59 78 2.9 43 11/13/80 (0942) 62 79 3.1 44 11/13/80 (1234) 65 82 3.3 43-11/13/80 (2130) 70 86 3.6 43 After final power rise at R = 43 REDUCE POWER AND REORIFICE FOR HIGHER RESISTANCE 11/14/80 (0240) 41 55 2.1 57 Starting conditions 11/14/80 (0425) 44 58 2.3 57 11/14/80 (0650) 47 63 2.6 56 11/14/80 (0900) 53 67 3.0 56 11/14/80 (1108) 55 70 3.2 56 11/14/80 (1154) 59 75 3.4 53 After power rise that initiated terperature redistribution REDUCE POWER TO REPEAT PREVIOUS POWER RISE l 3.2 11/14/80 (1805) 56 72 53 11/14/80 (2100) 59 75 3.5 54 11/15/80 (0352) 63 78 3.6 52 11/15/80 (0530) 66 82 3.9 52 11/15/80 (0801) 69 84 4.1 52 End of RT-500n REDUCE POWER TO ~60% AND PIORIFICE FOR NORMAL OPERATION 3-5
Initial conditions of 40% power, a core pressure drop of 1.6 psid and a resistance of ~45 were established. Dua power was increased in steps of ~3% at ~3%/ min, waiting at least 2 hours between steps, to 70% at a core pras-sure drop of 3.6 psid, with no unusual behavior. The power was then reduced to 40% and the core was reorificed for a. core pressure drop of 2.I' paid with a resistance of ~57, maintaining essentially the same relative distribution of flow through the refueling regions. The power was again increased, in steps, towacd 70%. After the power increase to 59%, a region exit tempera-ture redistribution was noted. Af ter nonitoring core behavior at 59% power for ~3 hours the power was reduced to ~56%. Again af ter monitoring core
- performance for ~3 hours and observing no unusual behavior, the power rise to ~59% was repeated. Power increases were continued to 70% power and a core pressure drop of 4.1 psid, with no unusual behavior. The region exit tennerature distribution recained stable with the exception of a partial restoration of the pre-event temperatures in two regions. At this point it was deciued to terninate testing pending detailed evaluation of the data obtained during the tenperature redistribution. Power was then reduced to
~60% and the core was reorificed for normal operation. During this reori-ficing, beginning at a core AP of ~2.8 psid, the region exit temperatures returned to its pre-event distribution. The core pressure drop as a function of power, obtained during this test, is shown in Fig. 3-2.
After completion of the above testing it was learned that some of the steam generator (SG) data tape was not readabic. It was subs.'equently decided to repeat the second resistance line of_the above test in order to obtain SG data and to determine whether the observed tenperature redistribu-tion was repeatable. In addition, the power was to be reduced fron 70% to 40% in steps to obtain note data on the return of the tenperatures to their pre-event distribution. The sequence of events for this second test, perforced on Decenber 12 to 14, 1980 is given in Table 3-2. A sunnary of the test procedure is given in the following paragraph. 3-6
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TABLE 3-2 SEQUENCE OF EVENTS (December 1980) Power Flow Core AP Date (Time) % % (psi) Resistance Connents STEADY-STATE CONDITIONS PRIOR TO POWER INCREASE AND/OR POWER DECREASE 12/12/80 (1510) 41 53 1.9 56 Starting conditions 12/12/80 (1626) 46 58 2.3 55 12/13/80 (0646) 52 66 2.8 55 12/13/80 (0940) 55 71 3.1 54 12/13/80 (1108) 58 73 3.3 54 12/13/80 (2240) 61 76 3.5 53 After power rise that initiated temperature redistribution 12/14/80 (0204) 62 75 3.5 33 12/14/80 (0504) 67 81 4.0 53 12/14/80 (0602) 69 83 4.2 53 After final power rise 12/14/80 (0955) 57 71 3.2 53 12/14/80 (1104) 55 69 3.0 53 12/14/80 (1210) 52 67 2.8 53 12/14/80 (1339) 49 64 2.6 53 12/14/80 (1511) 46 62 't . 4 54 12/14/80 (1704) 43 56 2.1 54 12/14/80 (1742) 40 53 1.8 54 End of RT-500H 3-8
Initial conditions of ~40% power, a core pressure drop of 1.9 psi and a resistance of ~56 were established. Again the power was increased toward 70% in steps of ~3% at 3%/nin. During the power rise from ~58% the region exit temperature redistribution again occurred and was renarkably sinilar to that observed during tha tlovember 1980 testing. Power increases were con-tinued to 69% and a core AP of 4.2 psid with no unusual behavior. The power was then reduced to 60% followed by a reduction to ~40% in steps of ~3% at 3%/ min. The region exit temperatures first showed evidence of returning to their pre-event distributions following the power reduction from ~52% (core 6P ~2.8 psid) to ~49% (core AP ~2.6 psid). The core was then reorificed at 40% power for nornal operation. The core pressure drop as a function of power, obtained during this test, is shown in Fig. 3-3. To illustrate the remarkable sinilarity of the November and Deceneer region exit temperature redietributions, the initiating conditions for the two events are compared in Table 3-3. Note that in both cases the redistri-butions were initiated by transient peak core pressure drops of 3.8 psid and the resulting steady state core pressure drop following the initiating load increase is 3.4 psid. 3-9
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TABLE 3-3 TEMPERATURE REDISTRIBUTION INITIATING CONDITIONS November Decenber Power increase, % 55 + 59 58 + 61 Flow, % 70 + 75 73 + 76 Core AP, psid Steady state 3.2 + 3.4 3.3 + 3.4 Transient peak 3.8 3.8 3-11
- 4. DATA AND INTERPRETATION 4.1. OBSERVATIONS ON DATA 4.1.1. Measured Region Exit Temperature Redistributions The region outlet temperatures were continuously measured during both the November 14, 1980 and December 13, 1980 temperature redistributions.
The temperatures before and af ter the November redistribution are shown in Figs. 4-1 and 4-2. These are typical behavior of region exit temperatures for both redistributions. Note that in general the inner regions (1 through-
- 19) increascd in temperature, while the boundary regions generally decreased. During most power increases one would expect all regions to increase. The temperature redistribution is essentially the same for the November 14, 1980 and the December 13, 1980 events.
4.1.2. Other Effects of the Redistribution 4.1.2.1. Nuclear Channel Deviations.* Several of the nuclear channel devi-ations exhibit small abrupt changes at the time of the redistribution fol-- lowed by a gradual change-(see Fig. 4-3). While the deviations exhibited changes, they were not cyclic; rather they simply stabilized at new levels. 4.1.2.2. Core Reactivity Perturbations. Careful examination of the nuclear channel signals during the initiation of the region exit temperature redis-tribution reveals the existence of a small reactivity insertion of about +1d (~0.00007 ap) not due to rod control motion.
- Nuclear channel deviations are defined as (Xi - X) where Xi is the signal from nuclear channel i and X is the average of the six channels.
During testing without RCDs installed, nuclear channel deviations proved to be highly reliable and sensitive indications of fluctuations. 4-1
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4.1.2.3. Gap' Temperature Change. There are twenty-six thermocouples, which are traversable across the core, installed in the calibration tubes at the ' core exit. Seventeen of these thermocouples are located at gaps between selected core support blocks (Fig. 4-4). These thermocouples indicate tem-perature changes during the temperature redistribution. The eignals of rep-resentative gap tr ermocouples are shown in Figs. 4-5 and 4-6 for the Novem-ber 14, 1980 event. The temperatures at gaps internal to the core clearly-increased while those adjacent to :he outer ring of regions generally decreased. 4.1.2.4. Core Resistance Change. . The core flow resistance parameter (see S . tion 3.2.2) shows a sharp decrease at the occurrence of the temperature redistribution for both the November 14, 1980 and December 13, 1980 events (see Fig. 4-7). The resistance dropped 4% to 5% in both cases. 4.1.2.5. Instrumented Control Rod Drive (ICRD) Thermocouple Measurements. The thermocouples in regions 5 and 35 control rod channels respond during the temperature redistribution. Region 5 responses are shown in Fig. 4-8, and those of region 35 in Fig. 4-9. Note the smooth behavior.of region 5 (except for signal noise) compared to the abrupt nearly step changes of region 35, clearly indicating a different behavior. The reasons for this difference are discussed later. 4.1.3. Temperature Redistribution Versus Fluctuation The temperature redistribution is clearly not a fluctuation. The parameters measured during the temperature redistribution do not fluctuate l as they did during the fluctuations experienced prior to installation of RCDs. This can be-illustrated by comparing the measured reactor parameters. For example, compare the displacement in the nuclear channel deviations of l the redistribution (Fig. 4-3) to those during fluctuations'(Fig. 4-10). Also compare region outlet temperatures and gap temperatures during and
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4.1.4. Sunnary of Data Observations The data consistently indicate the region exit temperatute redistribu-tion to be an event that sinultaneously affects region outlet temperatures, gap tenperatures, core resistance, and neutron flux to ex-core detectors. For fast responding instrunents (e.g., nuclear channels and ICRD control channel thernocouples) there is an abrupt change in measured paranetera gen-erally followed by a slower change, suggestive of thernal phenonena, until a new steady condition is reached. 4.2. ANALYSIS AND INTERPRETATIONS The previous discussions were concerned primarily with the presentation of some significant data and the comparisons to data of a typical previous fluctuation. In this section the observed neasurenents are explained and conpared to calculations and nechanisns which tend to reproduce the observed data. 4.2.1. Expected Versus Measured Tenperature Distribution In order to help understand the thermal behavior during the temperature redistributions, the expected temperature chatter were determined for each region on the basis of both calculations and pa.t dstory of performance during power rises. The expected values of the tenperature changes are for a load increase taken by withdrawal of the regulating control rod during nonfluctuating operation with no tenperature redistributions. In Figs. 4-13 and 4-14 the expected behavior is conpared with the ceasured behavior during the outlet tenperature redistribution of Decent er 13, 1980. Clearly the changes in tenperature for the inner regions (1 through 19) are larger than the expected value (see Fig. 4-13). On the other hand, the ceasured changes in the outer regions (20 through 37) are with one exception less than expected (see Fig. 4-14), or in fact decreased. 4-15
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l 4.2.2. Gap Temperature Changes The gap tenperatures are observed to change during the region outlet tenperature redistribution. The tenperature change can be explained (at least in part) by gap size changes which affect the heat transfer and flow (Ref. 3). A thernal flow model of the gap between regions had been previ-ous./ constructed (Ref. 3) and had been used to nake calculations of gap tenperature and flow responses. A series of steady state analyses were perforned for the reactor conditions before and af ter the temperature redistributi- as a function of gap size. The tenperatures before and after the tenperature redistribution were then tabulated from the neasured data. The tenperature before the redistribution indicates the initial size of the gap, and fron the change in tenperature the final gap size may be deternined as well. The results are shown in Table 4-1 and again in Fig. 4-15 on the core nap. The gaps interior to the core are indicated to have closed by 0.060 to 0.090 in. The gaps adjacent to the outer ring regions have noninally opened by 0.035 to 0.100 in. This result indicates a general redistribution of the gaps in the core. Additional transient calculations have been perforned which further support the gap redistribution theory. With the assunption that an initial gap of 0.180 in, is step increased to 0.240 in, when the redistribution occurs, the gap tenperatures through the power and flow transient of Noven-ber 14, 1980 were calculated. The results are shown in Fig. 4-16 (solid line) as conpared to the actual measured data (broken line). Note the excellent agreenent. Even better agreement could be obtained with a nore precise choice of initial and final gap size. 4-18
TABLE 4-1 CALCUIATED GAP CHANGES (November 14, 1980) T/C Tt T2 gi g2 82 - 81 3 1310 1380 0.090 ~0.0 -0.090 4 1130 1079 3.180 0.235 +0.055 7 1330 1390 0.090 ~0.0 -0.090 s 8 1192 1127 0.125 0.200 +0.075 11 1320 1375 0.075 ~0.0 -0.075 15 1310 1380 0.090 ~0.0 -0.090 18 1345 1400 0.075 ~0.0 -0.075 19 1275 1325 0.060 0.0 -0.060 20 1275 1325 0.060 0.0 -0.060 21 1280 1310 0.090 0.0 -0.060 22 1130 1100 0.180 0.220 +0.040 23 1280 1245 0.000 0.100 +0.100 24 1175 1150 0.145 0.180 +0.035 26 1037 965 0.250 0.340 +0.090 T1 = calculated gap temperature ( F) before redistributic:: T2 = calculated gap temperature ( F) af ter redistribution g1 = initial gap size (in.) R2 = final gap size (in.) a 4-19
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4.2.3, Core Resistance Chances Caltalacions have been perforned to estinate the gap redistribution necessary to cause the observed 5% reduction in resistance at the conditions of Novenbcr . . , 1980 and Decenber 13, 1980. Assuming a uniforn initial gap distribution, the change in hydraulic dinneter which is necessary to cause a 5% resistance drop was calculated. Constant gap area is assuned, i.e., gaps are opened and closed. Assuming that half of the gaps close and the i other half open, it has been shown that the resistance reduction can be 1 accomplished by displacenents of 0.060 in. The calculated displacenent necessary to cause the resistance drop is consistent with the displacenent necessary to explain the gap tenperatures changes. 4.2.4 Region 35 Jaws Calculation The neasured outlet tenperature of region 35 decreases at the tine the tenperature redistribution occurs. This can be attributed to several pos-sible 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). A change in the crossflow (jaw) gas entering (or leaving) the control chan-nel within the region is also an explanation for the abrupt tenperature response of the niddle and botton ICRD therr.ocouples. Calculations indicate that a jau size of 0.1 in. existing halfuay around a region (about 6 ft in length) can cause the observed change in region 35 exit tenperature. 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 botton of the block. This displacenent is again consistent with all other indicators of displacecents being on the order of about 0.060 in. to 0.10 in. 4-22
The ICRD thernoccuples at the middle and botton of the control rod hole exhibit a nearly stepwise increase in tenperature at the tine of the t_nper-ature redistribution. This sharp increase can be explained by an abrupt change in either hotter or colder crossflow into or out of the control rod channel. Figure 4-9 shows the observed behavior during the November 14, 1980 event. The cost probable explanation is that a colder crossflow (jaw) before the tenperature redistribution event is suddenly shut off (e.g., jaw closed). This conclusion is based upon calculations perforned with a con-trol rod channel thernal/ flow nodel of estimated and actual tenperature rises to the niddle thercocouple before and after the outlet tenverature redistribution. 4.2.5. Type II Flow Segnents Previous analyses and data (Ref. 4) have consistently enphasized the , probable existence of a cool transverse flow along the thernocouple sleeve. l This flow can cause the neasured outlet temperature to be soneuhat different l fron the actual region outlet temperature for the boundary regions on a thernocouple string. The analysis to date indicates some differences of noninally 50*F to 150*F betwcen expected and ceasured region exit tenpera-ture. Analyses have shown that pressure gradients of 0.010 psid across the sleeves extending through individual core support blocks can cause trans-verse flow of about 10 lbn/hr and a tenperature error of about 50 F. Chane-ing the pressure gradient to 0.03 psid increases the Type II flou to ~30 lbn/hr and the temperature error to ~150*F. Therefore , changes in the gap flow through the support block gaps can cause changes in pressure differ-ences capable of causing changes in neasured outlet tenperature of 150 F even though the actual temperature of the outlet gas recains constant. Inter-region gap size changes on the order of 0.10 in, are capable of pro-ducing changes of this magnitude in the lateral pressure drop across the thernocouple sleeves, as well as changing the tenperature of the gas traversing the sleeve. 4-23
The Type II flow effects are sunnarized in Fig. 4-17 where the effect of inter-region gap changes are shoun. 4.2.6. Nuclear Channel Deviations A nuclear cbunnel " deviation" is defined as the individual channel's response ninus the average response of all channels. Nuclear channel devia-tiens have proven to be highly reliable and sensitive indicators of fluctuations (Refs. 3 and 4). The nuclear channel response for three typical channels during the load increase which initiated the region exit temperature redistribution is shown in Fig. 4-18. The corresponding nuclear channel deviations for all six channels are shown in Figs. 4-19 and 4-20. The deviations during the exit tenperature redistribution are characterized by snall initial offsets fol-lowed by a gradual (10 to 13 nin) approach to a new stable value. These abrupt offset deviation responses are significantly snaller than those typi-cally observed during fluctuations (prior to installation of RCDs), and they do not exhibit any cyclic behavior. For conparison, the deviation response during a cycle 2 fluctuation is shown in Fig. 4-21. The initial offset behavior of the deviations is caused by snall changes (<0.3%) in the neutron streaming through gaps in che side reflector. The larger (~0.7%) gradual changes are responses to thernal ef fects , e.g. , changes in core and reflector tenperatures and/or configuration resulting from the redistribution of gaps. While the deviation responses during the region exit tenperature redis-tribution are clearly not fluctuations, a conparison of the data in Figs. 4-19, 4-20, and 4-21 does indicate sone sinilarities. Note for exanple channel 5 in Fig. 4-19, which shows an initial snall offset followed by a slou response. This is quite sinilar to an extension of the circled areas in Fig. 4-21. This seens to indicate that the two responses are the result of sinilar phenonena. During fluctuations, gap changes cause the initial 4-24
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offset responses, and thermal effects are then evident for a few minutes until a second gap change occurs and causes a second offset response. In contrast, during the exit tenperature redistribution a very 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). 4.2.7. Reactivity Perturbations i Further evidence of core displacement is the small positive reactivity change (~ld) which occurred at the time of the exit region temperature redistribution. Perturbations to core reactivity of similar magnitude were
, observed during fluctuations prior to installation of RCDs (Ref. 3). How- l ever, during fluctuations these perturbations 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 dispiecement of core components so as to reduce the effective diameter of the core, i.e., a compression or tight-ening of the core so as to close the gaps between regions. The reactivity I l
perturbations occurring during the initiation of the region exit temperature redistributions correlate with the onset of the changes in the region exit and gap temperatures. 4.3. SCENARIO OF EVENTS The data and analysis clearly support a temperature redistribution phenonenon which is caused by a small physical displacement of the fuel elements. This displacement results in a " tightening" of the core in an "hourglassing" manner wherein the gaps between outer regions are increased and gaps between inner regions are decreased. The decrease in core flow resistance, the gap temperature data, and calculations are consistent with the opening and closing of gaps between regions by amounts of 0.060 to 0.100 in. 4-30
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 notion displacenents themselves, i.e., ~0.10 in. The nuclear channel deviation behavior and the snall (~16) reactivity changes during the initiation of the temperature redistribution are responses to the redistribution of gaps (i.e., core geometry) and the corresponding redistribution of temperatures. The redistribution of the inter-region gaps can cause changes in Type Il flow sufficient to cause changes in measured region exit tenperatures sufficient to explain the differences between the expected and neasured results for boundary regions. Regions 1 through 19 are not likely to be significantly affected by changes in jaws type crossflow. The tenperature redistribution in these interior regions can be explained by decreased gap cooling and a slight redistribution of flow due to increased bypass flow. Regions 20 through 37 are generally cooler than expected after a temperature redistribution. These effects can be explained by a combination of changes in crossflow (jaw flow), increased bypass cooling, and Type II flow effects. These effects are at present not separable but certain teasured data characteristics are suggestive of all three depending upon the region being investigated. The "hourglassing" or core " tightening" scenario is sunnarized in Fig. 4-22 which shows the correlation between various predictions and obs e rva tions . 4-31 u__________ _. _ _ _ _ _ . _ _ _ . _ _ _ _ _ _ _ _ _ _ _ . .__ - _ . . _ _ . _ _ _ ___ _ _]
3 E 2 e ! E 5 CORE i CORE k
$ c f ! / T E k f
BEFOPI AFTER REDISTRIBUTION PIDISTRIBUTION PPIDICTION OBSERVATION REACTIVITY INSERTION OF ~1d REACTIVITY INSERTION OF ~16 DUE TO CORE GEOMETRY CHM!GE DECREASE CORE- R BY ~5% FOR 4% to 3% DECREASE IN CORE R CORE REDISTRIBUTION OF 0.06 TO 0.10 IN. INNER REGION T-EXITS INNER PICION T-EXITS INCREASE DUE TO DECPIASED FLOU INCREASED AND DECREASED GAP COOLING EOUNDARY REGION T-EXITS BOUNDARY REGION T-EXITS DECPIASE DUE TO JAWS FLOW, DECREASED INCREASED GAP COOLING AND TYPE.II FLOW "JAUS" FLOW PATHS OPENED "JAUS" FLOW EVIDENT IN IN BOUNDARY REGIONS DUE PIGION 35 (ICRD) TO FUEL BLOCK DISPLACEMENT INTERIOR REGION GAPS GAP CHANGES DEDUCED FROM CLOSE, BOUNDARY PIGION TEMP. CHANGES CONSISTENT CAPS OPEN (HOURGLASS) WITH PREDICTION CHM GE IN TRANSVERSE FLOW DEDUCED CAP CHANGES ARE F1.TE 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 core region outlet thermocouples. Fig. 4-22. Temperature redistribution scenario 4-32
- 5. SAFETY CONSIDERATIONS 5.1.
SUMMARY
The core temperature fluctuations that were observed during cycle 1 and during cycle 2 prior to the installation of the region constraint devices (RCDs) were caused by a cyclic core component socion. The cycle is charac-terized by a slight movement of several core regions and side reflectors followed by a period of 5 to 10 minutes, and then another restoring movenent and a period of 5 to 10 ninutes (resulting in a 10 to 20 minute period). Testing to date with the RCDs in place has shown no evidence of flue-tuations; however during a rise to 70% power and with a high core pressure drop, a region exit temperature redistribution has been observed and repro-duced. Many of the characteristics . observed at the onset of fluctuations are apparent during this temperature redistribution, and it is concluded that the outlet temperature redistribution is the result of a single core displacecent event similar to that which was previously observed at the initiation of a fluctuation. The significant difference is that fluctua-tions were not induced with RCDs in place, due to the stabilizing influence provided by the RCDs. In fact, after each of the outlet tenperature redis-tribution events, the reactor was increased up to 70% power with a core pressure drop of ~4.2 psi. Core temperatures renained generally stable. Because the outlet temperature redistribution is caused by a cechanism similar to that which previously produced fluctuations, the safety evalua-tions for a fluctuation remain valid for the outlet tenperature redistribu-tion event. These have been presented at length in an earlier submittal to the Nuclear E+.gulatory Comnission (Ref. 1). 5-1
An evaluation of the consequences of the outlet temperature redistribu-tion to core fuel temperatures has been nade and is discussed in this sec-tion. Based on the results of this evaluation it is concluded that the reactor may be operated after the temperature redistribution, up to 100% power, without increasing the risk to public health and safety. 5.2. COMPARISON OF FLUCTUATION AND OUTLET TEMPERATURE REDISTRIBUTION CHARACTERISTICS Several similarities between the various tenperature and nuclear channel ceasurements at the initiation of a fluctuation and those for the temperature redistribution have led to the conclusion that the redistribu-tion is a single notion event similar to that which was previously observed at the initiation of a fluctuation. These data have been presented in detail in the previous section; the sinilarities are listed below:
- 1. Sudden drop in core resistance, attributed to the conbination of the gaps between regions into a few larger gaps.
1
- 2. The tenperature signatures of the thernocouples located in the gaps between core support blocks. The initial shape and nagnitude of the tenperature changes are nearly identical for several of these thernocouples.
- 3. Slight abrupt changes in the wide range nuclear channel signals caused by novenent of the permanent side reflector colunns opening neutron streaming paths.
- 4. Slight reactivity changes which can be explcined by slight changes in the effective core diameter.
- 5. The largest region tenperature changes occurring generally in the same regions, that is the regions in the NW sector of the core.
5-2
- 6. Tenperatures ceasured by the region 35 instrunented control rod drive thernocouples located in the niddle and botton of the region, which indicate the presence of jaws type crossflows.
5.3. SAFETY EVALUATION OF THE OUTLET TEMPERATURE REDISTRIBUTION 5.3.1. Wide Range Linear Channel Flux Signals The linear nuclear channel signals displayed both a small rapid change on some of the detectors and a very small (~16) reactivity change. Sinilar behavior has also been observed during fluctuations, and changes that were observed were about the same magnitude or larger than those observed during the outlet temperature redistribution. The abrupt changes are explained by small displacenents in the pernanent side reflector resulting in changes in snall neutron streaning paths. Not all detectors displayed this behavior, and tenperature feedback normally following real reactivity changes was not observed. Small reactivity changes are observed and are explained by core displacement causing an effectively snaller core diaceter. As discussed in the fluctuation safety analysis report (Ref. 1), the nost extrene change in reactivity which could result from the core notion is to compress the core so as to close all available gaps. The resultant reduction in neutron leak-age amounts to only 0.00015 ak. In FSAR Section 14.2.1.3, the effects of a reactivity change of 0.006 6k were evaluated and no danaging effects were found. Therefore, no significant effects from these snall reactivity changes due to core displacenent are possible. 5.3.2. Control Rod Insertability As discussed in the fluctuation safety analysis report, the naxinun nisalignnent of a control rod channel available if all gaps across the core are conbined is 1.5 in. Rod insertion tests were conducted using 1.6 1 . nisalignnent at the insertion location and 2.5 in. nisaliganent in tue rod channel; no appreciable increase in scran times was noted when compared to sinilar tests with the core aligned. With the RCDs installed only 5-3
negligible displacenent can occur at the top of the core and the naxinun displacement of the middle of a region relative to its ends is less than 1.5 in. The conclusion is that no predictable nisalignment in the core will interfere with the ability of the control rods to be inserted or withdrawn. 5.3.3. Structural Considerations 5.3.3.1. Possible Inpact Velocities. In section 5.3.1.5.2 of Ref. 1, a maximun fuel element impact velocity of 3 in./see was calculated. The ana-lytical nodel was one in which whola regions noved horizontally, driven by transverse pressure forces, until they impacted with neighboring regions. After the installation of the RCDs, this type of notion is no longer pos-sible, because the regions are restrained at the top. Another node is possible, however, within the geonetric constraints. In this mode a fuel colunn opens (jaws) at the interface between blocks to form a.two-link nechanism. This nechanism coves in a mode where the top and botton are stationary, while the middle hinge leads the notion. The transverse pressure differential which could form such a nechanisn was first calculated. The corresponding location of the niddle hinge was found to be about 104 inches (or less) from the top. It was then assumed that the nid-die hinge'noves through the estinated largest gap to contact an adjacent column. Using a conservative nethod and assening 100% core power condi-tions, the maxinun impact velocity was calculated to be 2.3 in./sec, which is less than the 3 in./sec found in Ref. 1. 5.3.3.2. Core Structural Loads. In the section above, it was shown that the naximun possible inp1ct velocity, if core notion occurs with the top restrained by the RCDs, is less than assumed in Ref. 1. Fbreover, the node of inpact is also sinilar from the standpoint of causing inpact loads. It can, therefore, be concluded that the structural loads are bounded by the results in Ref. 1, where they were shown to be sna11 compared to the load capacity of the fuel elenents. 5-4 4
5.3.3.3. RCD Structural Loads. In section 5.3 of Ref. 5, the cost highly stressed part of the RCD was found to be the Inconel pin, which is subjected to a maxinun load of 1,167 lb during normal operation. If core notion occurs as discussed above, the inpact load in the pins would be negligible because of their renoteness from the inpact zone. Subsequent to the impact, however, about 50% of the pressure force on the displaced colunn would be transferred to the column it leans on, and its associated RCD pin would experience a higher shear force. If it is assuned that the maxinun pin force of 1,167 lb increases by 50%, which is very conservative (since nost of the original force cones from a postulated leaning of seven colunns in the sane direction due to uneven irradiation shrinkage), the naximun stress in the pin would increase fron 36,000 psi to 54,000 psi, still below the
; yield limit of 134,000 psi by a large nargin.
5.3.4. Secondary Systen The secondary systens have been eliminated as the cause of fluctuations or the outlet temperature redistribution because steam temperature perturba-tions lag the prinary side (helium) temperature perturbations. Further, the helium tenperature perturbations are danped in the stean generator, i.e., the steam temperature perturbations are smaller. The principal concern in the secondary systen during fluctuations was the effect of varying stean temperatures on the fatigue stress limits of the stean generator modules, and for this reason the duration for fluctuations during testing uas strictly linited. During the tenperature redistribution event the stean 4 generator nodules experience a single tenperature decrease of less than 10*F, and thereafter respond normally to the power rise. The-single tenper-ature decrease, which is believed due to increasing cold bypass flow, is of no consequence to the fatigue stress limits in the steau generator nodules. 5-5
5.3.5. Bypass Flow Iacrease After the Outlet Tencerature Redistribution The bypass flow fractioa* was calculated to be ~12% before the outlet tenperature redistribution events, increasing to ~14% during the tenpera-ture redistribution. The effect of this 2% increase is to increase naxi-nun core fuel temperatures by about 15*F and average core fuel tenperatures by about 5"F. It should be noted that the bypass flow fraction does not increase as power is increased fron 70% to 100%. 5.3.6. Accident Analyses FSV accident analyses (Chapter 14) were reviewed to deternine if any reevaluation was required for plant operation after the tenperature redis-tribution. It was determined that the localized initial conditions created by the outlet temperature redistribution would not affect the accident con-sequences since the accidents are initiated at Technical Specification lin-its and operation following the temperature redistribution is within those linits. Safety systens such as the plant protective systens, the reserve shutdown systen, and the liner cooling systen which nay be required to effect a safe shutdown are neither associated with nor inpaired by the tenperature redistribution. It was concluded that no reevaluation of the FSAR accident analyses is required as a result of the observed outlet tenperature redistribution. 5.4. LONG-TERM OPERATION WITH Tile OUTLET TEMPERATURE REDISTRILUTION The two outlet temperature redistribution events which occurred during fluctuation testing below 70% were nearly identical: the sane regions
- Core bypass flow fraction is the portion of core cavity flow which does not pass over the region exit thernocouples. It includes the flow in the vertical gaps between fuel elenent colunns, between side reflector colunns and between the side reflector and the core barrel.
5-6
exhibited temperature decreases during the power rise and the nagnitude of the tenperature decreases was conparable. Also the core operating condi-tions of power, flow, and core pressure drop were very nearly the same. In addition to the cooling effect of increased gap flow, there are two probable causes of the tenperature decreases experienced by sone core bound-ary regions. These are cool bypass flows which enter the thernocouple probes at the side of the core support floor (Type II flow) and depress the thernocouple reading, and the opening of small horizontal gaps between the stacked columns in a region pernitting cool bypass flow (crossflow) to enter the regions. Analysis and evaluation of the data indicate both of these phenonena are occurring, as explained in Section 4. Crossflows during the temperature redistribution are evidenced by the rapid temperature changes neasured by the middle and botton thernocouples on the ICRD in region 35. However, Type II flow correlates with more of the observations. The observations indicating Type II flow are discussed below. The changes in stean generator nodule hellun inlet temperctures calculated assuming the Type 11 flow explanation generally compare core favorably with the neasured temperature changes chan if crossflow into these regions is assumed. The profiles of the tenperature traverses across these regions at core exit indicate the presence of Type II flow. Also, at the time of the region exit temperature redistribution, a reduction in the core flow resistance was observed. This is explained by a 2% increase in the core bypass flow due to the combination of the core bypass gaps into fewer, larger gaps. An alternative, independent method of calculating the core bypass flow based upon neasured region outlet and stean generator nodule inlet heliun temperatures yields a ~2% increase only if Type II flow is assumed to be affecting the region outlet temperature neasurenent. Further evidence of Type 11 flow is available from cycle 1 and cycle 2 operation. Type II flow best explains the differences between the ceasured 5-7
and calculated steam generator nodule temperatures. For steam generator nodules B l-5 and B 2-6 the calculated helium inlet temperature, based on the ceasured region outlet tenperatures, is lower than neasured. These nod-ules are adjacent to core regions 34, 37 and 20, which have measured region peaking factors (i.e., exit gas tenperatures) lower than predicted from core physics calculations. Type II flou provides a single explanation to resolve both of these differences. The regions which are nost affected are those on the core boundary and at the end of thernocouple strings. These regions have the greatest potential for a significant Type II flow effect since the sleeves containing their region exit gas thernocouples are open to the core boundary where cool Type II flow is available. Additionally, during a loop isolation event some of these regions have shown a sharp tenperature increase while the rest of the regions show the expected decrease. This " retrograde' behavior can.be explained by a sudden reduction in the Type II flow caused by a sudden loss of driving potential for the Type II flow. Finally, during fluctuations, the thermocouples in sone of these regions as well as sone nearby gap thernocouples have dis-played tenperature changes which do not correlate with the stean generator nodule heliun inlet temperatures and which can be explained by changes in the magnitudes and/or direction of the Type II flows. The majority of the observations indicate Type II flow as the primary mechaniso for the observed exit tenperature reductions in the boundary regions. Nevertheless, a temperaturc analysis was perforned for the region exit tenperature redistribution event to deternine the fuel tenperature . changes which could be caused by either of the two postulated mechanisas, i.e., Type II flow or jaws crossflow. Region 36 was selected because in both events it experienced the largest outlet temperature decreases. In both cases core physics calculations were used to determine the region power after the region exit tenperature redistribution. Then, for the case where Type II flow was assuned to enter the thernocouple probe, the fuel tenpera-tures vere calculated using this region power and the region flow rate inferred from the region flow control valve setting. 5-8
For the case where cool crossflow is assuaed, the effect of the cross-flow depends on its location and tenperature. A larger fuel tenperature increase occurs when the crossflow location is lower in the core and when the temperature of the crossflow is high. Crossflow calculations were per-forned for crossflow locations 1/3,1/2 and 2/3 of the way down the active - core. As discussed in Section 5.3.4.1, the probable location for a cross-flow gap to develop is about 104 in. or less from the top of the core, or about 1/3 of the way down the active core. In fact, it is nost probable the first crossflow gap developed near the top of the region. The additional calculations, at 1/2 and 2/3, were perforned as conservative cases. For all the cases, the temperature of the crossflow was taken to be 40% of the region temperature rise at the location of the cross-flow gap. This is also a conservative assunption, since the regions exhib-iting the largest outlet temperature redistribution are located on the core boundary where the bypass flow is only partially heated. Also , the tenpera-ture redistribution event is believed to create a conbination of the core bypass gaps into a few larger gaps, and flow in these gaps would not be efficiently heated. With these conservative assunptions the crossflow and region flow rates were then deternined from the core pressure drop and heat balance equations, using the region power fron the core physics calculations. For the case where the region 36 outlet temperature redistribution is assuned to be caused by a cool bypass flow entering the thernocouple probe (Type II flow), the redistribution event does not change the region fuel temperatures. For the case where cool crcssflow into the region is assumed, the effect of the crossflow is to decrease the peak fuel tenperature fron the value with no temperature redistribution, as is shown in Fig. 5-1. The temperature decreases are 135*F, 65*F, and 25*F for crossflow locations of 1/3,1/2 and 2/3 of the way into the active core, respectively. The tenper-atures are decreased because the crossflow into the region reduces coolant and fuel tenperatures in the botton of the region, where peak fuel tenpera-tures occur. The result is the peak fuel tenperature location shif ts to 5-9
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a point just above the crossflow location (for the 1/2 and 2/3 location cases), and even though the tenperature at this location is increased, it is lower than the fuel tenperature at the botton of the region without cross-flow. For the 1/3 location case the peak fuel temperature remains at the botton of the region since the fuel tenperatures 1/3 of the way into the core are noninally low. The region exit tenperature redistribution has occurred twice during testing wherein attempts (unsuccessful) were made to induce fluctuations. Each time the redistribution occurred at about 60% power, and after the redistribution the core power was increased to 70%. During and after these test periods, including up to 32 hours of stable reactor operation with the outlet temperatures redistributed, no increase in prinary coolant activity was observed. Since significant fuel danage would result in an increase in the primary coolant systen activity inventory, it is concluded that no sig-nificant fuel danage resulted fron reactor operation during or after the outlet tenperature redistribution events. After both tenperature redistribution events the core continued to be operated using the procedure that has been inplenented for both cycle 1 and cycle 2 operation, i.e., the flow control valves on the interior regions are adjusted such that their region exit temperatures are generally slightly greater than core average. The flow control valves on the boundary regions are adjusted to balance the hellun inlet and reheat stean tenperatures of the twelve stean generator nodules. This procedure ensures that those regions on the core boundary which are experiencing a Type Il flow into their thernocouple probes continue to be operated within the technical spec-ification limits. This procedure also ensures that the interior regions are operated within the linits of the Technical Specifications both before and after a region exit tenperature redistribution. This procedure will con-tinue to be used after a temperature redistribution and for operation up to 100% power. 5-11
6 REFERENCES
- 1. " Safety Evaluation - Reactor Outlet Temperature Fluctuations,' P-79,137, August 11, 1978.
- 2. GA RT-500, Revision H, December 1979 - PSC Subnittal to NRC P-80021, February 7,1980.
- 3. R. Hackney and J. Saeger, " Investigations'of the Fort St. Vrain Cycle 2 Reactor 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 Subnittal to NRC P-79287, November 28, 1979.
- 5. "SAR for Core Region Constraint Devices," P-79068, March 23, 1979.
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