ML19350A429
| ML19350A429 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 03/05/1981 |
| From: | TOLEDO EDISON CO. |
| To: | |
| Shared Package | |
| ML19350A426 | List: |
| References | |
| NUDOCS 8103160159 | |
| Download: ML19350A429 (12) | |
Text
{{#Wiki_filter:__ _ _ _ _ _ i t O. POST REFUELING PHYSICS TESTING , After the core reload for Cycle 2, where 25% of the core was replaced, testing was j performed to assure that the core performance was within acceptable limits of the predicted results. This was accomplished through zero power physics testing and power escalation testing, j The Zero Power Physics Testing, ST 5010.03, was conducted to: (1) Confirm the nuclear design characteristics of the core, i ~ (2) Validate the assumptions made in the safety analysis, and , (3) Validate the analyticci models used for predicting plant responses The final step towards criticality consisted of deborating the Reactor Coolant System (RCS) from 1685 ppm boron to 1280 ppm. Source range instrument N1-2 was used to monitor neutron flux and produce an inverse multiplication (1/M) plot. Boron concen-l trations in the RCS, makeup tank and pressurizer were determined by chemical analysis periodically during deboration. 4 , Deboration began from essentially an "all rods out" configuration at a constant feed and bleed rate of about 70 gpm. Count rates were obtained every fifteen minutes and plotted as inverse multiplication values versus gallons of demineralized water added.
- In addition, the RCS boron concentration was sampled every 30 minutes during debora-tion. Inverse multiplication values (1/M) were also plotted versus these boron con-centrations. Both plots are in Figures 1 and 2.
l When a 1/M value of 0.03 was reached, the deboration process w a stopped and critica-l lity verified by pulling control rod group 7 to achieve a constant startup rate. ! Reactor power was allowed to increase to approximately 10-8 amps on the intermediate range nuclear instrumentation and maintained at this level using control rods. Dur-ing the increase, the overlap between the source range and intermediate range instru-ment channels was verified to be at least one decade in accordance with Technical Specification 4.3.1.1.1. The observed overlap was approximately two decades. Maxi-
- mum letdown flow to the makeup tank was established to equalize the boron concentra-l tions of the RCS, makeup tank, and pressurizer. Equilibrium conditions were finally
, established at an RCS boron concentration of 1280 ppm with control rod group 7 at 90% withdrawn. i All Rods Out Critical Boron Concentration ' The all rods out (ARO) critical boron concentration at het zero powei (HZP) was measured and compared to design. This comparison was used as one of the criteria for establishing the validity of the core physics model. I Initial conditions were control rod group (CRG) 7 at 90% withdrawn, CRG 8 at 37.5% withdrawn, and all others at 100% withdrawn, and a boron concentration of 1296 ppa. To perform the ARO measurement, CRG 7 was fully withdrawn and the associated positive reactivity worth read from the reactimeter (strip chart recorder). This worth, 138 percent milli rho (pcm), was divided by the differential boron worth of -9.64 pcm/ ppm ; from section 4.12 figure 6 of the B&W Physics Test Manual and subtracted from the 1296 ppm value. An ARO critical boron concentration of 1310 ppm was obtained, well within the acceptance criterion of 1351 1 100 ppm. 810816.0gg9 - -
- FOST REFUELING PHYSICS TESTING '
PAGE 2 Temperature Coefficient of Reactivity The temperature coefficient of reactivity is defined as the fractional change in the core reactivity per unit change in the average core temperature. This quantity is the sum of the change in reactivity per unit change in the fuel temperature (Doppler coef ficient) plus the change in reactivity per unit change in the moderator tempera-ture (moderator temperature coefficient). The temperature coefficient was measured directly by observing the reactivity change induced by a change in the average core temperature. From the B&W Zer, Power Test Specification, the isothermal Doppler coefficient is -2.0 pcm. Using this value, the moderator temperature coefficient was indirectly abtained by subtracting the Doppler coefficient from the measured temperature coefficient. I.emperature ramps were initiated for each measurement to obtain the RCS temperature change sequence of approximately -50F.,+100F, and -50F. The temperature coefficient of reactivity, defined ac3<c y e/3 ,4 / d Thee. was figured for each temperature change. The change in reactivity was taken from the strip chart recorder and reactimeter. The second measurement with a +100F temperature change, since it covered a wider temperature range, was used for comparison to the acceptance criteria. The temperature coefficient measurements were made twice during the Zero Power Physics Testing. The first measurements were made at RCS conditions of 531 F and 2155 psig with a boron concentration of 1286 ppm. Control rod assembly group 7 was at 88.5%, group 8 was at 37.5% and all other rods were 100% withdrawn. The moderator coeffi-cient of reactivity was found to be positive for the first run so it was extrapolated to 95% full power where it was determined to be negative. The second set of measurements was made with a boron concentration of 985 ppm with j control rod groups 1-4 at 100% withdrawn, group 5 at 9% withdrawn, groups 6 and 7 at l 0% withdrawn and group 8 again at 37.5% withdrawn. Because of the lower boron con-centration, the moderator coefficient of reactivity was found to be negative. Measurement results for both runs are summarized in Table 3. All moderator coefficients were within the acceptance criteria of less than +0.9 x 10-4 a K/K/0F (9.0 pcm/0F) and the temperature coefficients were within + 40.g J/0F of the predicted value and thus met the acceptance criteria. Control Rod Reactivity Measurements During Zero Power Physics Testing at 531 F, measurements were made to determine the reactivity worths of the regulating rods (CRGs 5, 6, and 7) using the reactimeter. The " boron swap" method was utilized to determine the differential and integral worths. The " boron swap" technique consists of establishing a slow deboration rate while in-serting the control rods periodically in discrete steps to compensate for the reacti-vity change. Reactivity and control rod group position were continuously recorded versus time and analyzed to determine the differential reactivity due to each rod movement. Integral control rod worths were obtained by adding the differential changes in reactivity over the range of control rod travel. Measured worths are tabulated and compared with predicted values in Table 4. All rod worth data met their acceptance criteria.
POST REFUELING PIIYSICS TESTING PAGE 3 l Ejected Control Rod Worth The ejected rod worth test was performed to measure the reactivity worth of the single predicted worst case ejected rod. CRA rod 7/6 was the predicted most reac- , t've ejected rod. The measurement began with the following rod configuration: CR Groups 1-4 at 100% withdrawn CR Group 5 at 9% withdrawn CR Groups 6-7 at 0% withdrawn CR Group 8 was 37.5% withdrawn CRA rod 7/6 was borated out of the core and then reinserted utilizing a rod swap with CR group 5. The worth of CR group 5 inserted during the swap was obtained from the rod worth measurements performed earlier in the testing program. Both values of ejected rod worth were adjusted to the CRG insertion limit and then averaged to obtain the final worth. This resulted in an ejected rod worth of 0.819% thK/K which is with-in 20% of the predicted value of 0.77% 0,K/K. Following the method recommended by B&W, the calculated values were further error adjusted to account for measurement uncertainty. This value was compared to the safety analysis stuck rod worth. Predicted Value Measured (error adjusted) Value % Deviation c
- 0.77% D;K/K 0.819%CiK/K - 6.0%
, 1.0 (safety analysis) 0.911 + 9.8% Core Performance During Power Escalation Sequence Tests Af ter the operating characteristics of the reactor at zero power had buen verified, a progrem of power escalation was undertaken. Testing was conducted at two major power plateaus of 40% and 75% full power. After verifying operability of the incore neutron monitoring system, a core power distribution test was performed to identify possible anamolies in the power distribution and in various thermal hydraulic parameters. An additional test, power imbalance detector correlation, was performed at 40% full power to calibrate the NI power range imbalance gain to the incore detectors. An analysis of the significant parameters at each plateau was made prior to escalating
- power above that plateau. The following sections give more details relevant to the performance of the physics tests conducted during the escalation phase of testing.
l
- Power escalation tests are also to be performed at 100% full power. Ilowever, this plateau was not reached due to unexpected problems with the reactor coolant pump seals. When 100% full power is reached, the power escalation tests will be completed.
The remaining tests to be performed at 100% power are:
- 1) Core Power Distribution
- 2) Temperature Reactivity Coefficient
- 3) Power Doppler Reactivity Coefficient
POST REFUELING PHYSICS TESTING PAGE 4 Power Imbalance Detector Correlation Test, PT 5175.01 The relationship between axial power offset as indicated by the out-of-core power range nuclear instrumentation and the incore detectors was obtained by performance of PT 5175.01. During testing, the minimum DNBR and maximum linear heat rate (LHR) at various oftsets were calculated to verify that their respective limits were not ex-ceeded. This procedure was conducted at the 40% full power plateau. The test was performed by moving the axial power shaping rods until a desired offset was obtained. After waiting 15 minutes at this offset for conditions to stabilize, the following were obtained:
- 1. Core power distribution data including incore offset values and computed maximum linear heat rates and minimum DNBRs
- 2. Out-of-core offset values
- 3. Backup incore detector readings The of fset measured by each out of core detector was plotted against the correspond-ing incore detector offset value, and a least squares fit was made through these points to determine the slope. The initial performance of this test determined an out-of-core versus incore offset slope somewhat less than the 1.15 acceptance criter-ion, although still conservative. New gain settings were calculated and adjustments made on all power range NI channels.
The final computed slopes for the four out-of-core power range instruments were between 1.17 and 1.18, which satisfied the acceptance criterion. The maximum linear heat rate and minimum DNBR for the different of fsets were also within their respective accep-tance ranges. The results of this test can be found in Table 1. Core Power Distribution, PT 5175.02 During power escalation, power distribution data provided by the incore detector system was collected and analyzed at 40% and 75% full power steady state conditions. The measured values of total power peak (peak segment power / average segment power) and radial peak (maximum assembly power / average assembly power) were compared to B&W pre-dictions. The worst case maximum linear heat rate (MLHR), minimum DNBR and quadrant power tilt (QPT) were verified to be within design limits. All these parameters were used to verify the B&W nuclear and thermal-hydraulic calculated models, thereby estab-lishing the acceptability of these codes for safety evaluations. In addition, this procedure verified that values of DNBR and LHR extrapolated to the overpower trip set-point of the next power plateau remained within limits. This allowed escalation to the next power plateau without unknowingly exceeding these safety parameters. The nuclear enthalpy rise hot channel factor (FNH) and nuclear heat flux (Fq) were also shown to be acceptable. The total peaks for both the 40% and 75% full power tests were found to be within their respective acceptance ranges, however, the radial peaks for both power levels exceeded their limits. At 40% full power, the measured radial peak exceeded the predicted peak by 9.8%, outside of the 8.0% acceptance range, and at 75% full power, the measured radial peak exceeded the predicted by 9.1%, outside of its 5.0% acceptance value. The results of this test can be found in Table 2.
FOST REFUELING PIIYSICS TESTING PAGE 5 Further analysis of this problem by B&W using the 3-D FLA}iE computer code revealed that Cycle 2 tech specs would remain valid for a radial peaking difference at least as large as 107. full power. The analysis also took into consideration the lack of significant discrepancies between measured and predicted core behavior during Zero Power Physics Testing, and that the more important measurement of total peak was well within the acceptance range. From the above it was concluded that further power escalation was acceptable and did not compromise plant safety. m
1, POST REFUELING PHYSICS TESTING PAGE 6 TABLE 1 Power Imbalance Detector Correlation Summary of Initial Axial Offset Measurements Offset Initial Values 0C0
% Power ICO NI-5 NI-6 NI-7 NI-S 40.34 0.728 1.366 0.311 0.423 -0.0091 '40.43 6.70 8.078 6.81 7.05 6.39 40.57 10.79 10.94 10.05 10.34 9.6 39.86 -6.16 -6.55 -8.16 -7.69 -8.44 39.47 -11.80 -13.26 -15.16 -14.41 -15.27 39.88 -15.50 -17.23 -18.95 -18.26 -19.22 40.48 -18.63 -21.74 -23.48 -22.80 -23.76 Slope 1.11 1.14 1.13 1.13 Final Values After Gain Adjustments 39.98 -1.91 -2.75 -2.97 -2.63 -3.65 39.09 8.02 8.79 8.52 8.92 7.84
< 39.04 -8.32 -10.58 -10.76 -10.25 -11.42 Slope 1.18 1.18 1.17 1.18 ICO = In-core offset = (Powup - Pow 1w) (100%) (Powup + Pow 1w) OCO = Out-of-core offset = Channel Imbalance x 100% Channel Power
. POST REFUELING PHYSICS TESTING PACE 7 TABLE 2 Core Power Distribution
.40% F.P.
~ ~
75% F.P. Acceptance Criteria Hinimum DNBR > 1. 30 10.71 5.69 Extrapolated Maximum Linear Heat Rate < 20.17 kw/ft 10.67
. 13.9 Extrapolated DNBR > 1.30 5.7 3.8 Quadrant Power Tilt < 3.21 1.059 1.053 Worst Cise Maximum Level Linear Heat Rate 1 16. 2 kw/ f t 3.135 6.38 2 16.7 4.556 8.971 3 17.3 5.103 9.789 4 18.4 5.176 9.789 5 17.6 5.067 9.3176 6 17.2 4.884 8.757 7 16.8 3.827 6.929 RADIAL PEAK TOTAL PEAK predicted measur. accept. deviation predicted measur, accept. deviation 40% 1.226 1.36 8.0% 9.8% 1.489 1.574 12.0% 5.3%
75% 1.218 1.34 5.0% 9.1% 1.505 1.5872 7.5% 4.6%
~ - . - -- . -
POST REFUELING PHYSICS TESTING PAGE 8 PRE-CRITICAL TESTING Control Rod Trip Test Control rod drop times from fully withdrawn to 75% inserted were measured in pre-critical testing at RCS conditions of >525 F and 2150 psig with all RCFs running. 3 This was done in accordance with ST 5013.02, " Control Rod Assembly Insertion Time Test". All rods in groups 1 through 7 were tested by fully withdrawing a group and then dropping the rods into the core using the auxiliary power supply trip switches. Times to 3/4 insertion were read off multi-pen strip chart recorders. The measured times ranged from 1.156 seconds to 1.224 seconds. These values were all well under the 1.58 second acceptance criteria. Reactor Coolant Flow Coastdown The flow coastdown test was performed at HZP conditions prior to criticality with all reactor coolant pumps (RCPs) initially running. It involved tripping the RCP with the highest flow and measuring the rate of total reactor coolant system flow decrease. The reactor coolant flow versus time was then compared to the B&W acceptance curve. Even with a -2.5% conservatism added to the measured flowrate, the value still fell above the minimum acceptance range. Reactor Coolant Flow The reactor coolant flow with four reactor coolant pumps running was measured at hot zero power steady state conditions. The measured flow, + 2.5 percent for uncertain-ties, fell within the acceptance criteria: Measured Measured Acceptance Measured - Acceptance Flow (gpm) +2.5% (gpm) (gpm) -2.5% (gpm) (gpm) - 404,162 414,266 4429,400 394,058 2387,200 L. T "
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POST REFUELING PilYSICS TESTING PAGE 11 TABLE 3 TEMPERATURE COEFFICIENT OF REACTIVITY DURING ZERO POWER PIIYSICS TESTING Temperature Coefficient Moderator Tenperature Rod Configuration (44'/ F) Coefficient (3-'/0F) CRA Groups 1-6 @ 100% wd CRA Group 7 0 88.5% wd CRA Group 8 @ 37.5% wd -3.37 +16.63 RCS Boron = 1286 ppm CRA Groups 1-4 @ 100% wd CRA Group 5 @ 9% wd CRA Groups 6 & 7 @ 0% wa -95.56 -75.56 CRA Group 8 @ 37.5% wd RCS Boron = 985 ppm J e k-_~.-._._-'_-___-_=--.__. ^
s POST REFUEING PHYSICS TESTING PAGE 12 TABLE 4 ROD WORTH DATA DURING ZERO POWER PIIYSICS TESTING Measured Worth Predicted Worth Deviation CRA Group % d K/K % K/K % 7 1.44 1.35 -6.25 6 0.85 0.84 -1.17 5 1.206 1.27 5.3 5-7 3.496 3.41 -2.5 l k m..}}