ML20085L049
| ML20085L049 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 10/13/1983 |
| From: | Clayton F ALABAMA POWER CO. |
| To: | Varga S Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20085L052 | List: |
| References | |
| NUDOCS 8310210259 | |
| Download: ML20085L049 (50) | |
Text
Malhn Address :
. Atbama Power Company b J 600 North 18th Street l Post Offica Box 2641 Birmingham, Alabama 35291 Telephone 205 783-6081 F. L Clayton, Jr.
l %%;'LT,*"' AlabamaPbwer the southem electrc system l
October 13, 1983 Docket Nos. 50-348 .
50-364 Director, Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D.C. 20555 Attention: Mr. S. A. Varga Joseph M. Farley Nuclear Plant - Units 1 and 2 Positive MTC and F Delta H Technical Specification Changer Gentlemen:
Alabama Power Company has completed an evaluation of operating the Joseph M. Farley Nuclear Plant with a slightly positive moderator temperature coefficient (MTC) at low power levels and an increased enthalpy hot channel factor (F Delta H) limit below full power. The evaluation concluded that operation of the plant with both of the proposed changes does not exceed any safety-related design criteria and does result in significant operational benefits. Thus, Alabama Power Company requests the proposed Technical Specification changes be incorporated for both Units 1 and 2.
The current MTC Technical Specification requires a non-positive value at all power levels. The basis for this is to ensure that the value of the coefficient remains within the limiting conditions assumed in the FSAR accident and transient analyses. The proposed change allows a slightly positive MTC (< 0.5X10 4 Delta K/K/ F) below 70% Rated Thermal Power and a non-positive value at or above 70% Rated Thermal Power. In keeping with the bases, Westinghouse performed the necessary accident and transient analyses based on the new MTC values to ensure that the results remain within all limiting conditions and safety criteria. The Westinghouse safety analysis ( Attachment 1) provides the basis for the change and shows that the limiting conditions and safety criteria are met.
h ) b 8310210259 831013 f{$ V&
, n Mr. S. A. Varga October 13, 1983 U. S. Nuclear Regulatory Commission Page 2 The current F Delta H Technical Specifications limit increases linearly with a slope of 0.2 at reduced power levels. The proposed change allows a slope of 0.3 rather than 0.2, which increases the sometimes-too-restrictive allowable limit at reduced power where F Delta H is of less concern and core exit boiling is the primary concern. The liniit at full power remains the same. The Westinghouse safety analysis (Attachment 1) demonstrates that the current Farley Nuclear Plant safety analyses are not impacted by this proposed change. In addition, as a result of this evaluation, new core safety limits have been established. The overpower and overtemperature delta-T setpoints were reviewed in light of the new core safety limits and revised in order to optimize the available operational margin that exists between the setpoints and the core safety limits.
The primary benefit associated with the MTC change is the reduced burnable poison rods required to control peaking during the early portion of each cycle. A resultant benefit of the reduced burnable poison rods is a slightly longer fuel cycle. This is particularly important since Farley is implementing 18-month fuel cycles beginning with Cycle 3 on Unit 2 and Cycle 6 on Unit 1. Additionally, the smaller number of burnable poison rods will reduce handling requirements and resultant problems associated with the storage and disposal of spent burnable poison rods. The F Delta H change will reduce the probability of administrative rod control limits being imposed for low power operation, and thus improves operational flexibility.
The proposed changes to the Joseph M. Farley Technical Specifications are presented in Attachment 2. The changes include redefining the MTC limits in Specification 3.1.1.3, changing the 0.2 l multiplier to 0.3 in Specifications B.2.1.1 and 3.2.3, redefining the F( Delta I) band and the trip setpoints in Table 2.2-1, and replacing Figure 2.1-1 with new Figure 2.1-1.
Alabama Power Company requests approval of the proposed Technical Specification changes by January 10, 1984 to support the reload design and safety evaluation for Unit 1, Cycle 6. This evaluation must be completed prior to the Unit 1 shutdown for refueling which could occur as early as January 13, 1984. The requested changes are not cycle j
dependent; however, future cycle designs are dependent upon the approval of these changes. If approval is not granted by January 10, 1984, the necessary action to redesign the core and secure the required burnable
- poison rods may become a critical path item that could delay the scheduled restart of Unit 1.
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c Mr. S. A. Varga October 13, 1983 U. S. Nuclear Regulatory Commission Page 3 Alabama Power Ccmpany has reviewed these proposed changes and determined that they do not result in a significant increase to the probability or consequence of a previously-analyzed accident or significantly reduce in some way a safety margin. Additionally, Alabama Power Company has determined that the results of the changes are clearly within all acceptable design criteria, as evidenced in Attachment 1.
Therefore, the proposed changes are consistent with Item (vi) of the
" Examples of Amendments that are Considered Not Likely to Involve Significant Hazards Considerations" listed on page 14870 of the April 6, 1983 issue of the Federal Register, and a significant hazards consideration, as defined in 10CFR50.92, is not involved. The proposed change does result in some slight decrease in a safety margin, but the results of the change are clearly within all acceptable criteria with respect to the safety analysis specified in the Standard Review Plan.
The Plant Operations Review Committee has reviewed this proposed change to the Technical Specifications. The Nuclear Operations Review Board will review this change at a future meeting.
The class of this change is designated as Class III for Unit 1 and Class I for Unit 2 in accordance with 10CFR170.22 requirements. A check for $4,400 to cover the total amount of fees required is enclosed with this letter.
In accordance with 10CFR50.30(c)(1)(i), three signed originals and forty (40) additional copies of this proposed change are enclosed. As noted by the distribution, a copy of this letter is being sent to the Alabama State Designee in accordance with 10CFR50.91 (b)(1).
Yours very tr i l
( .
gQL. Clayton, Jr.
FLCJ r/JLO:ddr-D4 Attachments cc: Mr. R. A. Thomas SWORN TO A(D SUBSCRIBE BEFORE ME Mr. G. F. Trowbridge THIS (3 DAYOF([L440983 Mr. J. P. O'Reilly Mr. E. A. Reeves Mr. W. H. Bradford A htvDLbla_s Dr. I. L. Myers Notary Public My Commission Expires: Ho ,37
_ .-. _. . ._~ . - . . - . _ _ = _ . - . . .. _ . _ _ _ _
bc: Mr. W. O. Whitt
, 3 Mr. R. P. Mcdonald Mr. H. O. Thrash Mr. O. D. Kingsley, Jr.
Mr. W. G. Hairston, III Mr. J. W. McGowan Mr. C. D. Nesbitt Mr. R. G. Berryhill Mr. D. E. Mansfield Mr. J. A. Ripple j Mr. W. G. Ware --
Mr. L. B. Long Mr. J. R. Crane Mr. K. C. Gandhi Reference Listing ,
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WITH POSITIVE MODERATOR TEMPERATURE COEFFICIENT AND REVISED F LIMIT BELOW FULL POWER AH
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. . I INTRODUCTION 1
This safety analysis has been performed to support the proposed Technical l 1
Specification changes for J. M. Farley Units 1 and 2 which would allow a small, positive moderator temperature coefficient to exist at power levels below 70 percent power and a 0.3 multiplier for determining F3g limits at reduced power levels. As a result of the F 3g analysis, new core safety I
limits were established. The overpower and overtemperature delta-T setpoints were reviewed in light of the new core safety limits and were revised in order j to optimize the margin that exists between the setpoints and the core safety l limits. The results of the analysis, which are presented below, show that the .
proposed changes can be accommodated with ample margin to the applicable FSAR i
. safety. limits. .
The present J. M. Farley Technical Specifications require the moderator tem-perature coefficient (MTC) to be zero or negative at all times while the reac-tor is critical. This, requirement is overly restrictive, since a small.posi-tive coefficient at reduced power levels would have only a minor effect on the i safety analysis of the accident events presented in the FSAR. However, a l I
significant reduction in fuel cycle cost results because of a decrease in the l burnable poison inventory required each cycle. .This not only decreases costs j I
related to the purchase and handling of burnable poisons but also results in i an increased cycle length due to a decreased burnable poison penalty. This is important particularly for long (18 month) cycles which require a large number of burnable poisons to control MTC at beginning of cycle.
The proposed MTC Technical Specifications change allows a +5 pcm/*F* MTC below 70 percent of rated power, changing to a O pcm/*F MTC at 70 percent power and above. A power-level dependent MTC was chosen to minimte the effect of the specification change on postulated accidents at higt pegr levels. Moreover, as the power level is raised, the average coolz.i: . m ature becomes higher
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r a i w 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 l less need to allow a positive coefficient 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. 1 The present J. M. Farley Technical Specifications allow an increase in the F
3g limit at reduced powers according to the following equation:
F g = 1.55 (1 v 0.2(1 - P))(1-RBP (BU))
This limit sometimes proves to be restrictive at low power levels, with the result being that either core loading pattern modifications are required, more j burnable poison rods than would otherwise be necessary are needed, or l administrative control rod insertion limits are imposed. To change the i t.urrent 0.2 multiplier to a 0.3 multiplier would virtually eliminate tne need to use any of the corrective actions identified above. Note that the limit on F
H at full power remains the same. Significant increases to the F H limit occur at low powers where F aH is f less concern. At low powers, core exit boiling is of prime importance to plant safety and it is not a function of F,g.
The overpower and overtemperature delta-T trip functions are designed to ensure that the core thermal limits are not exceeded. The core thermal limits are provided for illustrative purposes in the J. M. Farley Technical j Specifications. The overpower and overtemperature delta-T trips are functions f of delta-T, Tavg, and pressure. The current constants for the trip functions i t
as given in the Technical Specifications are: ;
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0vertemperature Delta-T: Ky = 1.14, K2 = .01733, K3 = .000805 l Overpower Delta-T: K4 = 1.07, K5 = .02 for increasing Tavg and l I
0 for decreasing Tavg, K6 = .00116. I i
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r 'D I-These trip functions may ,sometimes cause reactor trips during large load [
changes such as a sudden load rejection. During a load rejection, Tavg l' l-increases and the margin to trip reduces. In some cases, a trip may occur, j resulting in an undesired shutdown or loss of plant availability. [
l A modification to the setpoints is proposed as follows: f I
I Overtemperature Delta-T: Ky = 1.22, K2 = .0154, K3 = .000635 ;
1 Overpower Delta-T: K4 = 1.08, K5 = unchanged, K6 = .00109 ,
1 Of particular note are the increases in Ky and K4 . These constants have l tha most impact in improving plant availability. I I
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ACCIDENT ANALYSIS l Positive Moderator Temperature Coefficient The impact of a positive moderator temperature coefficient on the accident analyses presented in Chapter 15 of the J. M. Farley Units 1 and 2 FSAR have been assessed. Those incidents which were found to be sensitive to positive j or near-zero moderator coefficients were reanalyzed. In general, these inci- j dents are limited to transients which cause the reactor coolant temperature to i ncrease. With one exception, the analyses presented herein are based on a +5 l
pcm/*F moderator temperature coefficient, which is assumed to remain constant for variations in temperature.
The analysis in which this is not the case is the control rod ejection analy-sis which is based on a coefficient equal to +5 pcm/*F at zero power nominal average temperature and which becomes less positive for higher temperatures.
This is 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 l with temperature. For all accidents which are reanalyzed at full power, the assumption of a positve 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.
In general, the reanalysis is based on the identical analysis methods, compu-ter codes, and assumptions employed in the FSAR; any exceptions are noted in the discussion of each incident. Accidents not reanalyzed include those resulting in excessive heat removal from the reactor coolant system for which a large negative moderator temperature coefficient is more limiting, and those for which heatup effects following reactor trip are not sensitive to the [
i moderator coefficient. Table I gives a list of accidents presented in the I
J. M. Farley FSAR and denotes those events reanalyzed for a positive coefficient.
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FAH Multiplier Change , [
l A proposed change from K=0.2 to K-0.3 in the following equation for the l Enthalpy Hot Channel Factor (FN H
) has been evaluated with regard to its effect on accident analyses: l F g = 1.55 (1.0 + K (1-P))(1-RBP(BU))
I Where P = the fraction of full power and the 0.3 multiplier referred to I l
above is the power correction constant, K. l The effect of the proposed change to the F aH Technical Specification on the accident analysis is realized through changes to the core safety limits. f Since the overtemperature and overpower delta-T trip setpoints are based I directly on the core safety limits, the setpoints must-be verified as-being applicable to the new core safety limits.
The proposed F g change only affects the core limits at very high pressure and low power levels. Since the steam generator safety valves prevent the I plant from reaching these limiting conditions, the constants of the overpower l I
and overtemperature delta-T protection setpoints are unaffected by this l change. Changing the F 6g multiplier sometimes impacts the axial offset l envelope in a manner that requires a change to the f(al) term. However, no l credit for the f(al) protection is assumed in the accident analyses.
Therefore, the safety analyses are not impacted by the proposed F H multiplier change. Note that the core limits, which are illustrated in the proposed Technical Specifications, have been changed.
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Verification of the applicability of the current overpower and overtemperature 1
! delta-T setpoints to the new core safety limits showed that there was excess conservatism between the sepoints and the core safety limits. As a result I these setpoints were modified to provide improved operating capability while maintaining sufficient safety margin to the core safety limits. Thus, the proposed setpoints reflect both the 0.3 F g multiplier (which has no impact on the safety analysis) and the removal of excess conservatism between the l
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setpoints and the core safety limits. There are no FSAR transients which i assume the overpower delta-T trip, therefore the changes to this setpoint equation will not impact the safety analysis. The overpower delta-T function is modified only to be consistent with the new overtemperature delta-T setpoints. Those incidents which rely upon the overtemperature delta-T trip were reanalyze'd. These are listed in Table I. Note that these incidents are also impacted by the positive moderator temperature coefficient. Thus, the reanalyses required for the modification of the setpoints were easily I incorporated into the positive moderator temperature coefficient safety f I
analysis.
The.following evaluations address the impacts of the positive moderator f temperature coefficient and the overtemperature delta-T setpoint change. The F3g multiplier and the, overpower delta-T setpoints are not addressed further l
since they have no impact on the safety analysis.
I. Transients Not Affected By a Positive Moderator Coefficient or Setpoint Change The following transients are not reanalyzed since they either result in a reduction in reactor coolant systeni temperature and are therefore sensitive to f I
a negative moderator temperature coefficient, or are otherwise negligibly l affected by a positive moderator temperature coefficient, in addition, none !
of these incidents rely on an overtemperature delta-T signal to actuate f reactor trip. Therefore, the sensitivity of the incidents to the trip need not be addressed on a case by case basis.
A. RCCA Misalignment I.
i The RCCA drop case presented in Section 15.2.3 of the FSAR is potentially affected by a positive moderator temperature coefficient. Use of a post- j tive coefficient in the analysis would result in a larger reduction in I i
core power level following the RCCA drop, thereby increasing the l probability of a reactor trip. For the return to power with automatic rod control case, a positive coefficient (which would only exist below 70 l
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percent power) would result in a small increase in the power overshoot.
However, the limit,ing 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. Startup of an Inactive Reactor Coolant Loop An inadvertent startup of an idle reactor coolant pump results in a decrea'se in core average temperature. As the most negative values of moderator reactivity coefficient produce the greatest reactivity addition, the analysis reported in the FSAR, Section 15.2.6, represents the limiting case.
C. Excessive Heat Removal Due to Feedwater System Malfunctions The addition of excessive feedwater or the reduction of feedwater tempera-ture are excessive heat removal incidents, which result in a reduction in primary coolant temperature. Thus, they are most sensitive to a negative moderator temperature coefficient. Results presented in Section 15.2.10 of the FSAR based on a negative coefficient represent the limiting case.
D. Excessive Load Increase Incident i An excessive load increase event, in which the steam load exceeds the core power, results in a decrease in reactor coolant system temperature. With l
l the reactor in manual control, the analysis presented in Section 15.2.11 of the FSAR shows that the limiting case is with a large negative modera-tor coefficient. If 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 shows that the minimum DNBR is not sensitive to moderator temperature coefficient.
Therefore, the results presented in the FSAR are still applicable for this i ncident.
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. . j E. Loss of Nomal Fe*dwater, Loss of All AC Power to the Station Auxiliaries ;
i The loss of nomal feedwater and loss of offsite power accidents (Sections l 15.2.8 and 15.2.9 of the FSAR) are analyzed to detemine the ability of l the secondary system to remove decay heat. These events are not sensitive I.
I to a positive moderator temperature coefficient since the reactor trip occurs at the beginning of the transient before the reactor coolant system j~
temperature increases significantly. Therefore, these events are not }
reanalyzed.
F. Inadvertent Operation of ECCS During Power Operation The inadvertent actuation of the ECCS results in a reduction in power, temperature, and pressure, as shown in Section 15.2.14 of the FSAR. Since temperature is decreasing, a positive moderator temperature coefficient i will result in the additic,n of negative reactivity. This will cause power L to decrease even more rapidly, and margin to DNB will increase. Thus, the l FSAR represents the limiting case and this transient need not be ;
reanalyzed. i I
G. Rupture of Main Steam Line/ Accidental Depressurization of the Main Steam System j Since the rupture of a main steam pipe is a temperature reduction tran- f sient, minimum core shutdown margin is associated with a strong negative {
moderator temperature coefficient. The worst conditions for a steamline l break are therefore those analyzed in the FSAR (Section 15.4.2.1). Thi s f is also true for the accidental depressurization of the main steam system (15.2.13).
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H. Major Rupture of a Main Feedwater Pipe This event is analyzed in Section 15.4.2.2 of the FSAR in order to deter-mine the effects of a reactor coolant system heatup following a feedline rupture. In particular, the long-tem ability of the auxiliary feedwater 0531T:6
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reactor trip occurs near the beginning of the transient before the reactor j coolant temperature rises significantly. Therefore, this event need not I
be reanalyzed.
k I. Loss of Coolant Accident (LOCA)
The loss of coolant accident (Section 15.4.1 of the FSAR) is analyzed to determine the core heatup consequences caused by a rupture rf 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 acci- !
dent is not reanalyzed since the Technical Specification requirement that the temperature coefficient be zero or negative at 70 pen:ent power or above ensures that the previous analysis basis for this event is not affected.
II. Transients Sensitive to a Positive Moderator Coefficient or Setpoint Change A. Uncontrolled Boron Dilution This event is analyzed in Section 15.2.4 of the FSAR. An uncontrolled boron dilution during refueling is not impacted by the proposed changes in l l
moderator temperature coefficient or setpoint since the proposed changes do not cpply to the refueling mode of operation. If a boron dilution incident occurs during startup, the FSAR shows that the operator has sufficient time to identify the problem and teminate the dilution before the reactor returns to criticality. Therefore, the value of the moderator coefficient has no effect on a boron dilution incident during startup. No l
l automatic trips are assumed during the boron dilution incident during [
startup, therefore, the setpoint change also has no impact in this mode. l l
The reactivity addition due to a boron dilution at power will cause an l
increase in power and reactor coolant system temperature. Due to the l
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temperature increas2, a positiva moderator confficient would add additional reactivity and increase the severity of the transient. With the reactor in automatic control, however, the rod insertion alams provide the operator with adequate time to teminate the dilution before l
shutdown margin is lost. A boron dilution incident with the reactor in j 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 before shutdown margin is lost is not affected by the moderator coefficient. !
B. Uncontrolled RCCA Bank Withdrawal From a Subcritical Condition ,
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Introduction ;
l A control rod assembly withdrawal incident when the reactor is subcritical l results in an uncontrolled addition of reactivity leading to a power excursion (Section 15.2.1 of the FSAR). The nuclear power response is characterized by a very fast rise tenninateu by the reactivity feedback of the negative fuel temperature (i.e. Doppler) coeffient. The power excur-sion causes a heatup of the moderator and fuel. The reactivity addition due to a positive moderator coefficient results in increases in ;:eak heat flux and peak fuel and clad temperatures. l-i Reactor trip is initiated due to a high neutron flux signal. Thus, the ,
proposed overtemperature delta-T trip setpoint does not affect this analysis.
I Method of Analysis l
l l The analysis was performed in the FSAR for a reactivity insertion rate of l 83 pcm/sec. This assumed reactivity insertion rate is greater than that [
for the simultaneous withdrawal of the combination of the two sequential !{
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control banks having the greatest combined worth at maximum speed l (45 inches / minute). A constant moderator temperature coefficient of l
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+5 pcm/*F was used in tha analysis. The digital computer codes, initial power level, reactor trip instrument delays, and setpoint errors used in the analysis are the same as used in the FSAR.
Results and Conclusions The nuclear power, coolant temperature, heat flux, fuel average tempera-ture, and clad temperature versus time for a 83 pcm/sec insertion rate are shown in Figures 1 and 2. This insertion rate, coupled with a positive moderator temperature coefficient of +5 pcm/*F, yields a peak heat flux slightly higher than that presented in the FSAR. However, the peak heat flux still remains below the full power heat flux. Therefore, the conclusions presented in the FSAR are still applicable.
C. Uncontrolled RCCA Bank Withdrawal at Power Introduction An uncontrolled control rod assembly withdrawal at power produces a mis-match 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. The overtemperature delta-T protection function is assumed to actuate reactor trip, as well as the high neutron flux trip. Due to the increase in Ky . the time of trip will be delayed. This will also reduce the margin to DNB. A discussion l
of this incident is presented in Section 15.2.2 of the FSAR.
l Method of Analysis The transient was reanalyzed employing the same digital computer code and assumptions regarding instrumentation and setpoint errors used for the FSAR. This transient is analyzed at 100, 60, and 10 percent power with a positive moderator temperature coefficient and the revised overtemperature delta-T trip setpoints. A constant moderator coefficient of +5 pcm/*F was used in the analysis for cases based on minimium feedback. The assumption l
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that a positive moderater co fficient exists at full power is conservative I
since at full power the moderator coefficient will actually be negative.
The reactivity coefficients used in the analyses for maximum feedback are the same as those used in the FSAR. Only the setpoint is changed.
Results Figures 3, 4 and 5 show the minimum DNBR as a function of reactivity insertion rate for each of the three power levels with minimum and maximum feedback. The limiting case is a reactivity insertion rate of 1 pcm/sec i from full power initial conditions, assuming minimum reactivity feedback. :
As can be seen from the figures, the positive moderator coefficient and f new setpoint do not lower the DNBR associated with a control rod assembly withdrawal at power below the limit value of 1.30. i Conclusions These results demonstrate that the conclusions presented in the FSAR are j still valid. That is, the core and reactor coolant system are not adversely affected since the high neutron flux and overtemperature aT trips prevent the core minimum DNB ratio from f alling below 1.30 fo'r this !
i ncident. ;
t D. Complete Loss of Forced Reactor Coolant Flow
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Introduction As demonstrated in the FSAR, Section 15.3.4, the most severe loss of flow transient is caused by the simultaneous loss of electrical power to all three reactor coolant pumps. This is more limiting than a reduction in l
flow caused by the loss of only one reactor coolant pump (Section l
15.2.5). The complete loss of flow transient is reanalyzed to detennine the effect of a positive moderator temperature coefficient on the nuclear power transient and the resultant effect on the minimum DNBR reached during the incident. The effect on the nuclear power transient will be 1
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limited to th; initial stages of th2 incident during which reactor coolant temperature increa,ses since this increase is teminated shortly after rec.ctor trip. Reactor trip is generated by an undervoltage signal, thus, the new overtemperature delta-T setpoint will not affect the results.
Method of Analysis Analysis methods and assumptions used in the re-evaluation were consistent with those employed in the FSAR.
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The digital computer codes used to calculate the flow coastdown and resul- l ting system transient are the same as those used to perfom the FSAR analysis. The analysis was done with a constant moderator coefficient of
+5 pcm/*F.
Results For the analysis perfomed with a +5 pcm/*F moderator coefficient, the reactor coolant average temperature increases less than 3*F above the initial value. Therefore, a positive moderator coefficient does not l
appreciably affect the reactor coolant system response or the minimum DNBR 1 reached during the transient, which remains above the limit value of !
l 1.30. Figures 6 and 7 show the nuclear power and heat flux transients, I
the flow coastdown, and the DNBR versus time.
Conclusions A positive moderator temperature coefficient does not appreciably affect the results of the complete loss of flow transient, and the minimum DNBR remains above the limit value for this incident. This case is analyzed
! since it is the most limiting loss of flow case 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 (which also does not rely on overtemperature delta-T protection) will also not be appreciably affected.
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E. Single R7 actor Coolant Pump Locked Rotor j l
Introduction The FSAR (Section 15.4.4) shows that the most severe locked rotor incident is an instantaneous seizure of a reactor coolant pump rotor at 102 percent l
power with three 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 conser- l-vatively assumed to occur at the beginning of the incident. The transient is reanalyzed, however, due to the potential effect on the nuclear power transient and thus on the peak reactor coolant system pressure and fuel !
temperatures. Reactor trip occurs due to a low flow signal, not on an overtemperature delta-T signal. l' Method of Analysis 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 consistent with those employed in the FSAR. An analysis was done at 70 percent power with a moderator coef-ficient of +5 pcm/*F. This case is sufficient to illustrate the impact on l the transient by a positive moderator coefficient, 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 I the FSAR. As shown in the table, the FSAR analysis at full power with a j zero moderator coefficient is more limiting than the 70 percent power case l with a positive moderator coefficient. Therefore, the conclusions pre- [
sented in the FSAR are still applicable.
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F. Loss of External Electrical Load and/or Turbine Trip Introduction l Two cases for this incident, analyzed at both beginning and end of life j conditions, are presented in Section 15.2.7 of the FSAR and involve the j use of two different assumptions. They are: j l
- 1. Full credit for the operation of the pressurizer spray and the i pressurizer power operated relief valves; and :
- 2. No credit for pressurizer spray or power operated relief valves.
l The beginning of life cases are reanalyzed to determine the effects of the {
positive moderator temperature coefficient. The end-of-life case, with ll pressure control, is reanalyzed since reactor trip occurs via an !
overtemperature delta-T signal. The result of a loss of load is a core l power level which momentarily exceeds the secondary system power removal I causing an increase in core water temperature. The consequences of the l reactivity addition due to a positive moderator coefficient are slight '
increases in both peak nuclear power and pressurizer pressure. The new
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l setpoint will delay reactor trip slightly for those cases relying on this f unction. Reactor trip for a loss of load or turbine trip is generated by j either a high pressurizer pressure or overtemperature delta-T signal. !
I Thus, the reanalysis must also include the new overtemperature delta-T !
setpoints.
1 i Method of Analysis l
A constant moderator temperature coefficient of +5 pcm/*F was assumed. l The new overtemperature delta-T trip setpoints were also included in the [
analysis. The method of analysis and assumptions used were otherwise in h
! accordance with those presented in the FSAR.
b l
I l
l 0531T: 6 ,
l
,,-,-..,,n-----,.--.m.-- - --,-. ,, _ , . _ , , - , . . . , , , , . , , --
. . - , . ,r, .
3 i
\ . .
Results The system transient response to a total loss of load from 102 percent power, minimum feedback, with the pressurizer relief and spray valves, is shown in Figures 8 through 10. For this analysis, peak pressurizer pressure reached a value of 2510 psia following a reactor trip on overtemperature delta-T. This is very close to the value presented in the FSAR. The minimum DNBR is reached shortly af ter reactor trip and remains above 1.30.
Figures 11 through 13 illustrate reactor coolant system response to a loss of load assuming no credit for pressure control. Peak pressurizer pres-sure reached 2551 psia following reactor trip on high pressurizer pres-sure. Again, this is essentially the same as the peak pressure reached in the FSAR analysis for this case. The DNBR increases throughout the transient.
The transient response for the case with maximum feedback and pressure control, is shown in Figures 14 through 16. Peak pressurizer pressure reached 2370 psia with a reactor trip on overtemperature delta-T. As for the other cases analyzed, there is little change in peak pressure from the FSAR analysis. The DNBR is never below the initial value and remains above 1.30.
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 or the proposed overtemperature delta-T setpoint since the minimum DNB ratio remains well above the 1.30 limit, and the peak reactor coolant pressure is less than 110 percent of design. Therefore, the conclusions presented in the FSAR are still applicable.
0531T:6
{
q , ,
D. Accidental Depressurization of tha RCS !
l l
Introduction l l
This event is analyzed in Section 15.2.12 of the FSAR using a zero modera- l tor coefficient in order to minimize negative reactivity feedback. A positive moderator temperature coefficient can also be considered as a negative density coefficient and therefore, the density reduction due to f the RCS depressurization causes a positive reactivity insertion and an j
'ncrease in nuclear power. Furthermore, the reactor trip is generated by the overte:perature delta-T function. The RCS depressurization incident is reanalyzed to detennine the impact on the nuclear power transient and l the minimum DNBR.
1 Method of Analysis l The analysis methods and computer code utilized in this analysis are con-sistent with those used in the FSAR with two exceptions. The moderator temperature is assumed to be a constant value of +5 pcm/*F. In addition, manual rod control was assumed. Action of the automatic rod control sys-tem would serve to reduce nuclear pcwer and temperature, which results in higher DNBRs. I I
Results f Figures 17 and 18 show the nuclear power, average temperature, pressure, and DNBR vs. time for the accidental depressurization of the RCS. The positive temperature coefficient causes power and temperature to increase as pressure decreases until a reactor trip occurs on an overtemperature !
delta-T trip signal. The DNBR decreases until a reactor trip occurs and f then increases. Although the minimum DNBR is less than the FSAR value, it i
is well above the limit value of 1.30.
I 0531T:6
. . j Conclusions i
A positive moderator temperature coefficient causes an increase in nuclear l power and temperature for this event. The analysis demonstrates that the l DNBR remains above 1.30 and thus, fuel integrity is maintained. l H. Rupture of a Control Rod Drive Mechanism Housing (RCCA Ejection) l 1
Introduction ;
i The rod ejection transient is analyzed at hot full power and hot zero j power for both beginning and end of life conditions. Since the moderator j temperature coefficient is negative at end of life, only the beginning of life cases are reanalyzed. The high nuclear power levels and hot spot f fuel temperatures resulting from a rod ejection are increased by a-posi-tive moderator coefficient. Reactor trip occurs due to a high neutron flux signal; therefore, the proposed setpoints do not impact the j analysi s. A discussion of this transient is presented in Section 15.4.6 of the FSAR.
i Method of Analysis !
The digital computer codes for analyses of the nuclear power transient and hot spot heat transfer are the same as those used in the FSAR. The ejec- l ted rod worths and transient peaking factors are also the same as reported !
in the FSAR. The moderator coefficient used for this transient was +5 pcm/*F at zero power and at full power. This is still a conservative l
assumpt'.on since the moderator coefficient actually is zero or negative above 70 percent power.
I Results i b
Peak fuel and clad temperatures and nuclear power versus time for both full power and hot zero power are presented in Figures 19 and 20. A summary of the reanalysis results is presented in Table III. The limiting peak hot spot clad temperature, 2538'F, was reached in the hot zero power l
0531T:6
i
. . I casa. Maximum fu21 temperatures are associated with the full power casa. l Although the peak , hot spot fuel centerline temperature for this transient l exceeded the melting point, melting was restricted to less than the inner-most 10 percent of the pellet.
Conclusions Since fuel and clad temperatures do not exceed the fuel and clad limits l
specified in the FSAR, there is no danger of sudden fuel dispersal into l the coolant, or consequential damage to the primary coolant loop. l Therefore, the effects of a positive MTC of the magnitude described above {
is acceptable.
I i
i l
}
I I
0531T:6 l
g
SUMMARY
AND CONCLUSIONS In crder to assess the effect on the accident analysis of operation of J. M. Farley Units 1 and 2 with a slightly positive moderator temperature coefficient, a safety analysis of transients sensitive to a positive moderator coefficient was perfonned. These transients included control rod assembly withdrawal from subcritical, control rod assembly withdrawal at power, loss of reactor coolant flow, loss of external load, RCS depressurization, and control rod ejection. Except as noted, the analyses employed a constant moderator l coefficient of +5 pcm/*F, independent of power level. The results of this f-I study show that the safety criteria are met and are conservative for the acci-l dents investigated at full power, since the proposed Technical Specification j.
requires that the coefficient be zero or negative at or above 70 percent power. l l
' Transients sensitive to the proposed-overtemperature 'uelta-T trip setpoints were control rod assembly withdrawal at power, loss of external load, and RCS depressurization. The reanalysis of those events incorporating the proposed I.
I setpoints was combined with the analyses incorporating the positive MTC. The results show that the safety criteria are met. The proposed overpower delta-T trip setpoints do not impact the safety analysis.
An evaluation was also made 'of the effect of increasing the allowable FaH limit below full power in which the power multiplier is changed from 0.2 to 0.3. The safety analysis is not impacted by this change.
In cormlusion, this study demonstrates that a small positive moderator coef-ficient, the revised setpoints, and the increased F H limit below full power do not result in the violation of any applicable safety limits in the FSAR.
0531T:6
l
. . l l
TABLE I l ACCIDENTS EVALUATED FOR POSITIVE MODERATOR TEMPERATURE COEFFICIENT OR SETPOINT EFFECTS FSAR Accident Time in Life 15.2.1
- Uncontrolled RCCA Bank Withdrawal from a B0C Suberitical Condition 15.2.2 +* Uncontrolled RCCA Bank Withdrawal at Power BOC 15.2.3 RCCA Misalignment BOC l
+* Uncontrolled Boron Dilution B0C 15.2.4
- Partial Loss of Forced Reactor Coolant Flow B0C 15.2.5 Startup of an Inactive Reactor Coolant Loop E0C 15.2.6 15.2.7 +* Loss of External Electrical Load and/or BOC Turbine Trip 15.2.8 Loss of Nonnal Feedwater Loss of All AC Power to the Station 15.2.9 Auxiliaries Excessive Heat Removal Due to Feedwater E0C 15.2.10 System Malfunctions 15.2.11 Excessive Load Increase Incident BOC/E0C
+* Accidental Depressurization of the RCS B0C 15.2.12 15.2.13 Accidental Depressurization of the E0C Main Steam Systcm Inadvertant Operation of the ECCS During BOC 15.2.14 Power Operation
- Complete Loss of Forced Reactor Coolant Flow BOC 15.3.4 80C 15.4.1 Loss of Coolant Accident E0C 15.4.2.1 Rupture of Main Steam Line
- Accidents Reanalyzed for Positive MTC
+ Accidents Reanalyzed for new setpoints l
0531T:6 i L
m --
g
. . l-i-
TABLE I (Con't) t i
ACCIDENTS EVALUATED FOR POSITIVE MODERATOR TEMPERATURE C0 EFFICIENT EFFECTS FSAR Accident Time in Life 15.4.2.2 Major Rupture of a Main Feedwater Pipe -
15.4.4
- Single Reactor Coolant Pump Locked Rotor BOC l 15.4.6 Rupture of a Control Rod Drive Mechanism Housing BOC/E0C j l
l l
l 1
l I
I
(
- Accidents Reanalyzed for. Positive MTC
+ Accidents Reanalyzed for new setpoints j 0531T:6
t i
j i
i TABLE II l
COMPARIS0N OF RESULTS FOR LOCKED ROTOR ANALYSIS This Study FSAR
-5 Moderator temperature coefficient, ak/k/*F 5 x 10 0 Initial power level, percent of nominal 72 102 Peak clad temperature during transient, 'F 1720 2031 Peak reactor coolant system pressure, psia 2463 2675 l
l.
l 0531T:6 . .
i TABLE III
SUMMARY
OF R0D EJECTION RESULTS BEGINNING OF CYCLE (This Study)
Hot Hot Zero Full Power Power Maximum fuel pellet average temperature, *F 3553 3892 Maximum fuel center temperature, *F 4110 4947 Maximum clad average temperature, *F 2538 2254 Maximum fuel enthalpy, cal /gm 151 169 0 <10 Fuel pellet melting, percent h
0531T:6 f
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e
~ .- - ~ g 8 g 8 g 8 N
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cf Fuel Average u
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Uncontrolled RCCA Bank Withdrawal at Pcwer .
i Minimum DNBR vs. Reactivity Insertion Rate
- i 100% Power
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, Figure 8 Loss of Lc,ad/ Turbine Trip - Minimum Feedback With Pressure Control Nuclear Power and DNBR vs. Time 1.2000 l
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. e Figure 9 Loss of Load / Turbine Trip - Minimum Feedback With Pressure Control .
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Y Figure 10 Loss of Load / Turbine Trip - Minimum Feedback With Pressure Control Pressurizer Pressure and Water Volume vs. Time 2600.0 :
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1 Figure 11 Loss of Load / Turbine Trip - Minimum Feedback No Pressure Control Nuclear Power and DNBR vs. Time 51 .2000 0
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l Figure 12 Loss of Load / Turbine Trip - Minimum Feedback No Pressure Control T
avg and T in vs. Time 600.03 --
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Figure 13 Loss of Load / Turbine Trip - Minimum Feedback No Pressure Control l
Pressurizer Prassure and Water Volume vs. Time 2600.0
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Figure 14 Loss of Load / Turbine Trip - Maximum Feedback With Pressure Control Nuclear Power and DNBR vs. Time i
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l Figure 15 Loss of Load / Turbine Trip - Maximum Feedback p With Pressure Control T
avg and T in vs. Time ,
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C g 580.00 - --
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n Figure 16 Loss of Load / Turbine Trip - Maximu'm Feedback With Pressure Control Pressurizer Pressure and Water Volume vs. Time i
i 2500.0 2400.0 -- -
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Figure 17 Accidental Depressurization of the RCS Nuclear Power and Pressurizer Pressure vs. Time 9
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Figure 18 Accidental Depressurization of the RCS T
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Figure 19 RCCA Ejection - BOL HZP Nuclear Power and Fuel Temperatures vs. Time 3 100.0 ;
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