ML20155C103
| ML20155C103 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 09/30/1988 |
| From: | VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.) |
| To: | |
| Shared Package | |
| ML20155C101 | List: |
| References | |
| NUDOCS 8810070238 | |
| Download: ML20155C103 (64) | |
Text
- _ _ _ _ _ _ _ _ _
I' ATTACINENT 1 TECHNICAL SPECIFICATIONS CHANGES FOR INCREASED FQ(Z) WITH 18% STEAM GENERATOR TLBE PLUGGING NORTH ANNA UNITS 1 A m 2 G
4 esi00702]S, f$,hhh 43S 1 DH AD PDC P
1
37
~
- g.
p [;
q l
EQWER DISTRIBUTION LIMITE HEAT FlVX HOT CHANNEL FACTOR-F {Z1 q
LIMITING CONDITION FOR OPERATION
.~
3.2.2 F (Z) shall be limited by the f'J1owing relationships:
g F (Z) s [
] [K(Z)] for P > 0.5 g
F (Z) s (4.38] [K(Z)] for P s 0.5 g
where P - THERMAL POWER RATED THERMAL POWER and X(Z) is the function obtained from Figure 3.2-2 for a given core height location.
APPLICABILITY: MODE 1.
EUDB:
With F (Z) exceeding its limit:
n a.
Reduce THERMAL POWER at least 1% for each 1% F (Z) exceeds the limit g
within 15 minutes and similarly reduce the Power Range Neutron Flux-High Trip Setpoints within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />; POWER OPERATION may proceed for up to a total of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />; subsequent POWER OPERATION may proceed provided the Overpower aT Trip Setpoints (value of K ) have been reduced at least 1% (in aT span) for each 1%
4 I
F (Z) exceeds the limit, g
b.
Identify and correct the cause of the out of limit condition prior to increasing THERMAL POWER above the reduced limit required by a, above; THERMAL POWER may then be increased provided F is demonstrated through incore mapping to be within its limit. n(Z)
NORTH ANNA UNIT 1 3/4 2-5
., _. _ _ _ _.. _ =. _ _ _ _ _. -, _
- - ~
r 1
POWER OISTRIBUTION LIMITS SURVEILLANCE RE0VIREMENTS 4.2.2.1 The provisions of Specification 4.0.4 are not applicable.
4.2.2.2 F (Z) shall be evaluated to determine if F (Z) is within its limit g
9 hy:
a.
Using the movable incore detectors to obtain a power distribution map at any THERMAL POWER greater than 5% of RATED THERMAL POWER.
b.
Increasing the measured Fn(Z) component of the power distribution map by 3% to account fdr manufacturing tolerances and further increasing the value by 5% to account for measurement uncertainties.
c.
Satisfying the following relationship:
N Fq (z) s 2.19 x
Kfz) for P > 0.5 P x N(z)
M Fg (z) s 2.19 x
Kfz) for P s 0.5 N(z) x 0.5 M
where F z) is the measured F z increased by the allowances for manufac9ur(ing tolerances and me9s(ur)ement uncertainty, 2.19 is the l
limit, K(z) is given in Figure 3.2-2, P is the relative THERMA 9 POWER, and N(z) is the cycle dependent function that accounts for power distribution transienti ancountered during normal operation.
This function is given in t.1 Core Surveillance Report as per Specification 6.9.1.7.
M d.
Measuring Fg (z) according to the following schedule:
1.
Upon achieving equilibrium conditions after exceeding the THERMAL POWER at which F of RATED THERMAL POWER *,n(z) was last determined by 10% or mo or 2.
At least once per 31 effective full power days, whichever occurs first, e.
With measurements indicating M
maximum Fg (z) over z g(g) 7 has increased since the previous determination of F" (z) either of g
the following actions shall be taken:
- 0uring power escalatinn, the power level may be increased until a power level for extended operation has been achieved and a power distribution map obtained.
NORTH ANNA - UNIT -1 3/4 2 6
POWER DISTRIBUTION LIMITS SURV IllANCE RE0VIREMENTS (Continued) 1.
F N (z) shall be increased by 2% over that specified in 492.2.2.c,or N
2.
phwe(r) days until 2 successive maps indicate thatshall be measu F
z N
maximum
[Fn (z) is not increasing.
over z K(z) f.
With the relationships specified in 4.2.2.2.c above not being satisfied:
1.
Calculate the percent F (z) exce:ds its limit by subtracting n
one from the measurement 711mit ratio and multiplying by 100:
N l[ maximum Fg (z)
-1]l x 100 for P a 0.5 over z q'
2.19 x Kfz)
I P x N(z) s N
I[ maximum Fg (z)
-1]h x 100 for P < 0.5 p
2.19 x Kfz) over z O.5 x N(z)
J 2.
Either of the following actions shall be taken:
a.
Power operation may continue provided the AFD limits of Figure 3.2-1 are reduced 1% AFD for each percent F (z) g exceeded its limit, or i
b.
Comply with the requirements of Specification 3.2.2 for F (z) exceeding its limit by the percent calculated above.
n g.
The limits specified in 4.2.2.2.c, 4.2.2.2.e and 4.?.:.e.f above are not applicable in the following core plane regi:,ns:
1.
Lower core region 0 to 15 percent inclusive.
4 2.
Upper core region 85 to 100 percent inclusive.
4.2.2.3 When F (z) is measured for reasons othat than meeting the require-f ments of Specincation 4.2.2.2, an overall measured F (z) shall be obtained n
from a power distribution map and increased by 3% to account for manufacturing tolerances and further increased by 5% to acount for measurement uncertainty.
NORTH ANNA - UNIT 1 3/4 2-7 I
1.20 (6.a,1.c )
i (10.93,0.9373) 0.80 S
~
.3 0.60 5
5 e(12.0.0.4566)
_0.40 0.20 0.00-h 8
9 10 11 12 0
1 2
3 4
5 6
COREHEICHT(FT)
Figure 3.2-2 NORWAlllED F0(r) AS A FUNCil0N 0F CORE HEl0HT fiORTH AtitlA - LillT 1 3/4 2-3
POWER DISTRIBUTION LIMITS HEAT FLUX H0T CHAPWEL FACTOR-F in g
LIMITING CONDITION FOR OPERATION 3.2.2 F (Z) shall be limited by the following relationships:
g F (Z) s [Ld2] [K(Z)] for P > 0.5 0
P F (Z) s [4.38] [K(Z)] for P s 0.'
n where P - THERMAL POWER RATED THERMAL POWER and K(Z) is the function obtained from Figure 3.2-2 for a given core height location.
APPLICABILITY: MODE 1.
ACTION:
With F (Z) exceeding its limit:
g a.
Reduce THERMAL POWER at least 17. for each 17. F (Z) exceeds the limit n
within 15 minutes and similarly reduce the Power Range Neutron Flux-High Trip Setpoints within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />; POWER OPERATION may proceed for up to a total of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />; cubsequent POWER OPERATION may proceed provided the Overpower AT Trip Setpoints (value of K ) have been reduced at least l'/. (in AT span) for each 1*/.
4 F (Z) exceeds the limit.
g b.
Identify and correct the cause of the out of limit condition prior l
to increasing THERMAL POWER above the reduced limit required by a, demonstrated through incore mapping to be within its limit. g(Z) above; THERMAL POWER may then be increased provided F is NORTH ANNA - UNIT 2 3/4 2-5 l
l l
POWER DISTRIBUTION LIMITS SURVEILLANCE RE0VIREMENTS 4.2.2.1 The provisions of Specification 4.0.4 are not applicable.
4.2.2.2 F (Z) shall be evaluated to determine if F (Z) is within its limit g
n by:
a.
Using the movable incore detectors to obtain a power distribution map at any THERMAL POWER greater than 5% of RATED THERMAL POWER.
b.
Increasing f ae measured F (Z) component of the power distribution fOr manufacturing tolerances and further map by 3% to account increasing the value by 5% to account for measurement uncertainties.
c.
Satisfying the following relationship:
N x
Kfz) for P > 0.5 Fg (z) s 2.19P x N(z)
N Fq (z) s 2.19 x
Kfz) for P s 0.5 N(z) x 0.5 M
where F z) is the measured F z increased by the allowances for manufac9ur(ing tolerances and me9s(ur)ement uncertainty, 2.19 is the Fl limit, K(z) is given in Figure 3.2-2, P is the relative THERMA 9 POWER, and N(z) is the cycle dependent function that accounts for power distribution transients encountered during normal operation.
This function is given in the Core Surveillance Report as per Specification 6.9.1.7.
N d.
Measuring Fg (z) according to the following schedule:
1.
Upon achieving equilibrium conditions after exceeding the THERMAL POWER at which F (z) was last determined by 10% or more of RATED THERMAL POWER *,gor 2.
At least once per 31 effective full power days, whichever occurs first.
e.
With measurements indicating N
maximum FO (z) over z K(z)
/
N has increased since the previous determination of Fg (z) either of the following actions shall be taken:
- 0uring power escalation, the power level may be increased until a power level for extended operation has been achieved and a power distribution map obtained.
NORTH ANNA - UNIT 2 3/4 2 6
,QWER DISTRIBUTION LIMITS SURVEILLANCE RE0VIREMENTS (Continuedl 1.
F M('z) shall be increased by 2% over that specified in 492.2.2.c,or N
2.
p9we(r) days until 2 successive maps indicate thatshall be measu F
z N
maximum
[Fn (z) is not increasing.
over z K(z) f.
With the relationships specified in 4.2.2.2.c above not being satisfied:
1.
Calculate the percent F (z) exceeds its limit by subtracting n
one from the measurement 711mit ratio and multiplying by 100:
N l[ maximum Fn (z)
-1]l x 100 for P a 0.5 over z
(
2.19 x Kfz) l P x N(z)-
s M
1[ maximum Fn (z) 1 x 100 for P < 0.5 p
2.19 x Kfz) f over z O.5 x N(z)
J 2.
Either of the following actions shall be taken:
a.
Power operation may continue provided the AFD limits of Figure 3.2-1 arn reduced 1% AFD for each percent F (z) g exceeded its limit, or b.
Comply with the requirements of Specification 3.2.2 for F (z) exceeding its limit by the percent calculated above, g
g.
The limits specified in 4.2.2.2.c, 4.2.2.2.e, and 4.2.2.2.f above are not applicable in the following core plane regions:
1.
Lower core region 0 to 15 percent inclusive.
2.
Upper core region 85 to 100 percent inclusive.
4.2.2.3 When F (z) is measured for reasons other than meeting the require-f ments of Speciflcation 4.2.2.2, an overall measured F (z) shall be obtained n
from a power distribution map and increased by 3% to account for manufacturing tolerances and further increased by 5% to acount for measurement uncertainty.
NORTH ANNA - UNIT 2 3/4 2-7
1.20 (6.0.1.C )
1.00w (10.93,0.9373) po,,a e
~
30.60 2
\\
5 (12.0.0.4566) 0.40 7=
0.20 0.00-I 2
3 4
5 6
1 3
g jg jj j2 CORT HEIGHT (FT)
Figure 3.2-2 NORWAlllEDF0(r)ASAFUNCil0NOFCOREHEIGHT l
liORTH AtitiA - 2 3,4,_g i
9 8
t ATTACHMENT 2 LOCA-ECCS SAFETY EVALUATION FOR INCREASED FQ(Z) WITH 18% STEAM GENERATOR TUBE PLUGGING NORTH ANNA UNITS 1 AND 2 E
1.0 INTRODUCTION
A reanalysis of the Emergency Core Cooling System (ECCS) performance for the postulated large-break LOCA has been performed in compliance with Appendix K to 10 CFR 50.
The results of this re-analysis are presented here, and are in compliance with 10 CFR 50.46, "Acceptance Criteria for Emergency Core Cooling Systems for Light Water Reactors."
This analysis was performed with the NRC-approved 1981 model with BART version of the Westinghouse LOCA-ECCS evaluation model (Ref.
1 and 2).
The analysis includes the evaluation model revisions described in Reference 16 and approved by the NRC in Reference 17.
The analytical techniques used are in full compliance with 10 CFR 50, Appendix K.
As required by Appendix K of 10 CFR 50, certain conservative assumptions were made for the LOCA-ECCS analysis.
The assumptions pertain to the conditions of the reactor and associated safety system equipment at the time that the LOCA is assumed to occur, and include such items as the core peaking factors, the containment pressure, and the performance of the Emergency Core Cooling System.
All assumptions and initial operating conditions used in this reanalysis were the same as those used in previous LOCA-ECCS analyses (Ref. 3 and 19), with two exceptions.
The steam generator plugging level was increased to 18% (from 7%
and 15% in References 19 and 3, respectively) and the maximum core peaking factor, FQ, was increased from 2.15 to 2.19.
With these changes incorporated into the analysis, it was founa that the LOCA analysis results continue to meet the 10 CFR 50.46 accepttnce criteria.
2.0 ACCIDENT DESCRIPTION A LOCA is the result of a rupture of the reactor coolant system (RCS) piping or of any line connected to the systen.
The system boundaries considered in the LOCA analysis are defined in the UFSAR.
Sensitivity studies (Ref. 7) have indicated that a double-ended cold-leg guillotine (DECLG) pipe break is limiting.
Should a DECLG break occur, rapid depressurization of the reactor coolant systen occurs.
The reactor trip signal subsequently occurs when the pressurizer low-pressure trip setpoint is reached. A safety injection system 130-J0E-2123S 1
l (SIS) signal is actuated when the appropriate setpoint is reached, activating the hig'1-head safety injection pumps. The actuation and subsequent activation of the Emergency Core Cooling System, which occurs with the SIS signal, assumes the most limiting single-failure event. These countermeasures will limit the consequences of the accident in two ways:
1.
Reactor trip and borated water injection complement void formation in causing rapid reduction of power to a rer,idual level corresponding to fission product decay heat.
No credit is taken in the analysit for the insertion of control rods to shut down the reactor.
2.
Injection of borated water provides heat transfer from the core and prevents excessive clad temperature.
Before the break occurs, the unit is in an equilibrium condition, i.e., the heat generated in the core is being removed via the secondary system.
During blowdown, heat from decay, hot internals, and the vessel continue to be transferred to the reactor coolant system. At the begirging of the blowdown
- phase, the entire reactor coolant system contains subcooled liquid that transfers heat from the core by forced convection with some fully developed nucleate boiling.
After the break develops, the time to DNB is calculated, consistent with Appendix K of 10 CFR 50. Thereafter, the core heat transfer is based on local conditions, with transition boiling and forced convection to steam as the major heat transfer mechanisms.
During the refill period, it is assumed that rod-to-rod radiation is the only core heat transfer mechanism.
The heat transfer between the reactor coolant system and the secondary system may be in either direction, depending on the relative temperatures. For the case of continued heat addition to the secondary side, secondary-side pressure increases and the main safety valves may actuate to reduce the pressure.
Makeup to tha secondary side is automatically provided by the auxiliary feedwater cystem.
Coincident with the safety injection signal, normal feedwater flow is stopped by closing the main feedwater control valves and tripping the main feedwater pumps.
Emergency feedwater flow is initiated by starting the auxiliary feedwater pumps. The secondary-side flow aids in the reduction of RCS pressure. When the reactor coolant system depressurizes to 594 psia, the accumulators begin to inject borated water into the reactor 130-J0E-21235 2
coolant loops.
The conservative assumption is then made that injected accumulator water bypasses the core and goes out through the break until the termination of bypass.
This conservatism is again consistent with Appendix K of 10 CFR 50. In addition, the reactor coolant pumps are assumed to be tripped at the initiation of the accident, and effects of pump coastdown are included in the blowdown analysis.
The water injected by the accumulators cools the core, and subsequent operation of the low-head safety injection pumps supplies water for long-term cooling.
When the refueling water storage tank (RWST) is nearly empty, long-term cooling of the core is accomplished by switching to the recirculation mode of core
- cooling, in which the spilled borated water is drawn from the containment sump by the low-head safety injection pumps and returned to the reactor vessel.
The containment spray system and the recirculation spray system operate to return the containment environment to subatmospheric pressure.
3.0 ANALYSIS The large-break LOCA transient is divided, for analytical purposes, into three phases:
blowdown, refill, and reflood. There are three distinct transients analyzed in each phase, including the thermal-hydraulic transient in the reactor coolant system, the pressure and temperature transient within the containment and the fuel clad temperature transient of the hottest fuel rod in the core. Based on these considerations, a system of interrelated computer codes has been developed for the analysis.
The description of the various aspects of the LOCA analysis methodology is given in WCAP-8339 (Ref. 8).
This document describes the major phenomena modeled, the interfaces among the computer codes, and the features of the codes that ensure compliance with 10 CFR 50, Appendix K.
The SATAfi-V I, C0CO, WREFLOOD, BART, and LOCTA-IV codes, which are used in the LOCA analysis, are described in detail in WCAP-8306 (Ref. 9), WCAP-8326 (Ref 10),WCAP-8171(Ref.
11), WCAP-9695 (Ref.
- 4) and WCAP-10062 (Ref. 5), and WCAP-8305 (Ref. 12),
respectively.
The BART code used for this analysis includes the revisions 130-J0E-2123S 3
described by References 6, 16 and 17. These codes nssess whether sufficient heat transfer geometry and core amenability to cooling are preserved during the time spans applicable to the blowdown, refill, and reflood phases of the LOCA.
The SATAN-VI computer code analyzes the thermal-hydraulic transient in the reactor coolant system during blowdown, and the C0C0 computer code calculates the containment pressure transient during all three phases of the LOCA analysis. The thermal-hydraulic response of the reactor coolant system during refill and reflood is calculated by the WREFLOOD computer code. A mechanistic estimate of the heat transfer coefficient in the core during reflood is provided by the BART computer code.
For the three phases of the LOCA, the LOCTA-IV computer code is used to compute the thermal transient of the hottest fuel rod.
SATAN-VI is used to determine the RCS pressure, enthalpy, and density, as well as the mass and energy flow rates in the reactor coolant system and steam-generator secondary, as a function of time during the blowdown phase of the LOCA. SATAN-VI also calculates the accumulator mass and pressure and the pipe break mass and energy flow rates that are assumed to be vented to the containment during blowdown. At the end of the blowdown, the mass and energy release rates during blowdown are transferred to the C0C0 code for use in the determination of the containment pressure response during this first phase of the LOCA.
Additional SATAN-VI output data from the end of the blowdown, including the core inlet flowrate and enthalpy, the core pressure, and the core power decay transicat, are input to the LOCTA-IV code.
With input from the SATAN-VI code, WREFLOOD uses a system thermal-hydraulic model to determine the core flooding rate (i.e.,
the rate at which coolant enters the bottom of the core), the coolent pressure and temperature, and the quench front height during the refill and reflood phases of the LOCA.
WREFLOOD also calculates the mass and energy flow rates that are assumed to be vented to the containment.
Since the mass flowrate to the containment depends upon the core pressure, which is a function of the containment backpressure, the WREFLOOD and COC0 codes are interactively linked. With the input and boundary conditions from WREFLOOD, the mechanistic core heat transfer model in BART calculates the fluid and heat transfer conditions in the core during reflood.
130-J0E-2123S 4
LOCTA-IV is used throughout the analysis of the LOCA transient to calculate the fuel and clad temperature of the hottest rod in the core. The input to LOCTA-IV consists of appropriate thermal-hydraulic outputs from SATAN-VI, WREFLOOD and
- BART, and conservatively selected initial RCS operating conditions. These initial conditions are summarized in Table 1 and Figure 1.
The axial power shape of Figure 1 assumed for LOCTA-IV is a chopped cosine curve that has been previously verified (Ref. 13) to be the shape that produces the maximum peak clad temperature.
The C0C0 code, which is also used throughout the LOCA analysis, calculates the containment pressure.
Input to C0C0 is obtained from the mass and energy flowrates assumed to be vented to the containment, as calculated by the SATAN-VI cnd WREFLOOD codes.
In addition, conservatively chosen initial containment conditions and an assumed mode of operation for the containment cooling system are input to C0C0. Thesa initial containment conditions and assumed modes of operation are provided in Table 2.
4.0 NON-LOCA SAFETY EVALUATION FOR 18% STEAM GENERATOR TUBE PLUGGING This North Anna Power Station LOCA-EC^S reanalysis has evaluated plant operation at steam generator tube plugging levels of up to 18% based on the acceptance criteria delineated in 10CFR50.46. An evaluation has been performed which concluded that reanalysis of non-LOCA accidents is not required to support this increased tube plugging level provided the measured RCS flow rate remains above the thermal design flow rate assumed for the safety analyses.
Steam generator tube plugging in sufficient quantity can potentially affect non-LOCA safety analysis due to reduced primary system flow. more severe pump coastdown characteristics, and the reduction of the reactor primary coolant system volume.
Primary flowrate becomes a key parameter in DNB limited events (e.g.,
Uncontrolled RCCA Bank Withdrawal at Power) when it falls below the thermal design flowrate.
Pump coastdown characteristics impact analysis results when they become more severe than the conservative values used in the loss-of-flow related analyses.
The reduced primary coolant system volume affects dilution times in uncontrolled boron dilution events.
130-J0E-2123S 5
i A conservative estimate of Ncrth Anna RCS flow versus tube plugging is provided in Reference 18. This estimate is based on past flow measurements taken at the North Anna Power Station for several levels of steam generator tube plugging.
More recent North Anna Unit 1 measurements at greater tube plugging levels l
validate the conservatism of the Reference 18 curve. A re-evaluation of the projection presented in Reference 18 indicates that the conservatively estimated flow rate at the proposed 18% plugging level is approximately equal to the North Anna thermal design flow. Therefore, while measured flow exceeds the ther.nal design flow, the current docketed licensing analyses remain valid for those events in which ficw rate is an important concern.
The loss-of-flow related analyses in Reference 15 used a limiting reactor coolant pump flow coastdown characteristic with the limiting initial thermal design flow rate. Since the conservatively estimated system flow rate equals the thermal design value, the coastdown flows for the 18% plugging level will be bounded by the coastdown flows in the Reference 15 analyses.
The impact of 18% tube plugging on dilution times in the uncontrolled boron dilution events was evaluated with respect to the analyses documented in Reference 15.
Relative to the boron dilution events, the evaluation indicated:
'For uncontrolled dilution during startup, time to criticality is 37 minutes.
This is more than adequate time for the operator to recognize the high count rate signal and terminate the dilution flow.
For uncontrolled dilution at power, the operator has ample time (greater than 15 minutes) after the over-temperature T alarm or trip to determine the cause of dilution, isolate the water source, and initiate reboration before total shutdown margin is lost due to dilution.
Tube plugging levels exhibit no influence on dilution times for the refueling mode of operation, since the steam generator volumes are not a part of the active system.
130-J0E-2123S 6
This evaluation shows that for steam generator tube plugging levels of up to 18 percent, no reanalysis of the DNBR related non-LOCA safety events is necessary and that the currently licensed analyses remain valid.
In the case of the uncontrolled boron dilution events, the available opera',,r response times for the startup aad at power evaluations are reduced but remain well above the minimum acceptance values.
5.0 LARGE BREAK LOCA RESULTS Tables 1 and 2, and Figure 1 present the initial conditions and modes of operation that were assumed in the analysis.
Table 3 presents the time sequence of events, and Table 4 presents the results for the double-ended cold-leg guillotine break for the C = 0.4 and 0.6 discharge coefficients. The D
double-ended cold-leg guillotine break has been determined to be the limiting break size and location based on the sensitivity studies reported in Reference 7.
The analysis resulted in a limiting peak clad temperature of 2165.2*F for 0.4 case, a maximum local cladding oxidation level of 5.77%, and a the C
=
D total core metal-water reaction of less than 0.3%.
The detailed results of the LOCA reanalysis are provided in Tables 3 through 6 and Figures 2A through 18B.
The figures show the following:
1.
Peaking Factor vs. Core Height - Figure 1 shows the chopped cosine power shape used in the analysis.
2.
Mass Velocity - Figures 2A and 2B show the mass velocity at the clad burst and hot-spot locations on the hottest fuel rod for the discharge coefficient used.
3.
Heat Transfer Coefficient - Figures 3A and 3B show the heat transfer coefficient at tue clad burst and hot-spot locations on the hottest rod for the discharge coefficient used. The values of heat transfer coefficient that are shown were calculated by the LOCTA-IV code prior to reflooding and the BART code for the remainder of the transient.
These are based on equations for heat transfer in the nucleate boiling, transition boiling, film boiling, and stean cooling regimes.
130-J0E-21235 7
4.
Core Pressure - Figures 4A and 4B show the calculated pressure in the core for the discharge coefficient used.
5.
Break Flowrate - Figures 5A and 5B show the calculated flowrate out of the break for the discharge coefficient used. The flowrate out of the break is plotted as the sum of flow at both the pressure vessel end and the reactor coolant pump end of the guillotine break.
Figures 6A and 6B show the calculated core 6.
Core Pressure Drop pressure drop for the discharge coefficient used. The core pressure drop is interpreted as the pressure immedtely before entering the core inlet to the pressure just oute', che cors outlet.
7.
Peak Clad Temperature - Figures 7A and 78 show the calculated hot-spot clad temperature transient and the clad temperature transient at the burst location for the discharge coefficient used.
The peak clad temperature for the limiting discharge coefficient of 0.4 is 2165.2*F at the 8.00 ft elevation in the core.
8.
Fluid Temperature - Figures 8A and 88 show the calculated fluid temperature for the hot spot and burst locations for the discharge coefficient used.
9.
Core Flow - Figures 9A and 98 show the calculated core flow, both top and bottom, for the discharge coefficient used.
Figures 10A and 10B show the reactor pressure
- 10. Reflood Transient vessel downcomer and core water levels for the discharge coefficient used.
Figures 11A and 118 show the core inlet velocity for the discharge coefficient used.
- 11. Accumulator Flow - Figures 12A and 128 show the calculated flow for the discharge coefficient used.
The accumulator delivery during blowdown i; discarded until the end of bypass is calculated.
Accumulator flow, however, is established in the refill-reflood calculations.
The accumulator flow assumed is the sum of that injected in the intact cold legs.
130-J0E-2123S 8
12.
Pumped ECCS Flow (Reflood) - Figures 13A and 138 show the calculated flow of the emergency core cooling system for the discharge coefficient used.
Figures 14A and 14B show the calculated 13.
Containment Pressure pressure transient for the discharge coefficient used.
The analysis of this pressure transient is based on the data given in Tables 2, 5, and 6.
14.
Core Power Transient - Figures 15A and 158 show the core power transient calculated by the SATAN-VI code for the discharge coefficient used.
15.
Break Energy Release - Figure 16A and 16B show the break energy released to the containment for the discharge coefficient used.
Figure 17A and 178 show the 16.
Containment Wall Heat Transfer containment wall heat transfer coefficient for the discharge coefficient used.
17.
Fluid Quality - Figures 18A and 18B show the fluid quality at the clad burst and hot-spot locations (location of maximum clad temperature) on the hottest fuel rod (hot rod) for the limiting breaks.
6.0 CONCLUSION
S for breaks up to and including the double-ended rupture of a reactor coolant pipe, and for the operating conditions specified in Tables 1 and 2,
the emergency core cooling systen will meet the acceptance criteria as presented in 10 CFR 50.46, as follows:
1.
The calculated peak fuel rod clad temperature is below the requirement of 2200*F.
130-J0E-21235 9
2.
The amount of fuel element cladding that reacts chemically with water or steam does not exceed 1% af the total amount of Zircaloy in tne reactor.
3.
The clad temperature transient is terminated at a time when the core geometry is still amenable to cooling.
The localized cladding oxidation limits of 17% are not exceeded during or after quenching.
4.
The core remains amenable to cooling during and after the break.
5.
The core temperature is reduced and the long-term decay heat is removed for an extended period of time.
The effects Of increasing the allowable steam generator tube plugging to 18%
has been assessed for existing non-LOCA event analyses.
This evaluc+1on has concluded:
1.
Current analyses for which RCS flow is an important concern remain valid as long as measured flow is greater than the thermal design flow assumed in safety analyses.
2.
The existing loss-of-flow related analyses assume a conservative reactor coolant pump flow coastdown characteristic which accommodates the effect of increased tube plugging on loop flow resistance.
3.
Boron dilution analyses assuming the reduced RCS volume associated with tube plugging result in dilution times which remain adequate for the required operator actions to be performed.
130-J0E-21235 10
10 CFR 50.59 SAFETY EVALUATION The proposed limit changes for steam generator tube plugging and fQ have been reviewed against the criteria of 10 CFR 50.59 and were concluded not to involve any unreviewed safety question. The specific bases for this determination are as follows:
1.
Since the proposed changes involve parameters which are not accident initiators, they will not increase the probability of occurrence of any malfunction or accident previously addressed. The reanalyzed large break LOCA analysis verifies that operation under the revised specifications would also not result in any increase in accident consequences over those in previously accepted analyses.
2.
No new accident types or equipment malfunction scenarios will be introduced as a result of operating in accordance with the revised specifications.
The change which potentially affects physical components in the plant systems (steam generator tube plugging) was explicitly included in the analysis and shown not to produce any new or unique accident precursors.
3.
The margin of safety, as defined in the basis for the plant Technical Specifications, is not reduced. The revised ECCS analysis meets the acceptance criteria of 10 CFR 50.46. Additionally, since evaluation of non-LOCA accidents concluded that acceptance criteria are met when considering the proposed changes, the current margin of safety is maintained for LOCA and non-LOCA accidents.
130-J0E-2123S 11
8.0 REFERENCES
- 1. Letter from J. R. Miller, NRC, to E. P. Rahe, Westinghouse, "Acceptance for Referencing of the 1981 Version of the Westinghouse Large Break E:CS Evaluation Model," December 1, 1981.
- 2. Letter from C. O. Thomas, NRC to E. P. Rahe, Westinghouse, "Acceptance for Referencing of Licensing Topical Report WCAP-9561, BART A-1:
A Computer Code for Best Estimate Analyses of Re'lood Transients," December 21,
- 1983, and Addenda 1 and 2.
- 3. Letter from W.
L. Stewart, Vepco, to U.S. Nuclear Regulatory Commission, Serial No.87-486, dated September 11, 1987.
- 4. Young, M.
Y.
et al.,
BART-A1: A Computer Code for the Best Estimate Analysis of Reflood Transients, WCAP-9695, January 1980.
- 5. Chiou, J.
S.
et al.,
Models for PWR Reflood Calculations using the BAR? Code, WCAP-10002, December 1981.
- 6. Letter from C. 0. Thomas, NRC, to E. P. Rahe Westinghouse, "Acceptance for Referencing of Licensing Topical Report WCAP-10484(P), Spacer Grid Heat Transfer Effects During Reflood," June 21, 1984.
- 7. R. Salvatori, Westinghouse ECCS Sensitivity Studies, WCAP-8356, July 1974.
- 8. F.
M.
Bordelon et al.,
Westinghouse ECCS Evaluation Model - Summary, WCAP-6339, July 1974.
- 9. F.
M.
Bordelon et al.,
SATAN-VI Program:
Comprehensive Space-Time Dependent Analysis of Loss-of-Coolant, WCAP-8306, June 1974.
- 10. F. M. Bordelon and E. T. Murphy, Containment Pressure Analysis Code (C0CO),
WCAP-8326, June 1974.
130-J0E-2123S 12
- 11. R.
D.
Kelly et al.,
Calculational Model for Core Reflooding After a Loss-of-Coolant Accident (WREFLOOD Code), WCAP-8171, June 1974.
- 12. F.
M.
Bordelon et al.,
LOCTA-IV Program: Loss-of-CoolLnt Transient Analysis, WCAP-8305, June 1974.
- 13. Letter from C.
M.
Stallings, Vepco, to E. G. Case, NRC, Serial No. 092, dated February 17, 1978.
- 14. Letter'from C. M. Stallings, Vepco, to E. G. Case, NRC, Serial No. 344, August 9, 1977.
1
- 15. Updated Final Safety Analysis Report - North Anna Power Station Units 1 and 2, Virginia Electric and Power Company, Rev. 6, June 1987.
- 16. M. Y. Young, "Addendum to BART-A1: A Computer Code for the Best Estimate Analysis of Reload Transients (Special Report: Thimble Modeling in Westinghouse ECCS Evaluation Model): WCAP-9561-P, Addendum 3 Revision 1,
July, 1986.
- 17. Letter from Charles E. Rossi NRC, to E. P. Rahe, Westinghouse, "Acceptance for Referencing of Licensing Topical Report WCAP-9561, Addendum 3, Revision 1," August 25, 1986.
- 18. Letter from R.
H. Leasburg, Vepco, to H. R. Denton, NRC, Serial No. 080, February 12, 1982.
- 19. Letter from W. L. Stewart, Vepco, to H. R. Denton, NRC, Serial No.85-077, l
dated May 2, 1985.
i 130-J0E-21235 13
TABLE 1 INITIAL CORE CONDITIONS ASSUMED FOR THE D0jJ8LE-ENDEDCOLD-LEGGUILLOTINEBREAK(DECLG) 4 l
Calculational Input CorePower(HWt)102%of2893-2951 Peak linear power (kW/ft),102% of 12.45 12.70 Heat flux hot-channel factor (F )
2.19 q
Enthalpy rise hot-channel factor (F I) 1.55 H
3 Accumulator water volume (ft, each) 1025 Reactor vessel upper head temperature equal to Thot Limiting Fuel Region and Cycle Cycle Region m
Unit 1 All All regions All regions Unit 2 All 130-J0E-2123S 14
TABLE 2 CONTAINMENT DATA (DRY CONTAINMENT) 6 3
Net Free Volume 1.916 x 10 ft Initial Conditions Pressure (total), psia 9.50 Temperature
- F 90 RWST temperature. *F 35 Outside temperature, F
-10 Containment Quench Spray System Number of pumps operating 2
Runout flowrate (each), gpm 2000 Actuation time, sec 59 Structural Heat Sinks Type / thickness (in.)
Area (ft ), with uncertainty Concrete /6 8,393 Concrete /12 62,271 Concrete /18 55,365 Concreto/24
- 1) 591 Concrete /27 9,404 Concrete /36 3,636 l
Carbon steel /0.375, Concrete /54 22,039 Carbon Steel /0.375, Concrete /54 28,933 Carbon steel /0.50, Concrete /30 25,673 Concrete /26.4 (floor), Carbon Steel /0.25, Concrete /120 12,110 Carbon steel /0.371 160,328 t
Stainless Steel /0.407 10,527 Carbon Steel /0.882 9,894 Carbon Steel /0.059 60,875 130-J0E-21235 15
TABLE 3 TIME SEQUENCE OF EVENTS FOR DECLG CD = 0.4 CD = 0.6 (sec)
(sec)
Start 0.0 0.0 Reactor trip 0.630 0.615 Safety injection signal 2.60 2.05 Accumulator injection 16.7 12.7 Pump injection 27.60 27.05 End of bypass 32.036 26.286 End of blowdown 32.036
'26.286 Bottom of core recovery 45.843 39.826 Accumulator empty 56.516 51.364 130-J0E-2123S 16 C
TABLE 4 RESULTS FOR DECLG C = 0.4 C
0.6 D
D Deak clad temperature, F 2165.2 1971.7 Peak clad locat1.1, ft 8.0 7.25 Local Zr/H O reat, ion 2
(max),%
5.77 3.38 Local Zr/H O location, f t 5.50 6.50 2
Total ;'r/H O reaction, %
< 0. 3
<0.3 2
Hot-rod burst time, sec 40.60 63.80 Hot-rod burst location, f t 5.50 6.50 130-J0E-21235 17
r TABLE 5 REFLOOD MASS AND ENERGY RELEASES DECLG (C = 0.4)
D Total . ass TotalEgergy
+
Time (sec)
Flow Rate (lb/sec)
Flow Rate (10 Btu /sec) 45.843 0.0 0.0 46.468 0.66 0.009 56.810 86.77 1.078 71.860 141.93 1.243 90.360 240.25 1.454 110.760 257.99 1.425 132.860 264.43 1.386 169.510 308.53 1.415 TABLE 6 BROKENLOOPACCUMULATORFLOWTOCONTAINMENTDECL5(C =0.4)
D a
Time (sec)
Mass Flow Rate (lbm/sec) 0.00 4095.55 1.01 3691.57 3.01 3155.57 5.01 2801.97 7.01 2542.34 10.01 2250.67 15.01 1913.20 20.01 1681.07 25.01 1519.24 i
f 30.01 1552.21 i
i f
for energy flowrate, multiply mass flow rate by a constant of 59.62 Btu /lbm.
a i
130-J0E-21235 18
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1 ATTACINENT 3 10CFR50.92h!GNIFICANTHAZARDSEVALUATION FOR i
INCREASED FQ(Z) WITH 18% STEAM GENERATOR TUBE PLUGGING l
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e BASIS FOR NO SIGNIFICANT HAZARDS DETERMINATION The proposed changes do not involve a significant hazards consideration because operation of North Anna Units 1 and 2 in accordance with these change would not:
1.
involve a significant increale in the probability or consequence of an accident previously evaluated. The revised LOCA analysis which supports these changes demonstrated t1at the ECCS acceptance criteria of 10 CFR 50.46 were met. Evaluation of tne non-LOCA accidents has shown that the acceptance criteria for t*,ese accidents are also met with no increase in accident' consequences.
2.
create the possibility of a new or different kind of accident from any accident previously identified.
The proposed changes involve changes in assumptions made for previously evaluated LOCA accidents.
The revised analysis included i.nese parameter changes and demonstrated that they would not cause a new accident.
In addition, the increase in steam generator tube plugging was evaluated for impact upon RCS flow and RCS coolant volume.
It has been demonstrated that the non-LOCA accidents for which these parameters are significant meet applicable acceptance criteria when considering the proposed changes. Thus, the proposed changes will not create the possibility of a new or different kind of accident.
3.
involve a significant reduction in a margin of safety. The revised ECCS analysis meets the requirements of 10 CFR 50.46.
Additionally, the non-LOCA accidents affected by the proposed changes meet their acceptance criteria.
The current margin of safety as established by meeting regulatory requirements (e.g.,
10 CFR 50.46) is therefore maintained for LOCA and non-LOCA accidents.