ML17319A636
| ML17319A636 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 09/22/1980 |
| From: | WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML17319A634 | List: |
| References | |
| NUDOCS 8010010543 | |
| Download: ML17319A636 (38) | |
Text
Attachment 2 to AEP:NRC:00453 Transient Reanalysis Report
'SAFETY ANALYSIS'OR OPERATION OF D. C. COOK UNIT 2 WITE A'OS'ITIVE MODERATOR COEFFICIENT INTRODUCTION This safety analysis has been performed to support the proposed Technical Specification change for D. C. Cook Unit 2 which would allow a small, positive moderator temperature coefficient to exist.
at power levels below 70 percent power. The results of the analy-sis, wnich are presented below, show that the proposed change can be accommodated with ample margin to the applicable FSAR safety limits.
The present D. C. Cook Unit 2 Technical Specifications require the moderator temperature coefficient (MTC) to be zero or negative at all times while the reactor is critical. This requirement is overly restrictive, since a small positive coefficient at reduced power levels could result in a significant increase in fuel cycle flexibility, but would have only a minor effect on the safety analy-sis of the accident events presented in the FSAR.
The proposed Technical Specifications change, given in Attachment 1 to the letter, allows a +5 pcm/'F* MTC below 70 percent of rated power, changing to a 0 pcm/'F MTC at 70 percent power and above.
This MTC is diagrammed in Figure l. A power-level dependent MTC
- 1 pcm = 10 hk/k
was chosen to minimize the effect of the specification change on postulated accidents at high power levels. Moreover, as the power level is raised, the average coolant temperature becomes higher as allowed by the programmed average temperature controller for the plant, tending to bring the moderator coefficient more negative.
Also, the boron concentration can be reduced as xenon builds into the core. Thus, there is less need to allow a positive coefficient I
as full power is approached. As fuel burnup is achieved, boron is further reduced and the moderator coefficient will eventually become negative over the entire operating power range.
ACCIDENT ANALYSIS The impact of a positive moderator temperature coefficient on the accident analyses presented in Chapter 14 of the D. C. Cook Unit 2 FSAR(1) has been assessed. Those incidents which were found to be sensitive to positive or near-zero moderator coefficients were reanalyzed. In general, these incidents are limited to transients which cause the reactor coolant temperature to increase. With one exception, the analyses presented herein were based on a +5 pcm/'F moderator temperature coefficient, which was assumed to remain
- constant for variations in temperature.
The analysis in which this was not the case,was the control rod ejection analysis which was based on a coefficient equal to
+5 pcm/ F at zero power nominal average temperature and which became less positive for higher temperatures. This was necessary since the TWINKLE computer code, on which the analysis is based, is a diffusion-theory code rather than a point-kinetics approximation and the moderator temperature feedback cannot be artificially held constant with temperature. For all accidents which were reanalyzed, the assumption of a positive moderator temperature coefficient existing at full power is conservative since the proposed Technical Specification requires that the coefficient be zero or negative at or above 70 percent power.
Zn general, the reanalysis was based on the identical analysis methods, computer codes, and assumptions employed in the FSAR; any exceptions are noted in the discussion of each, incident. Accidents not re-analyzed included those resulting in excessive heat removal from the reactor coolant system for which a large negative moderator coefficient is more limiting, and those for which heatup effects following reactor trip are not sensitive to the moderator coefficient. Table I gives a list of accidents presented in the D. C. Cook Unit 2 FSAR and denotes those events reanalyzed for a positive coefficient.
Z. Transients Not Affected B a Positive Moderator Coefficient The following transients were not reanalyzed since they either result in a reduction in reactor coolant system temperature and are therefore
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sensitive to a negative moderator temperature coefficient, or are otherwise negligibly affected by a positive moderator temperature coefficient.
A. RCCA Misa'li nment/Dro The RCCA drop case presented in Section 14.1.3 of the FSAR is potentially affected by a positive moderator temperature coefficient. Use of a positive coefficient in the analysis would result in a larger reduction in core power level follow-ing the RCCA drop, thereby increasing the probability of a reactor trip. For the return to power with automatic rod control case, a positive coefficient (which would only exist below 70 percent power) would result in a small increase in the power overshoot. Westinghouse has performed extensive analyses in this area .which has demonstrated that the limiting conditions for this transient are at or near 100 percent power where the moderator temperature coefficient must be zero or negative.
On this basis, the analysis for this event was not repeated.
B. Startu of an Inactive Reactor Coolant Loo An inadvertent startup of an idle reactor coolant pump results in a decrease in core average temperature. As the most nega-tive values of moderator reactivity coefficient produce the greatest reactivity addition, the analysis reported in the FSAR, Section 14.1.7, represents'he limiting case.
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C. 'xc'e's'sive'e'atR'emova'1 Du'et'o'. Fee'dwater 'S 'stem Ma'1'funct'ions The addition of excessive feedwater or the reduction of feed-water temperature are excessive heat removal incidents, and are consequently most sensitive to a negative moderator temperature coefficient. Results presented in Section 14.1.10 of the FSAR Jl based on a negative coefficient, represent the limiting case.
D. 'xcessive Lo'ad increase An excessive load increase event, in which the steam load exceeds the core power, results in a decrease in reactor cool-ant system temperature. With the reactor in manual control, the analysis presented in Section 14.1.11 of the FSAR shows that, the limiting case is with a large negative moderator coefficient. Xf the reactor is in automatic control, the control rods are withdrawn to increase power and restore the average temperature to the programmed value. The analysis of this case in the FSAR show@that the minimum DNBR is not sensitive to moderator temperature coefficient. Therefore, the results presented in the FSAR are still applicable to this incident.
E. Los's'f Normal Feedwater, Loss of Offsite Power The loss of normal feedwater and loss of offsite power acci-dents (Sections 14.1.9 and 14.1.12 of the FSAR) are analyzed to determine the ability of the secondary system to remove decay heat. These events are not sensitive to a positive
moderator coefficient since the reactor trip occurs at the beginning of the transient before the reactor, coolant system temperature increases significantly. Therefore, these events were not reanalyzed.
F. Ru ture of a Main Steam Pi e Since the rupture of a main steam pipe is a temperature reduc-tion transient, minimum core shutdown margin is associated with a strong negative moderator=temperature coefficient. The worst conditions for a steamline break are therefore those analyzed in the FSAR (Section 14.2.5).
G. L'oss of Coolant Accident (LOCA)
The loss of coolant accident (Section 14.3 of the FSAR) is analyzed to determine the core heatup consequences caused by a rupture of the reactor coolant system boundary. The event results in a depressurization of the RCS and a reactor shutdown at the beginning of the transient. This accident was not reanalyzed since the Technical Specification requirement that the temperature coefficient be zero or negative at 70 percent power or above ensures that the previous analysis basis for this event is not affected.
XZ. Transients Sensitive to a Po'sitive'od'erator'o'e'ffi'c'i'ent A. pron'ilut'i"on As stated in Section 14.1.5 of the FSAR, an uncontrolled boron dilution incident cannot occur during refueling due 'to admin-istrative controls which isolate the reactor coolant system
~ ~ I from the potential source of unborated water. If a boron dilution incident occurs during startup, the FSAR shows that the operator has sufficient time to identify the probl'em and terminate the dilution before the reactor returns to criti-cality. Therefore, the value of the moderator coefficient has no effect on a boron dilution incident during startup.
The reactivity addition due to a boron dilution at power'ill cause an increase in power and reactor coolant system temperature. Due to the temperature increase, a positive moderator coefficient would add additional reactivity and increase the severity of the transient. With the reactor in automatic control, however, the rod insertion alarms provide the operator with adequate time to terminate the dilution before shutdown margin is lost. A boron dilution incident with the reactor in manual control is no more severe than a rod withdrawal at power, which is analyzed below and therefore this case was not specifically analyzed. Following reactor trip, the amount of time available be fore shutdown margin is lost is not affected by the moderator coefficient.,
B.'ontrol Rod Withdrawal From a Subcritical Con'di:t'i.'on Intro'duction A control rod assembly withdrawal incident when the reactor is subcritical results in an uncontrolled addition of reactivity leading to a power excursion (Section 14.1.1 of the FSAR). The
nuclear power response is characterized by a very fast rise terminated by the reactivity feedback of the negative fuel temperature (ie. Doppler) coefficient. The power excursion causes a heatup of the moderator and fuel. The reactivity addition'ue to a positive moderator coefficient results in increases in peak heat flux and peak fuel and clad temperatures.
Method of Anal sis The analysis was performed in the FSAR for a reactivity inser-tion rate of 75 pcm/sec. However, this value was found to be overly conservative based on the actual cycle values for D. C. Cook Unit 2 (2). A reactivity insertion rate of 60 pcm/sec which is still a very conservative value was used in the analysis with a po'sitive moderator coefficient. This assumed reactivity insertion rate is greater than that for the simultaneous withdrawal of the combination of the two sequen-tial control banks having the greatest. combined worth at maximum speed (45 inches/minute) . A constant. moderator temperature coefficient of +5 pcm/'F was used in the analysis.
The digital computer codes, initial power level, and reactor trip instrument delays and setpoint errors used in the analysis were the same as used in the FSAR.
Results and Conclusions The nuclear power, coolant temperature, heat flux, fuel average temperature, and clad temperature versus time for a 60 pcm/sec
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I insertion rate are shown in Figures 2 through 4. This insertion rate, coupled with a positive moderator temperature coefficient of +5 pcm/ F, yields a peak heat flux which does not exceed that presented in the FSAR. Therefore the conclusions presented in the FSAR are still applicable.
C. Uncontrolled Control Rod Assembl Withdrawa'1 'at.'o'wer Introduction An uncontrolled control rod assembly withdrawal at power produces a mismatch in steam flow and core power, resulting in an increase in reactor coolant temperature. A positive moderator coefficient would augment the power mismatch and could reduce the margin to DNB. A discussion of this incident is presented in Section 14.1;2 of the FSAR.
Method of Anal sis The transient was reanalyzed employing the same digital computer code and assumptions regarding instrumentation and setpoint errors used for the FSAR. This transient was only analyzed at 100 percent power with a positive moderator coefficient since this case is the most limiting of those presented in the FSAR. A constant moderator coefficient of +5 pcm/'F was used in the analysis. The assumption that a positive moderator coefficient exists at full power is conservative since at full power the moderator coefficient will actually be'egative.
For thi's case, the DNB evaluation was performed using the improved thermal design procedure (3) .
Results Figure 5 shows the minimum DNBR as a. function of reactivity insertion rate. The limiting case for DNB margin is a re-activity insertion rate of 0.6 pcm/sec from full power initial conditions which results in a minimum DNBR of 1.98. A positive moderator coefficient therefore does not lower the DNBR associated with a control rod assembly withdrawal at power below the limit value of 1.80.
Conclusions These results demonstrate that the conclusions presented in the FSAR are still valid. That is, the core and reactor coolant system are not adversely affected since nuclear flux and over-temperature hT trips prevent the core minimum DNB ratio from falling below 1.80 for this incident.
Loss of Reactor Coolant Flow Introduction As demonstrated in the FSAR, Section 14.1.6, the most severe loss of flow transient is caused by the simultaneous loss o' electrical power to all four reactor coolant pumps. This tran-sient was reanalyzed to determine the effect of a positive moderator temperature coefficient on the nuclear power tran-sient and the resultant effect on the minimum DNBR reached during the incident. The effect on the nuclear power transient would be limited to the initial stages of the incident during
1 ~ ii I IQ ~
which reactor coolant temperature increases since this increase is terminated shortly after reactor trip.
Method'f Anal sis Analysis methods and assumptions used in the re-evaluation were consistent with those employed in the FSAR.
The digital computer codes used to calculate the flow coast-down and resulting system transient were the same as those used to perform the FSAR analysis. The analysis was done with a constant moderator coefficient of +5 pcm/~F and the DNB evaluation was performed using the improved thermal design procedure (3).
Results For the analysis performed with a +5 pcm/ F moderator coefficient, the reactor coolant average temperature increases less than 2'F above the initial value. Therefore, a positive moderator coefficient does not appreciably affect. the reactor coolant system response or the minimum DNBR reached during the transient. For this case, a minimum DNBR of 2.08 was obtained. Figures 6 through 8 show the flow coastdown, the nuclear power and heat flux transients, and the minimum DNBR versus time.
Conclusions A positive moderator temperature'coefficient does not appreciably affect the result of the complete loss of flow transient, and the minimum DNBR remains above the limit value of 1.80 for this incident. This case was analyzed since it is the most limiting one presented in the FSAR.
Since the transient causes only a small change in core average moderator temperature, and the positive moderator coefficient does not appreciably affect. the nuclear power transient, the single pump loss of flow cases will also not be appreciably affected.
Locked Rotor Xntroduction The FSAR (Section 14.1.6), shows that the most severe locked rotor incident is an instantaneous seizure of a reactor coolant pump rotor at 100 percent power with four loops operating.
Following the incident, reactor coolant system temperature rises until shortly after reactor trip. A positive moderator coefficient. will not affect the time to DNB since DNB is conservatively assumed to occur at the beginning of the incident. The transient was reanlayzed, however, due to the potential effect on the nuclear power transient and thus on the peak reactor coolant system pressure and fuel temperatures.
Meth'od of Anal sis The digital computer codes used in the reanalysis to evaluate
the pressure transient and thermal transient were the same as those used in the FSAR. The assumptions used were also con-sistent with those employed in the FSAR. An analysis was done at 70 percent power with a moderator coefficient of'+5 pcm/'F in order to show that this case is not more limiting than the 100 percent power, 0 pcm/ F case presented in the FSAR. This case is sufficient to illustrate the impact on the transient by a positive moderator coe fficient, since the moderator coefficient will actually be zero or negative-at full power.
Results and Conclusions Table II compares results obtained for this case with those presented in the -FSAR. As shown in the table, the FSAR analy-sis at full power with a zero moderator coefficient is more limiting than the 70 percent power case with a positive moderator coefficient. Therefore, the conclusions presented in the FSAR are still applicable.
L'oss of External Electrical Load Introduction Two cases, analyzed for both beginning and end of life conditions, are presented in Section 14.1.8 of the FSAR-
- l. Reactor in automatic rod control with operation of the pressurizer spray and the pressurizer power operated relief valves; and
- 2. Reactor in manual rod control with no credit for pressurizer spray or power'operated relief valves.
As the moderator temperature coefficient will be negative at end of life, only beginning of life cases were repeated. The result of a loss of load is a core power level which momentarily exceeds the secondary system power removal causing an increase in core water temperature. The consequences of the reactivity addition due to a positive moderator coefficient. are increases in both peak nuclear power and pressurizer pressure'.
Method of Anal sis A constant moderator temperature coefficient of +5 pcm/~F was assumed. The method of analysis and assumptions used were otherwise in accordance with those presented in the FSAR. The improved thermal procedure (3) was utilized in the DNB evaluation.
Results The system transient response to a total loss of load from 102 percent power, with control rods in automatic control, assuming pressurizer relief and spray valves, is shown in Figures 9 and
- 10. Peak RCS pressure reaches 2567 psia following a reactor trip on overtemperature hT. This compares to a value of 2493 psia presented in the FSAR. A minimum DNBR of 2.30 is reached shortly after reactor trip.
ll and P
Figures 12 illustrate reactor coolant system response to a loss of load with rods in manual control, assuming no credit for pressure control. Peak RCS pressure reaches'604 psia following reactor trip on high pressurizer pressure. The peak pressure reached in the FSAR analysis for this case was 2597 psia. The minimum DNBR is initially 2.71 and increases throughout the transient.
Conclusions The analysis demonstrates that the integrity of the core and the reactor coolant system pressure boundary during a loss of load transient will not be affected by a positive moderator reactivity coefficient since the minimum DNB ratio remains well above the 1.80 limit, and the peak reactor coolant pressure is less than 110 percent of design. Therefore, the conclusions presented in the FSAR are still applicable.
Rupture of a Control Rod. Drive Mechanism Housin /Control Rod E'ection introduction The rod ejection transient is analyzed at full power and hot standby for both beginning and end of life conditions. Since the moderator temperature coefficient is negative at end of life, only the beginning of life cases were reanalyzed. The high nuclear power levels and hot spot fuel temperatures resulting from a rod ejection are increased by a positive
moderator coefficient. A discussion of this transient is presented in Section 14.2.6 of the FSAR.
Metho'd of Anal sis The digital computer codes for analyses of the nuclear and hot spot heat transfer are the same as those power'ransient used in the FSAR. The ejected rod worths and transient peaking factors were the same as reported in Reference 3. The moderator coefficient used for this transient was +5 pcm/'F at zero power nominal average temperature, decreasing to approximately +4 pcm/'F at full power T-a'verage. Th'is is still a conservative assumption since the moderator coefficient actually is zero or negative above 70 percent. power.
Results and Conclusions Peak fuel and clad temperatures and nuclear power versus time for both full power and hot standby are presented in Figures 13 through 16. A comparison of reanalysis and FSAR results is presented in Table III. The limiting peak hot spot clad temperature, 2469 F, was reached in the hot full power case.
Maximum fuel temperatures were also associated with the full power case. Although the peak hot spot fuel centerline temperature for this transient exceeded- the melting point, melting was restricted to less than the innermost 10 percent.
of the pellet.
As fuel and clad temperature do not exceed the fuel and clad
limits specified in the FSAR, there is no danger of sudden fuel dispersal into the coolant, or consequential damage to the primary coolant loop. Therefore, the effects of a positive MTC of the magnitude described above is acceptable.
SUMMARY
In order to assess the effect on accident analysis of operation of D. C. Cook Unit 2 with a slightly positive moderator temperature coefficient, a safety analysis of transients sensitive to a positive moderator coefficient was performed. These transients included control rod assembly withdrawal from subcritical, control rod assembly withdrawal at power, loss of reactor coolant flow, loss of external load, and control rod ejection. This study indicated that a small positive moderator coefficient does not result in the violation of any applicable safety limits for the transients analyzed.
Except as noted, the analyses employed a constant moderator coefficient of +5 pcm/ F, independent of power level. The results of this study are conservative for the accidents investigated at full power, since the proposed Technical Specification requires that the coefficient be zero or negative at or above 70 percent power.
REFERENCES
- l. Donald C. Cook Nuclear Plant, Unit 2, Final Safety Analysis Report, Amendment 75, dated April 1977 and supplements, Docket No. 50-316.
- 2. Reload Safety Evaluation Report, D. C. Cook Nuclear Plant, Unit 2, Cycle 2, July 1979, (letter No. AWF-311, 8/17/79).
- 3. Chelemer, H., Boman, L. H. and Sharp, D. R., "Improved Thermal Design Procedure," WCAP-8567, July 1975 (Proprietary) and WCAP-8568 (Non-Proprietary).
TABLE I ACCIDENTS EVALUATED FOR
POS'ITIVE'ODERATOR COEFFICIENT EFFECTS
'SAR Accident Time i:n'i:fe
- 14.1.1 RCCA Withdrawal from Subcritical BOC
- 14.1.8 Loss of Load/Turbine Trip BOC 1'4. 1. 9 Loss of Feedwater 14.1.10 Feedwater,Malfunction EOC 14.1.11 Excessive Load Increase BOC/EOC 14.1.12 Station Blackout
- 14. 2. 5 Steam Line Break EOC
TABLE II COMPARISON QF'ESULTS'QR'OCKED'OTOR 'ANALYSES This Stu'd FSAR Moderator temperature coefficient, hk/k/'F 5 x 10-5 Initial power level, percent of nominal 70 100 Peak fuel pellet av. temperature, 'F 2253 2906 Peak clad temperature during transient, 'F 1586 1878 Peak reactor coolant system pressure, psia 2521 2633
TABLE III SU1SGWY OF'OD EJECTION RESULTS'EGINNING 'OF CYCLE (This Study)
Hot Hot Zero Full Power Power Maximum fuel pellet average temperature, 'F 2935 4088 Maximum fuel center temperature, 'F 3440 4985 Maximum clad average temperature, 'F 2221 2469 Maximum fuel enthalpy, cal/gm 121 179 Fuel pellet melting, percent. <10
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CC LL P LLl LC Ch LaJ CD CD K cJ 0 10 . 20 30 40 50 60 70 80 90 100 I
POMER FIGURE 1 MOOERATOR TEMPERATURE COEFFICIENT VS POWER LEVEL
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10 3 10-'.0 2.5 5.0 7.5 10.0 12.5 15.0 17.5
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FIGURE 2 ROD WITHDRAWAL FROM SUBCRITICAL NUCLEAR POWER VS TIME .
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CORE MATER 1000 CLAD FUEL 750 500 0.0 2.5 5.0 7.5 10.0 12.5 -15.0 '7.5 20';0 TINE (SECONDS)
FIGURE 3 ROD WITHDRAWAL FROM SVBCRITICAL TEMPERATURE VS TINE
CD o 0.S c5 0.4 I
0.2 0.0 0.0 2.5 ~ 5.0 7.5 10.0 12.5 15.0 17.5 20.0 TIME (SECQNOS)
FIGURE 4 ROO WITHORAWAL FROM SUBCRITICAL HEAT FLUX VS TIME
2.4 2.3 2 2--
HIGH NEUTRON FLUX TRIP 2.0--
OVERTBlPERATURE
. 5T. TRIP 9.4 0.8 1,0 2.0 4 0 6.0 8.0 10.0 20.0 40.. 60. 80. 100.
REACTIVITY INSERTION (10 6 AK/SEC) ftGURE 5 ROD MITHDRAMAL AT POWDER
j.0 O.S 0.6 0.4 TINE {SECONOS)
FIGURE 6 LOSS OF FLOM FLOW VS TINE
1.2 1.0 HOT CHANNEL
&CD 0.8 0.6 4J4 CD
+as 0.4 eQ 0.2 0.0
'1.2 1.'0 AVERAGE CHANNEL
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eCQ 0.4 cled 0.2 Q.O 0 10 1.2 1.0 4L CD 0.8 O U CD 0.6 CJ +r 0.4 CC 0.2
'.0 0.0 5.0 10.0 15.0 20.0 TINE {SECONDS) .
FIGURE 7 LOSS OF FLOM r!~0</sfs <p'~5';5
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TINE (SECONDS)
FIGURE 8 LOSS OF FLOW DNBR VS TIME
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CD 0.6 mu 04
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'0 Zo 30 TIME (SECONOS) 50 FIGURE 9 LOSS OF LOAO AUTOMATIC ROO CONTROL WITH PRESSURIZER RELIEF ANO SPRAY C ~
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FIGURE 10 LOSS OF LOAD AUTOMATIC ROD CONTROL WITH PRESSURIZER RELIEF AND SPRAY
1.0 LQ Cl 0 8 CQ CD 0.6
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30 50 TIME {SECONOS)
'IGURE 11 LOSS OF LOAO MANUAL ROO CONTROL NO PRESSURIZER RELIEF OR SPRAY
560
~~ 540
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~~ 1000 750 V7 C) 500 250 0
590 580 570 560 10 20 . 30 50 TIME (SECONOS)
FIGURf 12 LOSS Of LOAO MANUAL ROD CONTROL NO PRESSURIZER RELIEF OR SPRAY
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TINE (SECONOS)
FIGURE 13 ROO EJECTION BOL HFP NUCLEAR POlrlER VS TINE
5000 4900'F MELTING
$ 000 FUEL CENTER TEMPERATURE 4000 FUEL AYERAGE TEMPERATURE 3000 2000 CLAD OUTER TEMPERATURE 1000 0
3 TIME (SECONDS)
FIGURE 14 ROD EJECTION BOL HFP TEMPERATURE VS TIME
10 1.0 1.5 2.0 2.5 3.0 TIME (SECONOS)
FIGURE 15 ROO EJECTION SOL HZP NUCLEAR POMER VS TIME
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MOO' 4000 FUEL CENTER TEMPERATURE I
3000 FUEL AVERAGE TBIP ERATURE CLAO OUTER
~i4IP ERATURE ZOOO j000- 1
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'IGURE 16 ROO EJECTION BOL HZP TBlPERATURE VS TINE