ML20010B561

Draft Testing at Fort St Vrain After Installation of Region Constraint Devices

ML20010B561
Person / Time
Site:	Fort Saint Vrain
Issue date:	02/05/1981
From:	GENERAL ATOMICS (FORMERLY GA TECHNOLOGIES, INC./GENER
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ML19263D739	List: ML19263D738 ML20010B552 ML20010B554 ML20010B556 ML20010B559 ML20010B561 ML20010B562
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FOIA-81-127 NUDOCS 8108170237
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{{#Wiki_filter:_ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ - - _ _ - - _ _ TESTING AT FORT ST. VRAIN AFTER INSTALLATION OF REGION CONSTRAINT DEVICES DRAFT I' GENERAL ATOMIC COMPANY FEBRUARY 5, 1981 8108170237'810529 - PDR FOIA MULLENB1-127 PDR

-o CONTENTS 1. INTRODUCTION. 1-1 2.

SUMMARY

2-1 3. TESTING 3-1 3.1. Steady-State Testing. 3-1 3.2. Fluctuation Testing 3-3 3.2.1. Objectives. 3................... 3.2.2. Test Strategy. .3-3 3.2.3. Test Procedure 3-3 4. DATA AND INTERPRETATION 4-1 4.1. Observations on Date. 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. Summary 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' 5. SAFETY CONSIDERATIONS 5-1 5.1. Summary 5-1 5.2. Comparison ot Fluctuation and Outlet Temperature Redistributica Characteristics. 5-2 f 111 p

^ l ,o 1 l 5.3. Safety Evaluation of the Temperature Shif t Event 5-3 5.3.1. Wide Range Linear Channel Flux Signals. 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. Byoass 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. TESTING ABOVE-70% POWER WITH REGION CONSTANT DEVICES. 6-1 6.1. Description of Test (RT-500) 6-1 6.2. Test Prccedure. 6-2 7. R;FERENCES. 7-1 APPENDIX A. A-1 FIGURES 3-1. Fluctuation test strategy.. 3-4 3-2. Core pressure drop as a function of power (November 1980 tusting sequence). 3-7 3-3. Core pressure drop as a function of power (December 1980 testing sequence).. 3-10 4-1. Region outlet *emperatures (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. Ndelear channel durations: channele 3, 4 and.5. 4-4' 4-4. Calibration tube thernosouple locations 4-6 4-5. Gap T/Cs 3, 4, and 5 for November 14, 1980 4-7 4-6. Jap T/Cs 7, 8, 9, and 10 for November 14, 1989 4-8 ~4-7. Core coolant flow resistance: Ncvember 1<4, 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 durations during fluctustions 4-12 iv - r .. - ~

FIGURES (Continued) 4-11. Temperature redistribution, T/Cs 11 and 13 4-13 4-12. Temperature fluctuation, T/Cs 11 and 13. 4-14 4-13. Expected versus measured region exit temperatures (interior regions) 4-16 4-14. Expected versus measured region exit temperatures (boundary regions) 4-17 4-15. Calculated gap redistributions for November 14, 1980 4-20 4-16. Gap outlet temperature calculated versus measured during t,emperature redistribution 4-21 4-17. Type II flow calculations. 4-25 4-18. Nuclear channel response during temperature redistribution 4-26 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 5-1. Region 36 peak fuel temperature af ter the region exit temperature redistribution assuming Type II flow or i crossflow. 5-10 TABLES 3-1. Sequence of events (November 1980) 3-5 3-2. Sequence of events (December 1980) 3-3 3-3. Temperature redistribution initiating conditions 3-11 4-1. Calculated gap changes. 4-19 V

.{ 1. INTRODUCTION During the initial rise-to power program in October 1977, while approaching 60%- power, temperature fluctuations were. observed in the primary coolant circuit at individual core region outlets and at the steam generator -module inlets of the Fort St. Vrain reactor. A comprehensive program of investigation into the nature and cause of the-fluctuations was initiated immediately. The fluctuation investigations lead to the design and fabri-cation of region constraint devices (RCDs) as a permanent solution to the fluctuations. These mechanical 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 at the top of the core to near nominal values. Steady-state testing was performed during initial operation following installation of the RCDs to verify 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 fluctuations was performed in November and Decem- -ber of 1980. The results of these tests are summarized in this report. e 6 1-l'

e =. y [.} - V 2.

SUMMARY

4 The steady-state testing conducted after the installation of RCDs confirmed that the installation of RCDs had a minimal effect on the overall ~ core performance. Testing wherein attempts were made to induce fluctuations, after installation of region constraint devices, was first conducted on November-12 through November 15, 1980. Power levels from 40% to 70% were sur;2yed at two sets of core orifice' positions (core flow resistances) with a maximum 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 from 55% to 59% l power at the higher core resistance, a region exit temperature redistribu-tion was observed. This redistribution of region outlet temperatures resulted in several boundary region outlet temperatures, particularly 1n the NW corner of the core, decreasing while inner core region outlet tem-e peratures generally increased somewhat more than expected from the power change. This redistribution of temperatures generally persisted throughout-the remainder of the test. Steam generator helium inlet temperatures also recorded a change in distribution. Due to the loss of some of the steam generator data as well as'to a desire to confirm repeatability of the nhenomenon, the test was rerun along the higher core resistance line on December 12 through 14, 1980. Power I levels from 40% to.70% were surveyed up to a maximum core pressure drop of 4.2 psid. A region outlet temperature redistribution almost identical to that observed in November was encountered under essentially the same conditions. Again no fluctuations were observed. 2-1 m ~ l

The retion exit-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 displace-ments are not cyclic. These small displacements cause changes in gap dis-tribution (between regions), crossflow, and in the amount of transverse helium flow along the sleeve (s) surrounding the region exit temperature thermocouples. These observations are consistent with a " tightening" of the core, wherein the outer region gaps generally are increased and inner region gaps are generally decreased. Testing below 70% power has been completed. The testing has demonstrated 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 manner above 70% power without increased risk to the health and safety of the public. 2-2 ~

2 3. TESTING After installation of the region constraint devices (RCDs), both steady-state and fluctuation tests were performed. The objectives of these tests were to eve.luate the effect of the RCDs on steady-state core performance and on the fluctuation threshold. 3.1. STEADY-ST/TE TESTING Three steady-state t2?ts were performed during initial operation with RCDs installed to verify that the overall steady-state core performance was unaffected by the presence of the RCDs. These tasts provided data on the region peaking f actor (RPF) distribution, temperature profiles along the region exit thermocouple (T/C) calibration tubes and selected orifice calibraflons. ~~ RPF distributions were measured for power levels from 5% to 69% power. These measured data were compared to the computed RPF distribution. Simi-larly the measured and computed RPF distributions ebtained prior to instal-lation of the RCDs were compared. (The computed and measured RPFs were com-pared ra :er than the RPFs themselves since all measured data did not have the same orifice configuration and/or control rod positions before and after installation of the RCDs.) These data were used to evaluate the effect cf RCDs on the RPF distribution. Temperature profiles were obtained by moving a T/C inside the cali-bration tubes of the core outlet thermocouple penetrationt in approximately one-inch increments 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 inlet and outlet temperature were very nearly the same for meaeurements made both before and after installation of the RCDs. Temperature traverses were neasured in each of the sesen (7) penetration 4 3-1

d ? tubes at ~70%: power. These. traverses were compared to those measured before l-installation of the RCDs. ' -Orifice calibration data were measured for the orifice valves in

y regions 10, 28 and 34.

These regions represent a refueled interior region (10), a refueled boundary region (28) and a boundary region (31) which had not been refueled. \\ Results of the analyses of these test steady state data showed no significant measurable effects on core performance. Comparisons of these data with data obtained prior to installation of the FCDs are summarized below. 1. Core reactivity is unaffected. 2. The temperature profiles along the calibration tube penetrations are basically the same' with some 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 J changed. From the results of these studies it was concluded that the core is perforuing as expected with no significant measurable changes as a result of installation of the LCDs. 3.2. FLUCTUATION TESTING 3.2.1. Objectives ET-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 k[

originally performed in November 1978 during cycle 1 operation as a part of the fluctuation test program. While this test has undergone numerous 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 performed in November and December 1980. The purpose of this test was to demonstrate that the installation of region constraint devices solved the fluctuation problem or to determine what impact their installation had on the fluctuation threshold. 3.2.2. Test Strategy RT-500H was a test wherein attempts were made to initiate fluctuations, thereby determining 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 psi at 40% power with a core flow resistance

• of ~45.

The power would then be increased in steps of ~3%/ min (waiting at least 2 hours between steps) to 70% power. This procedure was planned for three values of resis-tance. A schematie-of-this test strategy-is-shown -in-Fig 3-1. ihe details of the test (RT-500H) are given in Ref. 2. 3.2.3. Test Procedure RT-500H was initially performed on November 12 to 15,1980. The sequence of events for that test are given in Table 3-1. A su= mary of the test procedure is given in the following paragraph. The core flow resistance parameter is defined as: 13 AP P R = 2 x 10 T. ( )1 85 where AP = measured core pressure drop (psid), P = PCRV precoure (psia), T = average circulator inlet temperature (*R), and 6 = total circulator flow rate (lbm/hr). 3-3

c. 1 R = cogs ptow ass. stance R3>R z>R g R R ~ 4. 3 3 z l E,t

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l (ORE LP l 77tes:Ao'b / ~%O % ~ 70 % CoR= 20WEPs Fig. 3-1. Fluctuation test strategy 3-4

TABLE 3-1 SEQUENCE OF EVENTS (November 1980) Power Flow Core AP Date (Time) (psi) Resistance Comments STEADY-STATE CONDITIONS PRIOR TO PORER INCREASE 11/12/80 (1043) 40 54 1.6 45 Starting conditions 11/12/89 (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 2.5 43 11/13/80 (0942) 62 79 3.1 44 11/13/80 (1224) 65 82 3.3 43 11/i3/80 (2130) 70 86 3.6 43 After final power rise at R = 43 l 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 temper 4ture redistribution l REDUCE POWER TO REPEAT PREVIOUS POWER RISE 11/14/80 (1805) 56 72 3.2 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-500H REDUCE P0kT.R TO 40% AND REORIFICE FOR NORMAL OPERATION 3-5

e f Initial conditions of 40% power, a core pressure drop of.1.6 psid and a resistance of ~45 were established. The power was increased in steps of - ~3% at ~3%/ min, waiting at least 2 hours between steps, to 70% at a core pres- ~ sure drop of 3.c psid, with no unusual behavior. The power was then reduced to 40% and the core was uniformly reocificed for a core pressure drop of 2.1 psid with a resistance of_~57, maintaining the same core flow distribution. The power was again increased, in steps, toward 70%. After the power increase to 59%, a region exit temperature redistribution was noted. After monitoring core behavior at 59% power for ~3 hours the power was reduced to ~55%. Again af ter monitoring core performance for ~3 hours and observing no unusual behavior, the pcser 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 temperature distribution remained reable with the. exception of a partial restoration of the pre-event temperatures in two regions. At this point in the testing it was decided to terminste testing pending detailed evaluation of t'ne data obtained during the temperature 4 redistribution. Power was then reduced to ~60% and the core was reorificed for normal o, ration. During this reorificing, beginning at a core AP of r ~2.8 psid, the region exit temperatures returned to its pre-erent distribu-tion. 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 readable. It was subsequently decided to repeat tn. second resistance line of the above test in order to obtain SG data and to determine whether the observed temperature redistribu-tion was repeatable. In addition, the power was to be reduced from 70% to 40% in steps ta obtain more data on the return of the temperatures to their pre-event distribution. f-The sequence of events for this secon'd test, performed on December 12 to 14, 1980 is given in Table 3-2. A summary of the test procedure is given in the following paragraph. I i 3-6

4 S.o I 4 0W 4.0 S ,a y y A p ,8 M 3# .AN M-q ,U ,D' Q h p", f 'Gf i f p' / M Rf3HOLb U f-Ling ,e ,-O c Gf / d /.o 'IG Co y POWGg (c,/Q C <C/26 Fig. 3-2. Core pressure drop as a function of power (November 1980 testing sequence) '3-7 ..a,.,,_....m. ,--,,s....

TABLE 3-2 SEQUENCE OF EVENTS (December 1980) Power Flow Core AP Date (Time) (psi) Resistance Comments STEADY-STATE COB *DITIONS 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 34 12/13/80 (1108) 58 73 3.3 54 After power rise that initiated temperature redistribution 12/13/80 (2240) 62 76 3.5 53 12/14/80 (0204) 62 75 3.5 53 j 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 i 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 2.4 54 12/14/80 (1704) 43 56 2.1 54 12/14/80 '(1742) 40 53 1.8 54 End of RT-500H i 3-8

a s 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% st 3%/ min. During the power rise to ~58% the regf on exit' temperature redistributicn again occurred and was remarkably simflar to that observed during the November 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-et ent distributions following the power reduction from ~52% (core AP ~2.8 psid) to 49% (core AP ~2.6 psid). The core was then reorificed at 40% power for normal operation. The core pressure drop as a function of power, obtained during this test, is shown in Fig. 3-3. To illustrate'the remarkable similarity of the November and December regicn exit temperature redistributions, 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 drsp following the initiating load increase is 3.4 psid. 3-9

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d i.- l TABLE 3-3 TEMPERATURE REDISTRIBUTION INITIATING CONDITIONS I Noveciber i December. i Power increase, % 55 + 59 58 + 61 Flow, % 70 + 75 75 + 76 Core AP, psid .Stea'dy state 3.2 + 3.4 3.3 + 3.4 Transient' peak 3.8-3.8 4 4 1 5 a 4 1 a r b 4 3-11 .-=.... - - ,. ~. _. -.,.....

r-_ i 4. DATA AND INTEPERETATION 4.1. OBSERVATIONS ON DATA 4.1.1. {pasuredRegionExitTemperatureRecistributions The region outlet temperatures were c:atinuously measured during both the November 14, 1980 and December 13, 1980 tempcrature redistributions. The temperatures before and after the November v.sdistribution 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) increased 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 signala during the initiation of the region exit temperature redis-tribution reveals the existence of a small reactivity insertion of about +14 (~0.00007 Sp) not due to rod control motion. Nuhlear channel deviations are defined as (X i - X) where Xi is the signal from nuclear channel i and X is the average of the six channels. During testing without RCDs installed, nucitar channel deviations proved to be highly reliable and sensitive indications of fluctuations. f 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 thermcouples indicate tem-perature changes during the temperature redistribution. The signals of rep-resentative gap thermocouples 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 the outer ring of regions generally decreased. 4.1.2.4. Core Resistance Change. The core flow resistance parameter (see Section 3.2.2) shows a sharp decrease at the occurre'nce 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 redistributiot. 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 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 the redistribution (Fig. 4-3) to that of fluctuations (Fig. 4-10). Also compare region outlet temperatures and gap temperatutes during and after tra redistribution (Fig. 4-11) to those during a typical fluctuation (Fig. 4-12). 4-5

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4.1.4. Summary of Data Observations i The data consistently indicate the region exit temperature redistribu-tion to be an event that simultaneously affects region outlet temperatures, gap temperatures, core resistance, and neutron flux to ex-core detectors. For fast responding instruments (e.g., nuclear channels and ICRD control channel thermocouples) there is an abrupt change in measured parameters gen-erally followed by a slower change, suggestive of thermal phenomena, until a new steady condition is reached. 4.2. ANALYSIS AND INTERPPITATIONS 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 measurements are explained and compared to calculations and mechanisms which tend to reproduce the observed data. 4.2.1. Fxpected Versus Measured Temperature Distribution In order to help understand the thermal behavior during the temperature redistributions, the expected temperature chenges were determined for each l region on the basis of both calculations and past history of performance during power rises. The expected values of the temperature changes are for a power increase taken by withdrawal of tl.e regulating control rod during nonfluctuating operation with no temperature redistributions. The expected behavior is compared with the measured behavior during the outlet temperature redistribution of December 13, 1980. Clearly the changes in temperature for the inner regions (1 through 19) are larger than the expected value (see Fig. 4-13). On the other hand, the measured 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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4.2.2. Cap Temperature Changes I The gap' temperatures are observed to change during the region outlet' - temperature redistribution. The temperature change can be explained (at least in part) by gap size changes which affect the heat transfer-and flow (Ref.'3). A thermal flow model of the gap between regions had been~previ-ously constructed (Ref. 3) and had been used to make calculations of gap i temperature and. flow responses. A series'of steady state analyses were performed for the reactor conditions before and after the temperature redistribution as a function of gap size'. The temperatures before and after the temperature redistribution were then. tabulated from the measured data. The temperature before the redistribution indicates the initial size of the gap, and from the change in temperature the final gap size may be determined as well. The results are i shown in Table 4-1 and again in Fig. 4-15 on the core map, t The gaps interior to the core are indicated to have 41osed by 0.060-to 0.090 in. The gaps adjacent to the outer rink regions have nominally opened by 0.035 to 0.100 in. This result indicates a general redistribution of the gaps in the core. 1 Additional transient calculations have been performed which further support the gap redistribution theory. With the assumption that an initial gap of 0.180 in. is step increased to 0.240 in, when the redistribution occurs, the gap temperatures through the power and flow transient of Novem-4 ber 14, 1980 were calculated. The results are shown in Fig. 4-16 (solid line) as compared to the actual measured data (broken line). Note the i excellent agreement. Even better agreement could be obtained with a more ' precise choice of initial and final gap size. 4 l i 4-18 i =...

TABLE 4-1 CALCULATED GAP CHANGES (November 14, 1980) T/C Tt T2 81 82 82 81 3 1310 1380 0.090 ~0.0 -0.090 4 1130 1079 0.180 0.235 +0.055 7 1330 1390 0.090 -0.0 -0.090 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 redistribution T2 = calculated gap temperature (*F) after redistribution gi initial gap size (in.) g2 = final gap size (in.)z 4 9 4-19

c E3 \\ CALCUIATED gg CHANGE IN 2 GAP (INCHES) \\2s / t' g [ s \\37 Q,, - 36 2e Q a - \\s - '848 Sss 7 ex' x x x x x J2 6 I 3 IC 23 26

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N"%%%9 h ~ Fig. 4-15. Calculated gap redistributions for November 14, 1980 4-20

GRP OUTLET TEMPERATURES A7 .015, 610 .320 .180" .240" 1t50 c > - - - - - - - - -- - - }- - - - - - - - - - - - ~ - - e-TC 4 I t4LASchltb :D/JA [t1 14

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It30 - -8 4 4 4 l P Alf_t{L A T 4 0:4 LtdAOL.J3. L. l l O e ta a .i i u n e u . ---.-,-----------l-------------l------------i------------ g it10-t-. ec 23 l l l l <t 2: w e-. 3090.--...--.....j....._ ;_e d,a.4 a @.f.... ....a -------------!......-...... % p! IFO i O 5 to 15 20 25 30 T[ME-MINUTES I il 61 3 3 GAP =.180* GAP =.240 # Fig. 4-16. Cap outlet temperature calculated versus measured during temperature redistribution

I 4.2.3. Core Resistance Changes Calculations have.been 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.. Assuming a uniform initial gap. distribution, the change in hydraulic diameter which is necessary to cause -a 5% resistance drop was calculated. Constant gap area 'is assumed, i.e., 6 gaps are opened and closed. Assuming that half of the gaps close and the other half open, it has been shown that the resistance reduction can be accomplished by displacements of 0.060 in. The calculated displacement necessary to cause the resistance drop is ~ consistent with the displacement necessary to explain the gap temperatures changes. i J 4.2.4. Region 35 Jaws Calculation The measured outlet temperature of region 35 decreases at the time the temperature redistribution occurs. This can be attributed to several pos-sible causes, one of which is a change in the influx of cold crossflow int,o 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 fcr the abrupt temperature response of the middle and bottom ICRD thermocouples. t Calculations indicate that a jaw size of 0.1 in. existing halfway around a region (about 6 ft in length) can cause the observed change in region 35 exit 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 of about 0.060 in. to 0.10 in. i I I t I 4-22

~ [ The ICRD thermocouples at the middle and. bottom of the control rod' hole exhibit aLnearly' stepwise increase in temperature at the tira of the temper-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 evast. The most probable explanation is that a colder crossflow (jaw) before the temperature' redistribution event is suddenly shut off (e.g., jaw closed). This conclusion is based upon calculations performed with a con-trol rod channel thermal / flow model of estimated and actual temperature rises to the middle thermocot ple before and af ter the outlet temperature redistribution. 4.2.5. Type II Flow Segments A -Previous analyses and data (Ref. 4) have consistently emphasized the probable existence of a cool transverse flow along the thermocouple sleeve. I This flow can cause the measured outlet temperature to be somewhat different than the actual region outlet temperature for the boundary regions on a j thermocouple string. The analysis to date indicates some differences of nominally 50*F to 150*F between expected and measured region exit tempera-f 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 lbm/hr and a temperature error of about 50*F. Chang-ing the pressure gradient to 0.03 psid increases the Type II flow to ~30 lbm/hr and the temperature error to ~150*F. Therefore, changes in the gap l flow through the support block gaps can cause changes in pressure differ-ences capable of causing changes in measured outlet temperature of 150*F even though the actual temperature of the outlet gas remains constant. Inter-region gap size changes on the order of 0.10 in. are capable of pro-I. ducing 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. i. .4-23

I The Type II flow effects are summarized in Fig. 4-17.where the effect 'of inter-region gap changes are shown. 4.2.6. Nuclear Channel' Deviations b i A nuclear channel " deviation" is defined as the individual channel's response minus the average ~ response of all channels. Nuclear channel . deviations 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 deviation for all six chan-nels are shown in Figs. 4-19 and 4-20. The deviations during the exit tem-perature 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 significantly smaller thar 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. 4-21. The initial offset behavior of the deviations is caused by small changes (<0.3%) in the neutron streaming through gaps in the side reflector. The larger (~0.7%) gradual changes are responses to thermal effects..e.g., changes in core and reflector temperatures and/or configuration resulting from the redistribution of gaps. j While the deviation responses during the region exit temperature redis-tribution are clearly not fluctuations, a comparison of the data in Figs. 4-19, 4-20, and 4-21 does ,'cate some similarities. Note for example channel 5 in Fig. 4-19, tk .a shows an initial small offset followed by a slow response. This is quite similar to an extension of the circled areas in Fig. 4-21. This seems to indicate that the two responses are the result of similar phenomena. 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 temperature 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 Further evidence of core displacement is the small positive reactivity change (~14) which occurred at the time of the exit region temperature redistribution. Perturbations to co-reactivity of simil2: magnitude were observed during fluctuations prior to installation of RCDs (Ref. 3). How-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 displacement of core componer:4 ao as to reduce the effective diameter of the core, i.a., s compression or tight-ening of the core so as to c' lose the gaps between regions. The reactivity perturbations occurring during the initiation of the region exit temperature redistributions correlate with the onset of the changes in the region 2xit and gap temperatures. 4.3. SCENARIO OF EVENTS The data and analysis clearly support a temperature redistribution phenomena which is caused by a small physical displacement of the fuel ele-ments. This displaces st results in a " tightening" of the core in an " hour-glassing" manner wherein the outer region gaps are increased and inner region gaps 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

4 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., ~0.10 in. The nuclear channel deviation behavior and the small (~1d) reactivity changes during the initiation of the temperature redistribution are responses to the redistribution of gaps (i.e., core geometry) and the corre;ponding redistribution of temperatures. The redistribution of the inter-region gaps can cause changes in Type II flow sufficient to cause changes in measured region exit temperatures sufficient to explain the differences betteen the expected and measured results. Regions 1 through 19 are not likely to be significantly affected by change's in jaws type crossflow. The temperature 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. 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 measured data characteristics are suggestive of all three depending upon the region being investigated. The "hourglassing" or core "ttghtening" scenario is summarized in Fig. 4-22 which shows the correlation between various predictions and observations. i 4-31

R R E E F p _E' CORE C oR E. E j h j C o T J BEFORE AFTER REDISTRIBUTION REDISTRIBUTION PREDICTION OBSERVATION REACTIVITY INSERTION OF ~1d REACTIVITY INSERTION OF ~14. DUE TO CORE GEOMETRY CHANGE DECREASE CORE R BY ~5% FOR 4% to 5% DECREASE IN CORE R CORE REDISTRIBUTIONF OF O.06 TO 0.01 IN. INNER REGION T-EXITS INNER REGION T-EXITS INCREASE DUE TO DECREASED F'0W INCREASED AND DECREASED GAP COOLING BOUNDARY REGION T-EXITS BOUNDARY REGION T-EXITSs DECREASE DUE TO JAWS FLOW, DECREASED INCREASED GAP COOLING AND TYPE II FLOW " JAWS" FLOW PATHS OPENED " JAWS" FLOW EVIDENT IN IN BOUNDARY REGIONS DUE REGION 35 (ICRD) TO FUEL BLOCK DISPLACEMENT INTERIOR REGION GAPS GAP CHANGES DEDUCED FROM CLOSE, BOUNDARY REGION TEMP. CHANGES CONSISTENT CAPS OPEN (HOURGLASS) WITH PREDICTION CHANGE IN TRANSVERSE FLOW DEDUCED CAP CHANGES ARE RATE ALONG T/C SLEEVE SUFFICIENT TO CAUSE CHANGE (i.e., TYPE II FLOW) IN TYPE II FLOW Fig. 4-22. Temperature redistribution scenario 4-32

, =-. r 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 motion. The cycle is charac-terized'by a slight movement of several core regions and side reflectors i followed by a period of 5 to 10 minutes, and then.another restoring movement and a period of 5 to 10 minutes (resulting in a 10 to 20 minute period). l i Testing to date with the RCDs in place has shown no evidence of fluc-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. Bhny of the characteristics observed of the onset of fluctuations are apparent during this temperature redistribution, and it is concluded that the outlet temperature redistribution is the result of a sing 16 core displacement event similar to that which was previously observed at the initiation of a fluctuation. The significanc 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 temperature redis-tribution events, the reactor was increased up to 70% power with a core pressurra drop of 4.1 psi. Core temperatures remained generally stable. Because the outlet temperature redistribution is caused by a mechanism-similar to that which previously produced flucteations, the safety evalua-- tions for a fluctuation remain valid for the outlet temperature redistribu-tion event. These have been presented at length in an earlier submittal to the Nuclear Regulatory Commission (Ref. 1). 2 5-1 4

An evaluation of the consequences of the outlet temperature redistribu-tion to core fuel temperatures has been made 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 temperature and nuclear channel measurements at the initiation of a fluctuation and those for the temperature redistribution have led to the conclusion that the redistribu-tion is a single motion event similar to that was which previously observed at the initiation of a fluctuation. These data have been presented in detail in the previous section; the similarities are listed below: 1 1. Sudden drop in core resistance, attributed to the combination of the gaps between regi.ons into a few larger gaps. 1 1 2. The temperature signatures of the thermocouples located in the gaps between core support blocks. The initial shape and magnitude of the temperature changes are nearly identical for several of these thermocouples. 3. Slight abrupt changes in the wide range nuclear channel signals caused by movement of the permanent side reflector columns opening neutron streaming paths. 4. Slight reactivity changes which can be explained by slight changes in the effective core diameter. 5. The '- gest region temperature changes occurring generally in the sam. regions, that is the regions in the NW sector of the core. 5-2

I 6. Temperatures measured by the region 35 instrumented control rod drive thermocouples located in the middle and bottom 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 (~14) reactivity change. Similar 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 radistribution. The aerupt 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 nornally following real reactivity changes was not observed. Small reactivity changes are observed and are explained by core displacement causing an effectively smaller core diameter. As discussed in the fluctuation safety analysis report (Ref. 1), the most extreme change in reactivity which could result from the core motion is to cotgress the core so as to close all available gaps. The resultant reduction in neutron leak-age amounts to only 0.00015 ok. In FSAR Section 14.2.1.3, the effects'of a 4 reactivitychangeof0.006akwereevaluatedandnodamagingeffectsw;re found. Therefore, no significant effects from these small reactivity

• changes due to core displacement are possible.

5.3.2. Control Rod Insertability 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. Rod insertion tekts were conducted using 1.6 in. misalignment at the insertion location and'2.5 it. misalignment in the rod channel; no appreciable increas in scram times was noted when compared to similar tests with the core aligned. With the RCDs installed only 5-3

a r negligible displacement can occur oc 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. The conclusion is that no predictable misalignment 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 Impact Velocities. In section-5.3.1.5.2 of Ref. 1, a . maximum fuel element impact velocity of 3 in./see was calculated. The ana-lytical model was one in which whole regions moved horizontally, driven by transverse pressure forces, until they impacted with neighboring regions. After the installation of the RCDs, this type of motion is no longer pos-sible, because the regions are restrained at the top. Another mode is possible, however, within the geometric constraints. In this mode a fuel column opens (jaws) at the interface between blocks to form a two-link mechanism. This mechanism moves in a mode where the top and bottom are stationary, while the middle hinge leads the motion. The transverse pressure differential which could form such a mechanism was first calculated. The corresponding location of the middle hinge was found to be about 104 inch (or less) from the top. It was then assumed that the middle hinge moves through the esti=ated largest gap to collide with an adjacent column. Using a conservative method and assuming 1007 core power condi-tions, the maximum 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 maximum possible impact velocity, if core motion occurs with the top restrained by the RCDs, is less than assumed in Ref. 1. Moreover, the mode of impact is also similar from the standpoint of causing impact loads. It can, therefore, be concluded that the structural loads are bounded by the results in Ref. 1, where they were shown to be small compared to the load capacity of the fuel elements. 5-4 - l

5.3.3.3. RCD Structural Loads. In section 5.3 of Ref. 5, 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. If core motion occurs as discussed above, 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 would be transferred to the column it leans on, and its associated RCD pin would experience a higher shear force. If it is assumed that the maximum pin force of 1,167 lb increases by 50%, which is very conservative (since most of the original force comes from a postulated leaning of seven columns in the same direction due to uneven irradiation shrinkage), the maximum stress in the pia would increase from 36,000 psi to 54,000 psi, still below the yield limit of 134,000 psi by a large margin. 5 3.4. Secondary System The secondary systems have been eliminated as the cause of fluctuations or the outlet temperature redistribution because steam temperature perturba-tions lag the primary side (helium) temperature pertt eations. Further, the helium temperature perturbations 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 the temperature redistribution event the steam generator modules experience a single temperature decrease of less than 10*F, and thereafter respond normally to the power rise. The single tem-perature decrease, which is believed due to the increasing the cold bypass flow, is of no cor. sequence to the fatigue stress limits in the steam -generator modules. I I 5-5 l

~l L 5.3.5. Bypass Flow Increase After the' Outlet Temperature Redistribution The bypass flow fraction was calculated to be 13% before the-outlet 3 temperature redistribution events, increasing to 14.5% during the tempera-i ture redistribution. The affect of this 1.5% increase is to increase maxi-mum core fuel temperatures by about 15'F and average core fuel temperatures by about 5'F. It should be noted that the bypass. flow fraction does not increase as power is increased from 70% to 100%. 5.3.6. Accident Analyses 4 FSV accident analyses (Chapter 14) were reviewed to determine if any reevaluation was required for plant operation after the-temperature redis-tribution. It was determined that the localized initial conditions created I by the outlet temperature redistribution would not' affect the accident con-sequences since the accidents are initiated at Technical Specification lim-its and operation following 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 effect a safe shutdown are neither associated with nor impaired by the temperature redistribution. I It was concluded thac no reevaluation of the FSAR accident analyses is required as a result of the observed outlet temperature redistribution. ( 5.4. LONG-TERM OPERATION WITH THE OUTLET TEMPERATURE REDISTRIBUTION The two outlet temperature redistribution events which occurred during fluctuation testing below 70% were nearly identical: the same regions ex-hibited temperature decreases during the power rise and the magnitude of the 4 temperature decreases was comparable. Also the core operating conditions of power, flow, and core pressure drop were very nearly the same. 4 J 4 5-6 --p ,r y v ,m. ym.---,.- w w i-p

In addition to the cooling effect of increased gap flow, there are two probable causes of the temperature decreases experienced by some' core bound-ary regions. These are cool bypass flows which enter the thermocouple probes at the side of the core support floor (Type II flow) and depress the thermocouple reading, and the opening of small horizontal gaps between the stacked columns in a region permitting cool bypass flow (crossflow) to enter the regions. Analysis and evaluation of the data indicate both of these phenomena are occurring, as explained in Section 4. Crossflows during the temperature redistribution are evidenced by the rapid temperature changes measured by the middle and bottom thermocouples 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. l l l The changes in steam generator codule helium inlet temperatures calculated assuming the Type II flow explanation generally compares more favorably with the measured temperature changes than if crossflow into these regions is assumed. The profiles of the temperature traverses across these regions at cor exit indicate the presence of Type II flow. Also, et 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 measured region outlet and steam generator module inlet helium temperatures yields a ~2% increase only if Type II flow is assumed to be affseting the region outlet temperature measurement. Further evidence of Type II flow is available from cycle 1 and cycle 2 operation. Type II flow best explains the differences between the measured and calculated steam generator module temperatures. For steam generator modules B l-5 and B 2-6 the calculated helium inlet temperature, based on the measured region outlet temperatures, is lower than measured. These mod-ules are adjacent to core regions 34, 37 and 20, which have measured region 5-7

peaking factors (i.e., exit gas temperctures) powers lower than predicted from core physics calculations. Type II flow provides a single explanation to resolve both of these dliferences. The regions which are most affected are those on the core boundary and at the end of thermocouple strings. These regions have the greatest potential for a significant Type II flow effect since the sleeves containing their region exit gas thermocouples are open to the core boundary where cool Type II flow is available. Additionally, during a loop isolation event some of these r' d ons have i shown a sharp temperature increase while the rest of the regions shcw 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 'or the Type II flow. Finally, during fluctuations, the thermocouples in some of these regions as well as some nearby gap th'rt ouples have dis-played temperature changes which do not correlate w he steam generator module helium inlet temperatures and which can br .ned by changes in the magnitudes and/or direction of the Type i; #1o s. The majority of the observations indicate Type II flow as the primary mechanism for the observed exit temperature reductions in the boundary regions. Nevertheless, a temperature analysis was performed for the region exit temperature redistribution event to determine the fuel temperature changes which could be caused by either of the two postulated mechanisms, i.e., Type II flow or jaws crossflow. Region 36 was selected because in both events it experienced the largest out'.et temperature decreases. In both cases core physics calculations ; --. sed to determine the region power after the region exit temperature redistribution. Then, for the case where Type II flow was assumed to enter the thermocouple probe, the fuel tempera-tures were calculated using this region power and the region flow rate inferred from the region flow control valve setting. For the case where cool crossflow is assumed-the effect of the cross-flow depends on its location and temperature. A larger fuel temperature increase occurs when the crossflow location is lower in the core and when 5-8 i

the temperature of the crossflow is high. Crossflow calculations were per-formed for crossflow locations 1/3,1/2 and 2/3 of the way down the active As discussed in Section 5.3.4.1, the probable location for a cross-core. flow gap to develop is about 104 in. or less from the top of the active core, or about 1/3 of the way down the active core. In fact, it is most probable the first crossflow gap developed near the top of the region. The additional calculations, at 1/2 and 2/3, were performed as conservative cases. For all the cases, the temperature of the crossflow was taken to be 407. of the region temperature rise at the location of the cross-flow gap. This is also a conservative assumption, 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 tempera-ture redistribution event is believed to create a combination of the core bypass gaps into a few larger gaps, and flow in these gaps would not be efficiently heated. With these conservative assumptions the crossflow and region flow rates were than determined from the core pressure drop and heat balance equations, using the region power from the core physics calculations. 4 For the case where the region 35 outlet temperature redistribution is assumed to be caused by a cool bypass flow entering the thermocouple probe (Type II flow), the redistribution event does not change the region fuel temperatures. For the case where cool crossflow into the region is assi 95, the effect of the crossflow is to decrease the peak fuel temperature from 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 temper-atures are decreased because the crossflow into the region reduces coolant and fuel temperatures in the bottom of the region, where peak fuel tempera-tures occur. The result is the peak fuel temperature location shifts to a joint just above the crossflow location (for the 1/2 and 2/3 location cases), and even though the temperature at this location is increased, it is lower than the fuel temperature at the bottom of the region without 5-9 i

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I / ( 7 Mid)L ' t" Yb f3OJG J7Lx r inD -** _yg i e r/ i i iV l j Fig. 5-1. Region 36 pe'k fuel temperature after the region exit temperature edistribution assuming Type II flow or crossflow 5-10

crossflow. For the 1/3 location case the peak fuel temperature remains at the bottom of the region since the fuel temperatures 1/3 of the way into the core are nominally low. The region exit temperature 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 primary coolant activity was observed. Since significant-fuel damage would result in an increase in the primary coolant system activity inventory, it is concluded that no sig-nificant fuel damage resulted from reactor operation during or after the outlet temperature redistribution events. i After both temperature redistribution events the core continued to be operated using the procedure that has been implemented 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 helium inlet and reheat steam temperatures of the twelve steam generator modules. This procedure ensures that those regions on the core boundary which are experiencing a Type II flow into their thermocouple probes continue to be operated within the technical spec-ification limits. This procedure also ensures that the interior regions are operated within the limits of the Technical Specifications both before and after a region exit temperature redistribution. This procedure will con-tinue-to be used after a temperature redistribution and for operation up to 100% power. 4 i 5-11

6. TESTING AB0VE 70% POWER WITH REGION CONSTRAINT DEVICES 6.1. DESCRIPTION OF TEST (RT-500) RT-500 is a test that has been used in determining the fluctuation threshold as a function of core pressure drop versus power (or flow). The cest was originally performed in November 1978 during cycle 1 operation as a part of the fluctuation test program. While this test has undergone numer-ous revisions as a result of testing experience and for compatibility with cycle 2 testing, the basic test philosophy has remained unchanged. Ba si-cally this test is a series of attempts to initiate fluctuations using load ~ increases of ~3% at 3%/ min as the initiating, or trigger, mechanism. The core is initially orificed for a specified core pressure drop at a specified power level. This is done in such a way that the region exit gas tempera-tures have adequate margins from LCO 4.1.7 limits and that the SG module temperatures are within specified operating limits. Stepwise load increases are continued until fluctuations "are encountered or 100% power is reached. Revision J of RT-500 (i.e., RT-500J) is the test to be performed during the initial rise in power from 70% toward 100% wherein attempts will be made to initiate fluctuations with RCDs installed. The major difference in this test and the previous version (RT-500H)* is that a revised Fig. 1 is pro-vided. These temperature margins, defined as the LC04.1.7 limit minus the region exit temperature, have been changed to be more appropriate for cycle 2 testing. The margins are based on the maximum region exit temperature increases observed during cycle 2 fluctuation testing with RCDs installed.

• The revised version of RT-500H is denoted as version J rather than I to avoid confusing the letter I with the numeral 1.

6-1

Other minor changes in RT-500J are: 1. The use of circulator speed pulses as a trigger =echanism are deleted since this technique has never been shown to triggar fluctuations. 2. The requirement for collecting data with'the regulating rod and with the trim valves in manual is deleted since neither has been shown to influence fluctuations. 3. The waiting period, after attempting to initiate fluctuations by a load increase, is reduced from two to one hours since fluctuations initiated by load increases have always been detected within min-utes of the increase, and a 20 minute waiting period is sufficient to achieve core thermal equilibrium. 6.2. TEST PROCEDURE RT-500J is the test which will be performed during the initial rise in power from 70% toward 100% wherein attempts will be made to induce fluctua-tions in order to evaluate the degree of success RCDs have on eliminating fluctuations. Performance of RT-500J at power levels greater than 70% will be in accordance with the B-0 startup test program. A test plan will coor-dinate the overall test program. The detailed text of the test, including the safety evaluation, operating limits, data systems required and the test procedure, is given in Appendix A. 6-2

7. REFERENCES 1. " Safety Evaluation - Reactor Outlet Temperature Fluctuations," P-78137, August 11, 1978. 2. GA RT-500, Revision H, December 1979 - PSC Submitted 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 Submittal to NRC P-79287, November 28, 1979. 5. "SAR for Core Region Constraint Devices," P-79068, March 23, 1979, i O 4 7-1

APPENDIX A: RT-500J GA RT - 500 Revision J RT-500J POT REF NA Sheet 1 of REE REF NA DATE ISSUED _ REQUEST.FOR TEST REQUESTOR K. Asmussen/W. Simon SYSTEM 12 PURPOSE /0BJECTIVE - There are two main objectives of this test: 1. To determine the fluctuation threshold in terms of core pressure drop vs flow (power) for cycle 2 operation af ter installation of region constraint devices. 2. To obtain FM data during fluctuations (if they occur) with the revised instrumentation systems for comparison with cycle 2 data without region constraint devices. DESCRIPTION OF TEST - With the plant in normal operation, core orifices will be adjusted to achieve a specified core pressure drop. Load increases in ~3% steps will be used as trigger mechanisms to induce fluctuations and to determine the fluctuation threshold in terms of core pressure drop as a function of core flow rate (power level). If a fluctuation occurs, the step causing the fluctuation will be repeated to demonstrate repeatability. Attempts to initiate fluctuations will be performed first at ~40% power and then at ~10% intervals so as to provide a good definition of the stability threshold line. Part 1 of tho test encompases testing at <70% power while Part 2 refers to testing at >70% power. Revision E incorporates PSC comments on Revision D. REF SOP OR ABNORMAL CONDITIONS - SOP 12-04 SCHEDULE REQUIREMENTS - SAR 7 APPROVAL SHEET ATTACHED WORK ASSIGNED BY TO Name/Date EVALUATION COMPLETED Name/Date REVIEWED BY Nace/Date Distribution: K. Amussen E. Hill A. Kennedy F. Mathie R. Phelp G. Bramblett W. Franek J. Lopez R. Nirchl V. Tersie Requestor: GA-SD Ref. Library A-1

P GA-RT - 500 Revision J Sheet 2 of REQUEST FOR TEST (Continued) Revision F incorporates NRC comments on Revision E. Revision G incorporates a revised definition of a fluctuation and increases the estimated time spent in a fluctuating mode. These revisions are based on experience gained from RT-500F testing at 40% and 50% power. Revision H provi as for fluctuation testing with the Region Constraint Devices (RCDs) installed. Revision J deletes using the cire speed pulse as a trigger mechanism and the requirement for collecting fluctuation data with the reg rod and with the trim valves in manual. The waiting period after attempting to initiate fluctuations by a load increase is reduced from 2 hours to I hour. A new Fig. 1 is provided which is more consistent with cycle 2 test results after installation of RCDs. ANTICIPATED RESULTS/ ACCEPTANCE CRITERIA - The test will provide data to aid in predicting conditions for stable operation. Additionally, the test will demonstrate the effect of RCDs on fluctuations, and data will be obtained which will aid in understanding the fluctuation phenomenon and for compari-son with previous cycle 2 observations. There are no specific anticipated results or acceptance criteria. A-2

p This SAR is in support of RT-500J and is enclosed as Page 3 of of that RT even though the SAR is a separate uncontrolled document. RT-500 Rev. J Sheet of GENERAL ATOMIC COMPANY FORT ST. VRAIN NUCLEAR GENERATION STATION SAFETY ANALYSIS REPORT 1. INITIATING DOCUMENT: RT-500 Revision J 2. CATEGORY: Yes No PLANT CHANG 5.' x DOCUMENT CHANGE ONLY CLASS I x MAINTENANCE SAFE SHUTDOWN COOLING x TEST x 3. FAILURE MODES AFFECTED x 4. SAFETY RELATED COMPONENT, x i SYSTEM OR STRUCTURE CHANGE STATE IN ITEM 10 THE l 5. SAFETY SIGNIFICANT CHANGE x E 6. UNREVIEWED SAFETY QUESTION x 7. TECH SPECIFICATION CHANGE x 8. FSAR CHANGE x 9. APPLICABLE FSAR OR TECH SPEC SECTIONS RTVIEWED: T.S. LCO 4.1.4 and LCO 4.1.7 10. BASIS FOR SAFETY EVALUATION: (add additional sheets if required): See Attached Sheets 11. IS SAN DIEGO SAFETY ANALYSIS / LICENSING REVIEW REQUIRED? YES _ NO x 12. HAS SAN DIEGO SAFETY ANALYSIS / LICENSING REVIEW BEEN PERFORMED? YES x NO 13. INITI.10R/DATE LICENSING /DATE 14. GAC ENGR. REVIEW / DISPOSITION: I concur with the above safety evaluation ENGINEER /DATE A-3

/ RT-500 Rev. J Sheet 4 of 10. BASIS FOR SAFETY EVALUATION (Continued) During this test, operation will be within Tech Spec and design limits. The test is performed to investigate the influence of Region Constraint Devices on the threshold for fluctuations, and although they might be encountered, the time at fluctuation would be minimal. Consequently, any fatigue damage to the core and/or steam generators would be negligible. Main steam temperature fluctuation limits have been set accordingly. In_ addition, throughout the test, sufficient margins will'.se maintained between the Tech Spec limit and the region exit temperatures so that even during the maximum temperature increases observed during fluctuation testing with RCDs installed, the Tech spec limits will not be violated. Therefore, this test will not adversely affect the intcg'rity of the core or steam generator and will not affect public safety. Since first being encountered on October 31, 1977, fluctuations have been initiated several times in a continuing effort to understand their The power levels at which fluctuations were initiated have ranged cause. from 30% to 68%. A total of 65 hours was spent in a fluctuating mode in Cycle 1 and another 41 hours in Cycle 2. This is equivalent to about 640 cycles with an average period of about 10 minutes. Although the exact cause of the fluctuations is not known, there are several reasons for concluding that continued testing is safe. The total core power, flow, and average temperatures are relatively stable. An in-spection of the top plenum in December 1977 after fluctuation testing at power levels between 53% and 68% showed it was in good condition. An in-spection at that time of the control rods in region 34 (which were inserted i throughout the entire period of fluctuation testing) also showed no signs of excessive temperature or impact. During the first refueling outage, eleven blocks from region 35 were carefully inspected in the PSC hot service facil-ity and there.was no evidence of damage. An in-core inspection of region 35 A-4

RT-500 Rev. J Sheet 5 of and its surroundings with the fuel handling machine TV camera revealed no damage or excessive wear to any component. The upper plenum area looked fine; the gaps in the regions and side reflector surrounding the cavity left when region 35 was unloaded were very, regular with no evidence of wear or damage. An inspection of the core support blocks in regions 13 and 35 have likewise revealed no da= age. Every element removed from the core during the refueling of six regions has been photographed as has each block in five additional regions which were unloaded to permit installation of test assemblies. Examinations of these photographs have revealed no damage. Fluctuations have been initiated and observed on seven (7) occasions during cycle 2 at power levels between 38% and 63%. Four (4) of these were initiated by power increases and three (3) :ommenced during orifice valve adjustments. The data from the in pile test program has demonstrated that while the fluctuations encountered during cycle 2 appear to-be somewhat more regular and widespread throughout the core, it is basically the same phenomenon experienced during cycle 1. Eighty-four Region Constraint Devices were added to the top layer of hexagonal elements (keyed plenum elements) during the October-November, 1979 outage. These mechanical links are placed at locations in the core where three regions intersect and will provide inter-region keying and preclude the accumulation of large sized gaps which might result if several regions are displaced in the same direction. These have been installed to provide a permanent solution to the core temperature fluctuations that have previously occurred. In November and December 1980, fluctuation testing was performed with the RCDs installed. No fluctuations were detected at power levels up to 70% A-5 m

RT-500 Rev. J Sheet of and at core pressure drops up to 4.1 psid. It is important to note that most of the testing was above the fluctuation threshold line. 1 It has been shown that the installation of the RCDs has had no impact on the nuclear design or nuclear performance of the core at power levels up to 70%. Testing Above 70% Power (Part 2) Since fluctuations were first encountered, tests have been conducted under various core conditions. In large part, these tests were designed to gather specific information on what key parameter or combination of param-eters leads to the fluctuations, since this kncvledge could be instrumental in understanding their cause. These tests have shown fairly conclusively that power level is not by itself a parameter of primary importance to the fluctuation thresho1.d, and they have established core pressure drops as a key parameter, probably closely related to the cause ci the fluctuations. Another result from these tests is that it appears that the core pressure drop at which fluctuations are produced is highar at higher core power levels. Differences in fluctuation msgnitudes and character have been observed in the fluctuations that were initiated during cycle 1 and cycle 2 opera-tion. These aifferences have been carefully studied and reported exten-sively. No apparent correlation with power level has been noted, nor has a change been observed with time that would indicate increasing fluctuation magnitudes ot significant differences in character. All of the fluctuation testing limits and operating considerations as well as normal plant tech-nical specification limits and SOPS are in effect both below 70% power and above 70% power. One exception is the limit on nuclear detector fluctua-tions. This limit is increased from 1G% at <50% power to 20% for >70% A-6

RT-500 Rev. J Sheet of power. This increase is justified because nuclear channel fluctuations are believed to be due primarily to a streaming effect and are thus expected to be nearly proportional to the power (neutron flux) level. In this test, attempts will be made to initiate fluctuations at successively higher power levels. The magnitude and character of the fluctuations at each power level will be carefully observed for differences in addition to monitoring the Technical Specification and fluctuation testing limits. Consequently, testing above 70% will not affect public safety. Time Spent Fluctuating It is anticipated that so additional temperature fluctuations will be encountered with the RCDs installed. The purpose of this test is to demon-strate that fact or to determine what impact their installation has on the fluctuation threshold. In the event that fluctuations are again observed, it is anticipated that operation in the fluctuating mode would be limited to about 15 hours for completion of part 1 (<70% power) and another 10 hours for part 2 (70% to 100% power). This is based on previous test operating experience. It assumes that fluctuations are initiated from 2 to 4 times at 3 power levels or core configurations during part 1 and another 2 power levels or core configurations during part 2. A-7

~ U: ~ RT-500 Rev. J Sheet of PREFACE Revision A accomplished the following changes: .1. The Operating Limits Section was redefined to incorporate limits I required by the NRC. 2. The remainder of the previous litics were redesignated as-Operating Considerations. ^ 3. The equation for core resistance was redefined to better fit observed operating data. 4 e 4. Addendum I was added to determine the fluctuation threshold at 28% { power. Revision B accomplished the following changes: 1. The 10% limit on a nuclear channel fluctuation was extended to cover all six channels. 2. The required instrumentation was increased to have brush recorders for all twelve steam generator module outlet temperatures and all six nuclear channels; the steam generator temperatures will'be. monitored both by wide range brush recorders (700*F to 1100*F) and by either narro'w range brush recorders (100*F range,.zero suppressed) or digital display of fluctuation magnitudes from the steam generator data acquisition system. i 3.' The limit on-module main steam temperature at which testing is suspended until authorized by PSC management is increased from t 230*F to 150*F. A-8 l 4 -.. ~. ,, -. -, _.. _.. ~. __

? fl. i RT-500 Rev. J' Sheet of 4. In Fig. 1, the region temperature mismatch margin for region 12 is increased to 100*F.

5..

The instruments to be monitored by the trend recorders are not specified: any four thought to be of most use may be trended. 6. A two hour waiting period between fluctuation tests is specified. Revision C accomplished the following changes: 1. Corrective action is to be taken to stop the fluctuation if a module main steam temperature reaches 1025*F. 2. Editorial changes were made to the other limits on module main steam temperature. I 3. The test team members responsible for conducting the test are specified. 4. The physical location of the data systems to be monitored are specified, as are the respective team members responsible for monitoring them. Figure 2 of Addendum I has been " cleaned up" and updated to i reflect the current actual locations for thermocouples 3, 4, 5, 7, 19, 23, and 25. 6. In Fig. 1, the region temperature mismatch margin on regions 17, 18, 26, and 27 have been increased. A-9

RT-500 Rev. J Sheet of Revisicn D accomplished the following changes: 1 1. The detailed test procedure has been rewritten. The number of anticipated fluctuations and the total time spent in the fluctuation mode has been reduced. However, the basic test philosophy and the limits during fluctuatione remain unchanged. 2. RT-502 (Threshold Testing >70% Power) has been incorporated as Part 2 of this RT. 3. The objectives of the test have been modified to reflect testing during cycle 2 (for comparison with cycle 1) with the emphasis on gathering data to aid in predicting conditions for stable operation. 4. Addendum I of RT-500 Revision C has not been repeated here because it was successfully completed during cycle 1 testing. 5. A definition of a fluctuation has been included. 6. There have been numerous editorial changes (changes a e denoted by a D in the margin). Revisi'r E incorporates comments from PSC to delete the. tailed orifice adjustment procedure, update Data Sheet 1 to include all limits and other minor comments as noted in the left margin. Revision F incorporates comme..es from NRC as noted. Revision G incorporates a revised definition of a fluctuation and increases the estimated time spent in a fluctuation mode. These revisions A-10

RT-500 Rev. J Sheet of . nre based on experience gained f com RI-500F cesting at 40%.and 50% power. The increased time is the result of approximately 15 hours spent in a fluctuation mode during completion of the.first' half of RT-500F. ~ Revision H provides procedures for fluctuation threshold testing up to 100% reactor power af ter installation of the RCDs. In essence the test is a continuation of that previously performed in cycle 2. After establishing the core configuration for which fluctuations were initiated at the lowest power level during previous cycle 2 operation, an attempt will be made to initiate fluctuations by a power level increase. The power level will be increased incrementally up to 70% and if no fluctuations are encountered, the power level will be reduced to 40%, and the core AP increased by adjust-ing the orifice valve opening.- This will be repeated u,ntil a stable power operation at a power level of 70% with a core AP of about 4.5'is achieved or a fluctuation threshold established. Part 2 would be a continuation of this operation up to 100% power. Other than the method for determining system operating lines, the conditions and controls of this test are the same as those for the previous RT-500G. Test prerequisites, administrative con-trols, system and operating limita, and reporting requirements are all unchanged. Revision J deletes using the circulator speed pulse as a trigger mech-anism, since this has never been shown to trigger fluctuations. Since the 4 j reg rod and trim valves are not believed to influence fluctuations, the re-quirement for collecting data with the reg rod and trim valves in manual is deleted. The waiting period, after attempting to initiate fluctuations by a load increase, is reduced from 2 hours to 1 hour since fluctuations initi-ated by a load increase have always been detected within minutes following the loed increase.- In addition, a 20 minute waiting period is sufficient to achieve core thermal equilibrium. The temperature limits in Fig. I are revised to be more appropriate for cycle 2. These limits are based on the A _

f RT-500 Rev. J Sheet of maximum temperature increases observed during cycle 2 fluctuation testing with the RCDs installed. INTRODUCTION The collection of data from all cycle 1 core fluctuations indicates a distinct influence of core AP on the threshold for fluctuations.

However, the cyle 1 data shows a lot of scatter and fresh fuel has been loaded into six regions of the core; therefore, fluctuation threshold testing at the beginning of cycle 2 is necessary.

This RT will determine the effect of RCDs on the fluctuation threshold as a function of core pressure drop for cycle 2 (with a procedure aimed at minimizing the amount of scatter in the data). The FM Data System now in-cludes 24 traversable thermocouples, and two instrumented control and ori-ficing assemblies having in-core instrument packages. These data will be collected during fluctuations to aid in predicting conditions for stable operation and understanoing the fluctuation phenomenon. The method for triggering a fluctuation will be a 3% load inn;aase at 3% per-minute. If a fluctuation develops, the steps preceding and resulting in the fluctuation will be repeated to demonstrate repeatability and to provide a reasonably accurate determination of the threshold power. The test scope includes attempts to induce fluctuations for at least three values of core flow resistance. [ Testing will be conducted by the coordinated efforts of a test teca consisting of, but not limited to, the following members: l l 1. PSC Shift Supervisor i' 2. PSC Reactor Operator (s) l A-12

q RT-500 Rev. J' Sheet of 3. Test Coordinator 4. Core Performance Engineer 5. Data Systems Engineer The NRC will be provided, within one week, with a summary of test resulta for eact. power level. Included with these results will be notifica-tion of any change to the procedure as a result of the test results. OPERATING CONSIDERATIONS In addition to the normal plant operating procedures and limitations, the following should be observed:

1.

• T'he HRH and MS temperature imbalance between each SG module and the average for the loop should not exceed 30*F (in steady state) or the limits given in SOP 12-04, whichever are more restrictive.

In addition, the maximum individual module MS steady-state temper-ature should be limited to 995'F. The purpose of the 995*F limit is to provide margin on MS temperature should fluctuations occur. 2. Steady-state module helium inlet temperature shall be limited to 145'F about the mean or the limits given in SOP 12-04, whichever tre more restrictive. 3. In order to minimize the chance of getting into a degraded per-formance condition during fluctuation testing with RCDs installed, the maximum region outlet gas temperature during steady state conditions shall be limited as shown in Fig. 1. NOTE: The margins per Fig. I are based on data obtained during cycle 2 fluctuation t? sting with RCDs installed. A-13

i-RT-500 Rev. J Sheet of 4. The plant is defined to be in a fluctuation operating mode-when individual nuclear channels exhibit cyclic deviations from the average power equal to or greater than 0.5%. peak-to peak of full power not exceeding a 30-minute period. 5. Operation and/or testing at power levels >70% should be in accordance with the B-0 startup test program. 6. Throughout the duration of this RT, all plant control systems are to be in automatic, and with MS and HRH temperature controls set to a maximum of 980*F. NOTE: The reason for selecting the temperature setpoint at 980*F versus 1000*F is to allow margin for the temperature swings that may occur if fluctuations develop. i, LIMITS DURING FLUCTUATION TESTING f I Test Limits 1. Proposed testing will be conducted within the technical specification limits. 2. Throughout the test, the intent will be to minimize the time spent in fluctuation except when necessary to record FM data. When fluctuations are present, the following should be observed: A tempeature fluctuation of module main steam temperature a. about its mean of 10*F (20*F total amplitude) is acceptable with no specific time considerations. A-14 rwar =-- y-r

4 A RT-500 Rev. J Sheet of b. A temperature fluctuation of module main steam temperature about its mean greater than 110*F (20*F total amplitude) but less than 30*F (60*F total amplitude) should not exceed ane hour in duration per event. c. A temperature fluctuation of module main steam temperature about its mean of t30*F (60*F total amplitude) is cause to take immediate corrective action by reducing power to stop the fluctuations. d. A unidirectional module main steam temperature change of 6)*F' (excluding the average component of intentional steam temper-ature changes) is cause to take immediate corrective action by reducing power to stop the fluctuations. A module main steam temperature of 1025'F is cause to take e. immediate corrective action by reducing power to stop the fluctuation. f. A primary coolant activity increase greater than a factor of 25% but less than a factor of 5 over the prior equilibrium value for that power level is sause to take immediate corree-tive action by reducing powar to stop the fluctuations. 3. A limit of 10% of full power range on any nuclear channel will be maintained. 4. The helium purification system will be in service during all testing. A-15

RT-500 Rev. J Sheet of 5. An increase in primary coolant activity levels greater than a factor of five (5)-over prior equilibrium values for that power level during any fluctuation test will be cause for terminating the testing and proceeding with an orderly plant shutdown. Corrective Action 1. If any of the established limits or conditions outlined in items 2 through 4 above are exceeded during a fluctuation test, the test will be terminated, and further plant testing in the fluctuating mode will be suspended until specifically authorized by PSC management. 2. If any of the following conditions are exceeded, immediate action will be taken to terminate the fluctuation test and further testing in fluctuacion mode will be suspenced until authorization to proceed is obtained from the Commission: Technical Specification limits are exceeded. a. b. An increase in primary coolant activity levels greater than a factor of five (5) over prior equilibrium values for that power level. A temperature change of module main steam temperature of c. 150*F relative to the initial steady-state temperature and exclusive of temperature change due to load changes. d. A module main steam temperature which exceeds 1075*F. A-16 (

f RT-500 Rev. J Sheet of 3. If inadvertent fluctuations are observed (see operating consider-ation item 4 for the definition of a fluctuation) in normal opera-tion, corrective action will be taken to terminate the fluctua-tion, and PSC management authorization will be required prior to returning to a power level that would approach that level at which the inadvertent fluctuations were observed. INSTRUMENTATION / DATA SYSTEMS Through the duration of this test, the following data systems shall be operating and personnel should be present for monitoring: 1. Brush recorders (located in the auxiliary control room) with all steam generator module main steam outlet temperatures add nuclear channels. A data system engineer will be present to monitor the recorders. Both wide range brush recorders (700*F to 1100*F) and either narrow range brute recorders (100*F, zero suppressed) or digital display by the steam generator data acquisition system will be available to monitor the steam generator module main steam outlet temperatures. 2. Data logger (located in the control room). The core performance engineer will be present to monitor the core temperature limiting conditions for operation. 3. Th'e primary coolant activity monitor (located in the control room). { 4. FM data acquisition system. A-17

w RT-500 Rev. J Sheet of If any of the above systems becomes inoperable, testing shall be halted until the system is reinstated. If fluctuations are ncountered when any of these systems is inoperable, core power should be reduced until the fluctuations cease. During power increases and for a period of 1 hour fo' lowing the power increase, the following data system and data taking frequencies are desired: 1. Data logger on a fast sample rate (15 sec or less). 2. Steam generator Fox II computer on a fast sample rate (~5 sec). 3. Model verification computer. 4. FM data acquisition system. 5. Brush recorders. At periods during the test when the initial conditions for a fluctuation test are being established (orifice adjustments, flow / power changes), the following data systems and data taking frequencies are desired. 1. Data logger on a sample rate of 2 m!n or faster. 2. Steam generator Fox II computer on a smaple rate of 15 see or faster. 3. Model verification computer. 4. FM data acquisition system. 5. Brush recorders. The traversable thermocouples are to be pocit.oned per RT-524. d s- 'l l A-18

s o RT-300 Rev. J Sheet of PART 1: TESTING AT <70% POWER Initial Conditions l. Plant at approximately 40% power. 2. The orifices are to be adjusted such that the region exit tem-peratures and steam generator inlet temperatures are balanced per normal procedures and per the Operation Considerations section of this RT. Procedure 1. The objectiv'e of this test is to demonstrate that no fluctuations 1 occur with RCDs installed or to develop a core pressure drop vp core flow rate (or power) stability threshold line. Thus it is desired to attempt to initiate fluctuations at three or more values of core flow rate. This will be done by orificing the core to different flow resistances. Depending on the core flow rate at which fluctuations are initiated in the first test, higher or lower values of core resistance may be selected. To generate a reasonably good stability threshold line it is desired to initiate fluctuations at a lowest power level of about 40% to 50% and at about every 10% increase in power thereafter. NOTES: a. Each time fluctuations are initiated, Data Sheet 1 must be completed. b. The most effective means of halting fluctuations is by power reduction. Experience has shown that to halt a fluctuation the A-19

RT-500 Rev. J Sheet of power may have to be reduced by 5% to 10% below the power level which produced the fluctuation. c. Wait at Ia..t 1/2 netr.co reach thermal equilibrium prior to performing any fluctuation test, wait I hour after attempting to initiate a fluctuatior by a load increase before continuing. 2. For the first test configuration, adjust the core orifices in a series of steps to obtain an average core pressue drop of about 1.7 psi with a core resistance of about 46. Core resistance cor-responding to this core pressure drop and core flow rate may be calculated by: 13 AP Measured + Pressure Resistance = 2.0 10 Flow.85.TT460 1 in where AP measures is the measured core pressure drop in psi (item-71 in DF 76) Pressue is the circulator inlet helium pressure in psi (item 9 in DF 76) Flow is the total circulator flow in ibm /hr (item 72 in DF 76) T is the circulator inlet temperature in degrees Fahreheit in (average of items 1 and 2 in DF 76) Verify that the region outlet gas temperatures have adequate j margin from LCO 4.1.7 per Fig.1 and that the SG module temperatures are within the Operating Limits Section of this RT before proceeding. r A-20

RI-500 Rev. J Sheet of 3. Begin a series of power rises-by increasing turbine load at 3% per minute in incremental load changes of 3% (~9 MWe). Continue the incremental load increases until fluctuations develop or a-plant limit is reached. Prior to each incremental load increase, adjust orifices'as necessary to balance region outlet gas temperaturees and module inlet gas temperatures. In addition, adjust the reg-rod position according to normal operations practice. If 70% power is reached and no fluctuations have been encountered, reduce the power to 40% and begin a series of orifice adjustments to in- . crease the core AP to about 2.2 psi. The orifices will be closed incrementally so that approximately the same regional core flow distribution is maintained. Verify that the region outlet gas temperatures have adequate margin from LCO 4.1.7 per Fig.1 and that the steam generator module temperatures are within the oper- ] ating limits. Begin another series of power increases by increas-ing the turbine load at 3% per min in increments of about 3%. Continue these power increases until a power level of 70% is achieved or fluctuations are encountered. Repeat i:he above procedure until a stable operation at 70% power and/or a core AP of about 4.5 psid is achieved. i. 4. If fluctuations develop, there are three basic sets of data to 4 obtain: It is desirable to obtain FM data during the onset of a. all fluctuations. b. For one fluctuation with each core flow resistance, obtain FM data for one hour during the fluctuations. T A-21

RT-500 Rev. J Sheet of The operating limits stated in this RT must be adhe:ed to during the one-hour period. c. Deleted. 5. For each fluctuation encountered, repeat the step preceding the fluctuation and, if fluctuations are not encountered, that step which cuased the fluctuation. For example, if fluctuations are encountered during a power rise from 50% to 53% power, return to 47% power and repeat the 47% to 50% power rise. If no fluctuations occur, then repeat the 50% to 53% power rise. 6. When a fluctuation threchold has been defined per steps 2 through 5, return to the highest power level for which a fluctuation was -not initiated-{^7Y in the-above example). 9 7. Deleted. 8. The next test power level depends on the power level at which fluctuations were encountered in step 5 above. The objective is to initiate fluctuations at power levels approximately 10% apart; that is, at about 40%, 50%, 60% and 70% power. For the selected new power level the next value of core resistance can be calculated from the conditions which initiated the preceding fluctuation as follows: 2 [ FOLD \\ RNEW = ROLD (FNEW/ A-22

~..

y..

RT,500 Rev. J Sheet of where Roto is the resistance from the. preceding test. Foto is the core flow rate from the preceding test (step 5)'. FNEW is the flow rate, corresponding to the power. level where the next fluctuation is desired. 9. The starting point for the next test is with the core orificed.to-achieve the new. core resistance, RNEW, Per the equation given in step 2, and with a core pressure drop - 10% to 20% below that at which the preceding fluctuation was initiated. In getting to the new starting point, it is desired to keep the core pressure drop at or below the value at which the fluctuation test will be started to prevent inadvertent fluctuations. To do this it is suggested that: ~~ a. IT R IROLD, reduce flow before closing orifices. NEW b. If RNEW < ROLD* OPen orifices before increasing flow. 10. Repeat steps three through seven to obtain data for at least three values of. core resistance. To generate a reasonably accurate stability threshold, fluctuations should be initiated at a lowest nower level of about 40% to 50% and at increments of ~10% power Je this initial level. Depending upon the effect of RCDs on the fluctuation threshold, it may be necessary to vary attemperation flow (core P/F), region outlet temperature mismatches, or ' partially insert control rods (power flattening) in order to demonstrate the threshold at high power' levels. Any or all of these operations may be used as per-mitted by SOPS and Technical Specifications. Caution should be exercised to maintain the region outlet temperature margins for A-23

RT-500 Rev. J Sheet of l LCO 4.1.7 given in Fig. 1 and to not violate che,coru thermal safety limit on ' core power / flow ratio (SL 3.1, Fig. 3.1-2). PART 2: TESTING AT >70% POWER 4 From an initial steady-state. condition of ~70% power, the core power will be increased slowly (1/2% per min) to ~73% and stabilized. If no fluc-tuations occur, ' power will be reduced to 70%, stable operation achieved, and a 3% load increase at 3% per minute will be effected to attempt to trigger-fluctuations. This process of slow power increases, and then rapid power increases of 3% will be continued until fluctuations are ancountered or until 100% power or a plant limit is encountered. If fluctuations occur, data will be recorded for a short period of time and the step which initi-ated the fluctuation will be repeated to establish reproducibility of the onset of fluctuations. Initial. Conditions 1. Plant at approximately 70% power. 2. The orifices are to be adjusted such that the region exit tempera-tures and steam generator inlet temperatures are balanced per nor. mal procedures and per the Operation Considerations section of this RT. Procedure j 1. The objective of this test is to extand the core pressure drop vs-core flow rate (or power) stability threshold line developed in Part 1. A-24 ~ i ]

RT-500 Rev. J Sheet of NOTES: a. Prior to each incremental load increase, adjust orifices as necessary to balance region outlet gas temperatures and module gas temperatures. In addition, adjust the reg-rod position according to normal testing practice. b. The most effective means of halting a fluctuation is by power reduction. Experience has shown that to halt a fluctuation the power may have to be reduced to 5% to 10% below the power level which produced the fluctuation. c. Each time a fluctuation is initiated, Data Sheet 1 must be completed. d. Wait at least 1/2 hour to reach thermal equilibrium prior to performing any fluctuation test; wait 1 hour af ter attempting to initiate a fluctuation by a load increase before continuing. 2. Adjust the core orifices for normal operation, within the limits of Fig. 1. The core resistance corresponding to this core pres-sure drop and core flow rate may be calculated by the equation given in Part 1 procedure step 2. If the core orifices are opened as much as possible, the main steam temperature may be reduced to 40*F below the reheat tempera-ture setpoint, core control rods may be partially inserted (to flatten the power distribution and thereby permit further opening of orifices), or attemperation flow may be increased within the limits of SOPS and Technical Specifications to further reduce the 4 core pressure drop. A-25

RT-500 Rev. J Sheet of Verify that the region outlet gas temperatures have adequate margin from LCO 4.1.7 per Fig. 1 and that the SG module tempera-tures are within the Operating Limits Section of this RT before proceeding. 3. Increase power by -G% (~9 MWe) at 1/2% per min. If fluctuations do develop, go to step 6. If fluctuations are not initiated, decrease power by ~3% to achieve initial conditions once again. 4. Deleted. 5. If fluctuations are not initiated by step 3, increase power by ~3% at ~3% per min. If fluctuations develop, go to step 6. If flue-tuations do not develop, r2 peat steps 3 through 5 starting at the new power level (~G% above the preceding power Jevel). Continue with successively higher power levels until fluctuations do develop or until 100% uover or a plant limit is encountered. 6. When fluctuations develop, there are two basic sets of data to obtain: It is desirable to obtain FM data during the onset of all a. fluctuations. b. Obtain FM data for one hour during the fluctuations. The operating limits stated in this RT must be adhered to during the one-hour period. 7. For each fluctuation encountered, repeat the step preceding the fluctuation and if no fluctuations occur, that step which caused the fluctuations. These repeated power increases will be at A-26 u_

RT-500 Rev. J Sheet of ~3%/ minute. For example, if fluctuations are encountered during a power rise from 73% to 76% power,' return to 70% power and repeat-the 70% to 73% power _ rise, then if no fluctuations occur, repeat the 73% to 76% power rise. 8. Based on the results of the test, a determination will be made as to the necessity to repeat any portions of the test or two perform any additional tests utilizing different orifice patterns or core resistance values. If further testing is to be continued, proceed to step 9. 9. Repeat steps 3 through 7. 10. Establish normal operating conditions. A. w-g1--- .e ~ t ,r

4 RT-300 Rev. J Sheet 17 of 19 t / ,/ n n 20 10 a 2, 20 20 so 80 , 20 A

D 0

D

[9 1 30 n m 0 b0 20 20 0 ,s. f _.0 u a 1 u ~ u to ,, g l g

o g

/D 10 20 w m 5D u 20 c m a ,g ^O ^O ^ / I 4 Fig. 1. Required temperature margins (LCO 4.1.7 limit - TEET) ) A-28

,w oO } k i SHEET 18 IS DELETED A-29

, ~, 4;e .e-RT-500 Rev. J Sheet of DATA SHEET 1 Complete this data sheet if fluctuations were encountered. This data sheet is to certify that test limits were not exceeded. A. If any of the following limits are exceeded, testing must be stopped until further aathorization by PSC Management. Limit 1. Were OPERATING CONSIDERATIONS 1, 2 and 3 met? N/A Yes /NO 2. Time /date fluctuation started N/A 3. Power level at start of fluctuation ~ N/A 4. Time /date power reduced I hour 5. Time /date fluctuation stopped N/A 6. Power level when fluctuation stopped N/A 7. Maximum fluctuation on Nuclear Channel # N/A Peak Magnitude 110% (<70%) 120% (>70%) 8. Maximum fluctuation of Loop I MS Temp Module # N/A Fluctuation Magnitude 60*F P-P Hottest Module #

• F 1025'F j

9. Maximum fluctuation of Loop Il MS Temp Module # N/A Fluctuation Magnitude 60*F P-P Hottest Module # 1025*F ] 10. Equilibrium value of primary coolant activity for power level of test N/A 11. Maximum value of primary coolant activity i during test 25% increase ~, 12. Were data systems in service? Yes /No Required for testing 13. Was a purification train in service? Required for Yes /No testing D. For any of the following, testing must be stopped and reported to the NRC: 1. Any technical specification exceeded? Yes /No 2. Any MS temperature fluctuation > 150*2? Yes /No 3. Primary coolant activity >5 times normal? Yes /No (This requires an immediate orderly shutdown.) PSC SHIFT SUPERVISOR Signature /Date TEST COORDINATOR Signature /Date .{ l A-30 - - -. -}}