ML20043C065
| ML20043C065 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 05/29/1990 |
| From: | INDIANA MICHIGAN POWER CO. (FORMERLY INDIANA & MICHIG |
| To: | |
| Shared Package | |
| ML17328A734 | List: |
| References | |
| NUDOCS 9006010264 | |
| Download: ML20043C065 (25) | |
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f- 1 ATTACHMENT 1 TO AEP:NRC:1071I-REVISED PROPOSED UNIT 2 TECHNICAL SPECIFICATION CHANGE -
PAGE 2-2 90060 .
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- DESIGN FLOW - 91,600 GPM/ LOOP DESCRIPTION OF SAFETY LIMITS Pressure Power Tavg Power Tavg Power Tavg Power Tavg (psia) (frac) ( F) (frae) ( F) (frac) ( F) (frac) ( F) 1775' 0,00 615.4 0.98 583.8 1.02 580.9- 1.2 558.1 2000 0.00 631.8 0.86. 60$.8 - 0.96 597.5 1.2 568.5 2100- 0.00 639.1 0.82 614.0 0.96 601.6 1.2 573.1 4 2250 0.00 649.2 0.72 628'6 . 0.98 605.2 1.2 580.4 .,
2400 0.00 '659.0 0.62 642.0 1.1 599.0 1.2- 588.1 660 N
650 - 2400 PSIA 640~ 2250 PSI A 630-N 2500 PSIA T
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i-FRACTION OF; RATED THERMAL POWER Figure 2.1-1 Reactor Core Safety Limits Four Loops in Operation COOK NUCLEAR PLANT - UNIT 2 2-2 AMENDMENT NO.
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l ATTACHMENT 2 TO AEP:NRC:1071I CHANCE PAGES TO DESCRIPTIONS OF ANALYSES PERFORMED BY WESTINCHOUSE ELECTRIC CORPORATION, FOR DONALD C. COOK NUCLEAR PLANT UNIT.2 i
AND PREVIOUSLY SUBMITTED IN AEP:NRC:1071E
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B.3.8A.2 Feedwater System Malfunctions Causing an Increase in Feedwater Flow B.3.8A.2.1 Introduction Addition of excessive feedwater is a means of increasing core power above full power. Such transients are attenuated by the thermal capacity of the secondary plant and of the Reactor Coolant 1 System. De high neutron Dux trip, overpower AT trip and overtemperature AT trip prevent any power increase which could lead to DNBR less than the minimum allowable value in the event that the steam generator high level protection has not been actuated.
Excessive feedwater flow may be caused by the full opening of a feedwater control valve due to a Feedwater Control system malfunction or an operator error. At power conditions, this excess flow causes a greater load demand on the Reactor Coolant System due to increased subcooling in.
the steam generator. With the plant at no load conditions, the addition of cold feedwater will cause a decrease in Reactor Coolant System temperature and thus a reactivity insertion due to the g effects of the negative moderator temperature coefiicient of reactivity.
4-B.3.8A.2.2 Method of Analysis The excessive heat removal due to a feedwater system malfunction transient is analyzed by using ;
the detailed digital computer code LOFTRAN (Reference 5). This code simulates the neutron !
kinetics of the reactor coolant system, pressurizer, pressurizer relief and safety valves, pressurizer spray, steam generator, and steam generator safety. valves. The code computes pertinent plant variables including temperatures, pressures, and power level.
The system is analyzed to demonstrate acceptable consequences in the event of an excessive feedwater addition, due to a control system malfunction or operator error which allows a feedwater control valve to open fully. The following cases have been analyzed:
- 1. Accidental full opening of one feedwater control valve with the reactor at power assuming automatic and manual rod control and a conservatively large negative moderator temperature coefficient of reactivity. /
- 2. Accidental full opening of a feedwater control valve with the reactor at no load (Hot Zero Power) conditions and assuming a conservatively large negative moderator temperature coefficient of reactivity.
B-74 Revision 2
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i The analyses are performed to bound the reduced RCS temperature operation along with the range ;
1 of conditions possible for the uprating of Cook Nuclear Plant Unit 2. This accident is analyzed using the Revised Thermal Design Procedure with the initial conditions shown in Table B.2-4 .
l The reactivity insertion rate following a feedwater system malfunction is calculated with the j following assumptions:
A. Initial reactor power, pressure, and RCS temperatures are assumed to be at their conservative nominal values. Uncertainties in initial conditions are included in the limit DNBR, l B. For the feedwater control valve accident at full power, one feedwater control valve is assumed to malfunction resulting in a step increase to 150% of nominal feedwater flow to one steam !
generator.
C. For the feedwater control valve accident at no load conditions, feedwater control valve (
malfunction occurs which results in an increase in flow to one steam generator from zero to f I
200 percent of the nominal full load value.
D. For the zero load condition, feedwater temperature is at a value of 32 oF. l E. No credit is taken for the heat capacity of the RCS and steam generator thick metal'in attenuating the resulting plant cooldown. i F. The feedwater flow resulting from a fully open control valve is terminated by a steam generator !
high.high level trip signal which closes all feedwater control and isolation valves, trips the l main feedwater pumps and trips the turbme. j i
Normal reactor control system and engineered safety systems are not required to function. The reactor protection system may function to trip the reactor due to overpower or turbine trip on high.high steam generator water level conditions.
-i B.3.8A.2.3 Results In the case of an accidental full opening of one feedwater control valve with the reactor at zero power and the above mentioned assumptions, the maximum reactivity insertion rate is 113 pcm/sec (1 pcm = 10-5 Ak/k).
I B.'l$ Revision 2 1
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An analysis has been performed to demonstrate that the applicable DNB criteria are met. A conservative reactivity insertion rate of 120 pcm/sec was assumed to bound the reactivity insertion rate calculated for the zero power feedwater malfunction analysis. The method of analysis used is the same as discussed in Section B3.1 (Uncontrolled RCCA Withdrawal From ' A Suberitical ,
Condition Analysis), except that the analysis assumed four (4) reactor coolant pumps to be in operation as required by the Cook Nuclear Plant Unit 2 Technical Specifications in Mode 2. A ;
conservatively low value of Doppler Power Defect ( 1000 pcm) was assumed in this analysis.
Although the reactivity insertion rate for the zero power feedwater system malfunction is calculated assuming reactivity parameters representative of EOL core conditions, the DNB analysis was conservatively performed at BOL conditicm.
The DNB analysis performed for the hot zero power feedwater malfunction analysis with an insertion rate of 120 pcm/sec yields a minimum DNBR which remains above the safety analysis limit value for the mixed core.
The full power case (maximum reactivity feedback coefficients with manual rod control) gives the largest reactivity feedback and results in the greatest power increase. Assuming the reactor to be in the automatic rod control mode res'ults in a-slightly less severe transient. The rod control system is not required to function for an excessive feedwater flow event.' -r For all excessive feedwater cases, continuous addition of cold feedwater is prevented by automatic closure of all feedwater isolation valves on a steam generator high-high water level signal. In i
addition, a turbine trip is initiated. A reactor trip on turbine trip was then assumed as a means of terminating the transient analysis. The reactor trip prevents reactor coolant heatup consistent with the cooldown characteristics of the feedwater tralfunction event. The reactor trip on turbine trip was assumed as an anticipatory ttip. If the reactor trip was not assumed, the transient would L progress into a heatup event, in particular, a loss of normal feedwater due to the isolation which occurs on the high high steam generator water level signal. A reactor trip would then be provided by a low-low steam generator water level signal. The reactor trip on turbine trip is not. required for core protection for this event. The results (minimum DNBR) of the feedwater malfunction
- analysis would be essentially unchanged if the reactor trip was not assumed to occur on tubine trip.
l l-Following reactor trip and feedwater isolation, the plant will approach a stabilized condition at hot standby. Normal plant operating procedures may then be followed. The operating procedures would call for operator action to control RCS boron concentration and pressurizer level using the B-76 Revision 2
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CVCS and to maintain steam generator level through control of the main or auxiliary feedwater system. Any action required of the operator to maintain the plant in a stabilized condition will be !
in a time frame in excess of ten minutes following reactor trip.
Transient results, Figures B.3-41A through B.3-44A, show the nuclear power, T avg, pressurizer pressure and DNBR for the full power cases (with and without Rod Control). The DNBR does not drop below the safety analysis limit value.
Since the power level rises during the excessive feedwater flow incident, the fuel temperatures will also rise until after reactor trip occurs. The core heat flux lags behind the neutron flux response due to the fuel rod thermal time constant; hence, the peak heat flux does not exceed 118 percent of its nominal value (i.e., the assumed high neutron flux trip setpoint). The peak fuel temperature !
will thus remain below the fuel melting temperature.
The transient results show that DNB does not occur at any time during the excessive feedwater d flow incident; thus, the ability of the primary coolant to remove heat from the fuel rods is not reduced. The fuel cladding temperature therefore does not rise significantly_above its initial value during the transient. ;
The calculated sequence of events for the increase in feedwater flow for the full power cases are shown in Table B.3 7A. i j
j B.3.8A.2.4 Conclusions !
1 The results'of the analysis show that the DNB ratios encountered for an excessive feedwater ;
addition at power are above the limit value; hence, no fuel or clad damage is predicted.
Additionally, an analysis at hot zero power demonstrates that the minimum DNBR remains above the safety analysis limit for the reactivity insertion rate which occurs at no-load conditions following an excessive feedwater addition.
B.3.8B Excessive Heat Removal due to Feedwater System Malfunctions (Full VANTAGE 5 Core)
Excessive heat removal events due to feedwater system malfunctions are examined primarily to ,
demonstrate core protection. An increase in core power is nonconservative with respect to the DNB transient whereas the reduction in full power average temperature for the reduced B.77 Revision 2 L.
temperature operatien is a benefit for the at power events. The no load temperature does not change due to the reduced temperature and pressure operation. The r' eduction in RCS pressure is nonconservative with respect to the DNB transient. Also, the impact of the revised core limits as well as other design changes associated with the VANTAGE 5 transition as discussed in Section , l B.1 needs to be addressed. As a result, the excessive heat removal events due to feedwater system l malfunctions were analyzed. Feedwater System Malfunctions causing a reduction in feedwater l temperature as well as an increase in feedwater flow are considered.
l B.3.88.1 Feedwater Systern Malfunctions Causing a Reduction in Feedwater Temperature B.3.8B.1.1 Introduction Reductions in feedwater temperature will result in an increase in core power by initially decreasing reactor coolant temperature. Such transients are attenuated by the thermal capacity of the secondary plant and of the RCS. The high neutron flux trip, overtemperature AT trip, and overpower AT trip prevent any power increase which could lead to a Departure from Nucleate Boiling Ratio (DNBR) less than the limit value.
I A reduction in feedwater temperature may be caused by the accidental opening of a feedwater heater bypass valve which diverts flow around a portion of the feedwater heaters. In the event of q an accidental opening of the bypass valve, there is a sudden reduction in feedwater inlet _i temperature to the steam generators. At power, this increased subcooling will create a greater load demand on the RCS, With the plant at no load (Hot Zero Power) conditions, the addition of cold feedwater will cause a decrease in RCS temperature and, thus, a reactivi t y insertion due to the effects of the negative q
moderator temperature coefficient of reactivity. However, the rate of energy change is reduced as i
load and feedwater flow decrease so the transient is less severe than the full power case. The net I
effect on the RCS due to a reduction in feedwater temperature is that the reactor will reach a new
, equilibrium condition at a power level corresponding to the new steam generator AT. l l \
B.3.88.1.2 Method of Analysis This transient is analyzed by computing conditions at the feedwater pump inlet following opening of the heater bypass valve. These feedwater conditions are then used to perform a heat balance through the high pressure heaters. This heat balance gives the new feedwater conditions at the steam generator inlet.
i B 78 Revision 2
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'Ihe following assumptions are made:
- 1. Plant initial power level corresponding to guaranteed NSSS thermal output.
- 2. . Simultaneous actuation of either a low pressure heater bypass valve or a high pressure heater ,
bypass valve and isolation of one string of feedwater heaters. 3 B.3.88.1.3 Results Opening of either a low pressure heater bypass valve or a high pressure heater bypass valve causes a reduction in feedwater temperature which increases the thermal load on the primary system. The -
calculated reduction in feedwater temperature due to opening of a high pressure heater bypass valve is higher than that of the opening of a low pressure heater bypass valve and is less than 60 oF. This reduction in feedwater temperature results in an increase in heat load on the primary system of less than 10 percent of full power. The increased thermal load, due to opening of the high pressure heater bypass valve, would result in a transient very similar (but of reduced magnitude) to that presented in Section B.3.9B for an Excessive Increase in Secondary Steam Flow ]
incident, which evaluates the consequences of a 10 percent step load increase. Therefore, the results of this analysis are not presented. !
B.3.8B.1.4 Conclusior.s The decrease in feedwater temperature transient is less severe than the increase in secondary steam Dow event (Section B.3.9B). Based 'on results presented in Section B.3.9B, the applicable 1i acceptance criteria for the decrease in feedwater temperature event have been met.
B.3.8B.2 Feedwater System Malfunctions Causing an Increase in Feedwater Flow B.3.8B.2.1 Introduction Addition of excessive feedwater is a means of increasing core power above full power. Such i transients are attenuated by the thermal capacity of the secondary plant and of the Reactor Coolant 1 System. The high neutron flux trip, overpower AT trip and overtemperature AT trip prevents any. ,
power increase which could lead to DNBR less than the minimum allowable value in the event that the steam generator high level protection has not been actuated. l Excessive feedwater flow may be caused by the full opening of a feedwater control valve due to.
a Feedwater Control system malfunction or an operator error. At power conditions, this excess flow causes a greater load demand on the Reactor Coolant System due to increased subcooling in the steam generator. With the plant at no load conditions, the addition of cold feedwater will B49 Revision 2
i l
cause a decrease in Reactor Coolant System temperature and thus a resetivity insertion due to the effects of the negative moderator temperature coefficient of reactivity.
B.3.8B.2.2 Method of Analysis ,
The excessive heat removal due to a feedwater system malfunction transient is analyzed by using the detailed digital computer code LOFTRAN (Reference 5). This code simulates the neutron kinetics of the reactor coolant system, pressurizer, pressurizer relief and safety valves, pressurizer spray, steam generator, and steam generator safety valves. The code computes pertinent plant variables including temperatures, pressures, and power level.
The system is analyzed to demonstrate acceptable consequences in the event of an excessive feedwater addition, due to a control system malfunction or operator error which allows a feedwater control valve to open fully. The following cases have been analyzed:
- 1. Accidental full opening of one feedwater control valve with the reactor at power assuming automatic and manual rod control and a conservatively large negative moderator temperature coefficient of reactivity.
- 2. Accidental full opening of a feedwater control valve with the reactor at no load (Hot Zero Power) conditions and assuming a conservatively large negative moderator temperature coefficient of reactivity.
d The analyses are performed .to bound the reduced RCS temperature and decreased RCS pressure operation along with the range of conditions possible for the Cook Nuclear Plant Unit 2 core I
uprating and fuel upgrade. This accident is analyzed using the Revised Thermal Design Procedure with the initial conditions shown in Table B.2 5.
The reactivity insertion rate following a feedwater system malfunction is calculated with the .
following assumptions:
A. Initial reactor power, pressure, and RCS temperatures are assumed to be at their conscretive nominal values. Uncertainties in initial conditions are included in the limit DNBR.
l B 80 Revision 2 i
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B. For the feedwater control valve accident at full power, one feedwater control valve is assumed to malfunction resulting in a step increase to 150% of iaminal feedwater flow to one steam generator, C. For the feedwater control valve accident at no load conditions, feedwater control valve malfunction occurs which results in an increase in flow to one steam generator from zero to ,
200%'of the nominal full load.value. ,
D. For the no load condition, feedwater temperature is at a value of 32 oF.
E. No credit is taken for the heat capacity of the RCS and steam generator thick metal in-attenuating the resulting plant cooldown.
F. The feedwater flow resulting from a fully open control valve is terminated by a steam generator high.high level trip signal which closes all feedwater control and isolation valves, trips the main feedwater pumps and trips the turbine.
Normal reactor control system and engineered safety systems are not required to function. The reactor protection system may function to trip the reactor due to overpower or turbine trip on high-high steam generator water level conditions.
B.3.8B.2.3 Results In the case of an accidental full opening of cae feedwater control valve with the reactor at zero power and the above mentioned assumptions, the maximum reactivity insertion rate is 114.3 pcm/sec (1 pcm = 105 Ak/k).
An analysis has been performed to demonstrate that the applicable DNB criteria are met. A conservative reactivity insertion rate of 120 pcm/sec was assumed to bound the reactivity insertion rate calculated for the zero power feedwater malfunction analysis The method of analysis used is the same as discussed in Section B.3.1 (Uncontrolled RCCA Withdrawal From A Subcritical Condition Analysis), except that the analysis assumed four (4) reactor coolant pumps to be -in operation as required by the Cook Nuclear Plant Unit 2 Technical Specifications in Mode 2. A conservatively low value of Doppler Power Defect (-1000 pcm) was assumed in this analysis.
Although the reactivity insertion rate for the zero power feedwater system malfunction is calculated l assuming reactivity parameters representative of EOL core conditions, the DNB analysis was 1
I conservatively performed at BOL conditions.
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The DNB analysis performed for the hot zero power feedwater malfunction analysis with an insertion rate of 120 pcm/sec yields a minimum DNBR which remains above the safety analysis limit value for the full VANTAGE 5 core.
The full power case (maximum reactivity feedback coefficients with manual rod control) gives the largest reacthity feedback and results in the greatest power increase. Assuming the reactor to be in the automatic rod control mode results in a slightly less severe transient. The rod control system is not required to fenetion for an excessive feedwater flow event.
For all excessive feedwater cases, continuous addition of cold feedwater is prevented by automatic closure of all feedwater isolation valves on a steam generator high.high water level signal, in addition, a turbine trip is initiated. A reactor trip on turbine trip was then assumed as a means of terminating the transient analysis. The reactor trip prevents reactor coolant heatup consistent with the cooldown characteristics of the feedwater malfunction event. The reactor trip on turbine ;
trip was assumed as an anticipatory trip. If the reactor trip was not assumed, the transient would progress into a heatup event, in particular, a loss of normal feedwater due to the isolation which occurs on the high.high steam generator water level signal. A reactor trip would then be provided by a low low steam generator water level signal. The reacto trip on turbine trip is not required for core protection for this event. The results (minimum DNBR) of the feedwater malfunction analysis would be essentially unchanged if the reactor trip was not assumed to occur on turbine trip.
1 Following reactor trip and feedwater isolation, the plant will approach a stabilized condition at hot standby. Normal plant operating procedures may then be followed. The operating procedures would call for operator action to control RCS boron concentration and pressurizer level using the CVCS and to maintain steam generator level through control of the main or auxiliary feedwater j system. Any action required of the operator to maintain the plant in a stabilized condition will be in a time frame in excess of ten minutes following reactor trip.
Transient results. Figures B.3 41B through B.3-44B, show the nuclear power, T-avg, pressurizer pressure and DNBR for the full power cases (with and without Rod Control). The DNBR does not drop below the safety analysis limit value.
Since the power level rises during the excessive feedwater flow incident, the fuel temperatures will also rise until after reactor trip occurs. The core heat flux lags behind the neutron flux response due to the fuel rod thermal time constant; hence, the peak heat flux does not exceed 118 percent B-82 I Revision 2
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of its nominal value (i.e., the assumed high neutron flux trip setpoint). The peak fuel temperature will thus remain below the fuel melting temperature.
The transient results show that DNB does not occur at any time during the excessive feedwater ,
flow incident; thus, the ability of the primary coolant to remove heat from the fuel rods is not reduced. The fuel cladding temperature therefore does not rise significantly above its initial value during the transient.
The calculated sequence of events for the increase in feedwater flow for the full power cases are shown in Table B.3 7B.
B.3.8B.2.4 Conclusions The results of the analysis show that the DNB ratios encountered for an excessive feedwater addition at power are above the limit value; hence, no fuel or clad damage is predicted.
Additionally, an analysis at hot zero power demonstrates that the minimum DNBR reamins above the safety analysis limit for the reactivity insertion rate which occurs at no-load conditions following an excessive feedwater addition.
B.3.9A Excessive Imad Increase (Mixed Core)
B.3.9A.1 Introduction i The excessive increase in secondary steam flow is examined primarily to demonstrate core protection. Since the OTAT setpoint is changed for the VANTAGE 5 transition, the impact of the revised OTAT setpoint needs to be examined for this event. As such, the excessive increase in secondary steam flow is analyzed to determine the impact of VANTAGE 5 transition and other design changes as discussed in Section B.1.
l An excessive load increase incident is defined as a rapid increase in steam flow that causes a power mismatch between the reactor core power and the steam generator load demand. The reactor control system is designed to accommodate a ten percent (10%) step load increase and a five percent (5%) per minute ramp load increase in the range of 15 to 100 percent of full power. Any loading rate in excess of these values may cause a reactor trip actuated by the reactor protection
( system.
l l
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i This accident could result from either an administrative violation such as excessive loading by the operator or an equipment malfur.ction in the steam dump control or turbine speed control.
During power operation, steam dump to the condenser is controlled by reactor coolant condition signals, i.e., high reactor coolant temperature indicates a need for steam dump. A single controller ,
malfunction does not cause steam dump; an interlock is provided which blocks the opening of the ,
valves unless a large turbine load decrease or turbine trip has occurred.
Protection against an excessive load increase accident is provided by the following reactor protection system signals:
- Overpower AT ,
Overtemperature AT
- Power range high neutron flux
- Low pressurizer pressure B.3.9A.2 Method of Analysis This accident is analyzed using the LOFTRAN Code (Reference 5). This code simulates the neutron kinetics, RCS, pressurizer, pressurizer relief and safety valves, pressurizer spray, steam generator, steam generator safety valves, and feedwater system. The code computes pertinent plant variables including temperatures, pressures, and power level. l Four cases are analyzed to demonstrate the plant behavior following a 10 percent step load increase from rated load. These cases are as follows:
A. Manual rod control with minimum moderator reactivity feedback B. Manual rod control with maximum moderator reactivity feedback C. Automatic rod control with minimum moderator reactivity feedback l D. Automatic rod control wi.h maximum moderator reactivity feedback l
For the minimum moderator feedback cases, it was assumed that the core has a zero moderator l temperature coefficient of reactivity and the least negative Doppler only power coefficient curve.
This results in the least inheient transient response capability. The zero moderator temperature coefficient of reactivity bourds a positive moderator temperature coefficient for this cooldowTi ,
l l
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ATTACHMENT 3 TO AEP:NRC:1071I RECONCILIATION OF SUBMITTAL AEP:NRC:1071E i WITH THE THIRTEEN CONDITIONS IN THE SER TO WCAP 10444-P-A 1
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L RECONCILIATION OF THE DONALD C. COOK NUCLEAR PLANT, UNIT'2 CYCLE 8 SUBMITTAL, AEP:NRC:1071E, WITH THE.13 CONDITIONS IN THE SER-TO VESTINGHOUSE'S VANTAGE 5 FUEL TOPICAL REPORT WCAP 10444-P-A- -
NRC LIMITATIONS ON USE OF WCAP-10444 The NRC Staff reviewed Westinghouse's WCAP-10444, " Reference _ Core Report VANTAGE ~5 Fuel Assembly," and concluded in a Staff Safety Evaluation deport (SER), Reference 1, that the generic' topical; report was'an acceptable reference-to support plant-specific. applications =for
~
+
use of VANTAGE 5 fuel, provided thirteen -conditions' identified in :the SER were addressed by the licensees. These thirteen conditions were' ,
appropriately considered in Attachment 4'to Reference 2,--
AEP:NRC: 1071E, Each of the thirteen conditions is addressed below,- r and a reference is provided for the. specific section of Reference 2 or s other appropriate documentation where the condition is discussed, i 1 .
L-NRC-Condition 1:
The statistical convolution method described in WCAP-10125 for the- >
evaluation of initial' fuel rod to nozzle growth has .not .been approved. '
This method should not be used in VANTAGE 5; l Response:
, The statistical convolution method was not used for fuel evaluation.
To' determine the initial. fuel rod-to-nozzle-growth gap from fuel rod irradiation growth, the worst-case fabrication tolerances were used to evaluate fuel rod performance as' discussed in.Section 2.0, Page 5 of i
Reference 2. This evaluation was in compliance with Condition 1 of >
l the VANTAGE 5 SER (Reference 1).
NRC Condition 2:
For each plant application, it must be demonstrated that'the LOCA/ seismic loads considered in WCAP-9401 bound the plant in-question; otherwise additional analysis will be required to demonstrate the fuel assembly structural integrity,
ll
Response
The LOCA/ seismic loads considered in WCAP-9401 do not bound Donald C.
Cook Nuclear Plant Unit 2. An analysis was performed to demonstrate the fuel assembly structural integrity and the results are presented in Section 2.0, Page 9 and Section 5.3, Page 37, of Reference 2.
NRC Condition 3. ;
An irradiation demonstration program should be performed to provide {
early confirmation performance data for the VANTAGE 5 design. j
Response
A demonstration program was successfully performed to determine early performance data on the VANTAGE 5 fuel assembly design features. .The VANTAGE 5 demonstration program at commercial reactors is described in Section 1.0, Page 3 of Reference 2. ,
NRC Condition 4: I For those plants using the ITDP, the' restrictions enumerated in Section 4.1 of this report (SER) must be addressed-and information regarding measurement uncertainties must be provided.
Response
The Revised Thermal Design Procedure (RTDP) (WCAP-11397-P-A), an l
extension of ITDP, was used for Donald C. Cook Nuclear Plant Unit 2.
The restrictions for ITDP for this condition are also applicable to '
RTDP. Westinghouse has addressed the restrictions enumerated in Section 4.1 of the NRC's generic VANTAGE 5 SER in a Westinghouse letter to the NRC Staff in March 1985, Reference 3. The ITDP/RTDP instrument uncertainty methodology used for Donald C. Cook Nuclear Plant, Unit 2 with VANTAGE 5 fuel is presented in WCAP-11656, Reference 4. Measurement uncertainties are documented in WCAPs-12576/12577 and provided to the NRC Staff in Attachment 4 of this submittal, Reference 5.
NRC Condition 5: '
Thc WRB-2 correlation with a DNBR limit of 1.17 is acceptable for application to 17x17 VANTAGE 5 fuel. Additional data and analysis are required when applied to 14x14 or 15x15 fuel with an appropriate DNBR limit. The applicability range of VRB-2 is specified in Section 4.2.
i M '
Response
VANTAGE 5 17x17 fuel is proposed to be used at Donald C. Cook Nuclear Plant Unit 2. As described in Section 4.0, Page 15 of Reference 2, the WRB-2 correlation with a DNBR limit of 1.17'was used for the i VANTAGE 5 fuel with the NRC-approved ITDP methodology. The WRB ,
correlation is supported by'the DNB test: data contained.in Appendix A
~
to WCAP-10444-P A, and was applied within its. approved range-of applicability for Donald C. Cook Nuclear Plant Unit 2. i NRC Condition 6:
For 14x14 and 15x15 VANTAGE 5 fuel designs, separate analyses will be-required to determine a transitional mixed core penalty. Juus mixed core penalty and plant-safety margin to-compensate for the penalty, should be addressed in the plant Technical Specifications-Bases.
Response
As noted above in Condition 5, VANTAGE 5 17x17 fuel is-proposed for
- Donald C Cook Nuclear Unit 2.- The Westinghouse transition core DNB-methodology as applied to the Donald C.' Cook Nuclear Unit.2 is discussed in Section 4.0 of Reference-2. Reference-2 incorporates the- !
NRC-approved change to the generic._ VANTAGE 5 transition core penalty discussed in the Westinghouse October 1987 letter to the NRC Staff regarding transition core effects, Reference 6. The transition core-penalty is covered by the margin maintained between thel design and -i safety analysis DNBR. limits. The' proposed changes to Technical Specification 2.1.1, Reactor Core Safety Limits-Bases, in Reference 9 3 (see also Appendix-A to Reference 2) address these margins. 1 i
NRC Condition 7:
Plant-specific analysis should be performed to show that the:DNBR limit will not be violated with the higher value of F 6H' '
Response
The Core DNB methodology as applied to Donald C. Cook Nuclear Plant Unit 2 with VANTAGE-5 fuel is presented in Section 4.0, Page 15, of Reference 2. Section 5.1 of Reference 2 contains plant-specific analysis results in Section 5.0 which-support the use of'F of 1.65 ;
forVANTAGE15fuelduring'thetransitionperiodandwitha# Null-core of VANTAGE 5 fuel. All safety criteria are met with an F f 1.65 as demonstrated in Section 5.0. AH i
i
- .1
l 7
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- k NRC Condition 8:
The plant-specific safety analysis for the steam system piping failure event should be performed with the assumption of loss of offsite power if that is the most conservative case.
Response
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. The Donald C. Cook Nuclear Unit 2 plant specific analysis for the Main Steam Piping Rupture event has been performed for the VANTAGE 5 Fuel ,
transition. This analysis included both the with f and without power. ;
cases. The DNB design basis was confirmed for the fuel' transition as presented in Section B.3.11-of' Appendix 3 of Reference 2.
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NRC Condition 9:
?
With regard to the RCS pump shaft seizure accident,- the fuel failure criterion should be the 95/95 DNBR limit. The mechanistic. method mentioned in WCAP-10444 is not acceptable. !
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Response
Consistent with the original design basis requirements for Cook Nuclear Plant Unit 2, a maximum. clad temperature limit of 2700 F and total metal water reaction at the hot spot less than 16% have been used as the acceptance criteria for the RCS pump shaft. seizure'(locked rotor) event analyzed.for the VANTAGE 5 transition. The. purpose-of this limit is to ensure that a coolable geometry is maintained and.is further described in Reference 8. The results of the~ analysis presented in Section B.3.5.2 of. Appendix B to Reference 2 showed that 1 the above criteria are satisfied for the locked rotor event analyzed for the VANTAGE 5 transition.
Westinghouse also performed a separate analysis, using the LOFTRAN, t FACTRAN and THINC Computer Codes to determine the number of rods that !
experience DNB during the locked rotor accident for the Cook Nuclear Plant Unit 2 VANTAGE 5 fuel transition. Any rods,which. violated the 95/95 DNBR lim 1t were assumed to fail in this analysis. This analysis
Unit 2 Technical Specifications was used in' the safety. analyses for i the VANTAGE 5 fuel.
NRC Condition 11:
The LOCA analysis performed for the refergnce plant with higher Fq 'of _7 2.55 has shown that the PCT limit of 2200 F is violated during transitional mixed core. Plant specific LOCA analysis must be done-to show that with the appropriate value of F the 2200 F criteria can be ,'
met during use of transitional mixed core.q
Response
t In accordance with Condition 11 of the VANTAGE 5 Safety Evaluation ;
Report (SER), Donald C. Cook Nuclear Plant Unit 2 specific LOCA 4 analyses were performed with consideration of transitional core !
effects. The LOCA.results referred to in Condition 11'were for large j break LOCA and the new large break'LOCA analysis _is summarized in Section'5.2.1 of Reference 2 and reported ~in detail in Section C.2 of i Appendix C to Reference 2. As described therein, the ECCS acceptance criteria of 2200 F is met for both of the following limiting ;
conditions:
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- 1. Cp - 0.6, 3588 MWt Core Power, High Temperature (T hot - 615.2'F).
High Pressure (PRCS- ~
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Minimum S1 with RHR crosstle valves open. }
Limiting break case, i.e., this case had highest PCT for all t cases analyzed. 1 I
PCT - 2140 F + 50 F (transition core penalty) - 2190 F
- 2. CD - 0.6, 3413 MWt Core Power, High Temperature (Thot - 611.2D),
High Pressure (PRCS - 2313 psia), Fq - 2.335,
[AH-1.644, Minimums 1withRHRcrosstievalvesclosed.
PCT - 2090 F + 50 F (transition core penalty) - 2140 F' NRC Condition 12:
Our SER on Westinghouse's extended burnup topical report WCAP.10125 is not yet complete; the approval of the VANTACE 5' design for operation to extended burnup levels is contingent on NRC_ approval of WCAP.10125.
However, VANTACE 5 fuel may be used to those burnups to which.
Westinghouse fuel is presently operating. Our review of-the
[. Westinghouse extended burnup topical report has not identified any=
safety issues with operation to the burnup value given in the extended
~
burnup report.
Response
WCAP.10125 has been approved (see Reference 7). The extended burnup methodology contained in this topical has been applied-and is
- addressed in Section.2.0, pago 5 of Appendix D to Reference 2.
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P NRC Condition 13 >
Recently, a vibration problem has been reported in a French reactor-having 14. foot fuel assemblies; vibration below the fuel assemblies in the lower portion of the reactor vessel is damaging the movable incore instrumentation probe thimbles. The staff is currently evaluating the implications of this problem to other core having 14 foot long fuel bundle assemblies.- Any limitations to the 14. foot core design resulting from the staff evaluation must be addressed in plant specific evaluations,.
Response
Donald C. Cook Nuclear Unit 2 has 12. foot long fuel assembly bundles..
There have been no fuel assembly vibration problems observed in 12 foot corea. Therofore, the above condition is not. applicable.
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< L1, _ INRC Staffiletter.(C. O.1 Thomas) to Westinghouse (E. P, Raho) -
Re: " Acceptance for Referencing of Licensing Topical ~Repert,.
WCAP 10444,."VANTACE 5 Fuel Asssably, "' undated.
\
- 2. Attachment' 4' ho the Donald C.' Cook Nuclear Plant, Unit 2 Cycin 8 Reload Licensing Submittal, AEP:NRC:1071E, dated February 6, 1990, and modifications to Attachment 4 included in letter AEP:NRC:1071H, dated April 6,.1990 and in Attachment 2 of AEP:NRC:10711, this submittal.
- 3. Westinghouse letter, E. P. Rahe, Jr. , to C. O. Thomas (NRC),
Response to Request No. 1 for Additional Information on WCAP 10444 entitled "VANTACE 5 Fuel Assembly" (Proprietary)
NS NRC 85 3014, dated March 1, 1985.
- 4. " Westinghouse Improved Thermal Design Procedure Instrument Uncertainty Methodology for the Virgil C. Summer Nuclear Power Station," WCAP 11656, December 1987.
- 5. Attachment 4 to letter AEP:NRC:10711, this submittal,
" Westinghouse Revised Thermal Design Procedure Instrument to Uncertainty Methodology for American Electric Power D. C. Cook Unit 2 Nuclear Power Station," WCAP 12576 (Proprietary) and WCAP 12577 (Non Proprietary).
- 6. Westinghouse letter, W. J. Johnson to M. W. Hodges (NRC),
NS NRC 87-3208, " VANTAGE 5 DNB Transition Core Effects," October 2, 1987.
- 7. " Extended Burnup Evaluation of Westinghouse Fuel," WCAP 10125 P A, December 1985.
- 8. NS NRC-89 3466, "Use of 2700 F PCT Acceptance Limit in Non 14CA-Accidents," Letter from W. J. Johnson (Westinghouse) to Mr. Robert C. Jones (NRC), October 23, 1989.
- 9. Attachment 2 to AEP:NRC:1071E, Donald C. Cook Nuclear Pinnt Unit 2 Cycle 8 Reload Licensing Submittal, dated February 6, 19 %
05/24/90 s TAC:90027.C I 1
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ATTACildENT 4A TO AEP:NRC:10711 COPIES OF Tile PROPRIETARY VERSION OF "I'STINC110USE REVISED TilERMAL DESIGN PROCEDURE INSTRUMENT UNCERTAINTY METil0D0 LOGY FOR AMERICAN ELECTRIC POWER, D. C. COOK UNIT 2 NUCLEAR POWER STATION,"
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a 1 ATTACHMENT 45 TO AEP:NRC 10711 TEN COPIES OF THE NON PROPRIETARY VERSION OP
" WESTINGHOUSE REVISED THERMAL DESIGN PROCEDURE INSTRUMENT UNCERTAINTY METHODOLOGY FOR AMERICAN ELECTRIC POWER, D. C. COOK UNIT'2 NUCLEAR' POWER STATION,*
WCAP.12577 L
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