ML20215B602
| ML20215B602 | |
| Person / Time | |
|---|---|
| Site: | Catawba  |
| Issue date: | 06/12/1987 |
| From: | DUKE POWER CO. |
| To: | |
| Shared Package | |
| ML20215B599 | List: |
| References | |
| TAC-65589, TAC-65590, NUDOCS 8706170379 | |
| Download: ML20215B602 (200) | |
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l TABLE 3.4-1 y
REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES VALVE NUMBER FUNCTION
\\
NI59 Accumulator Discharge NI60 Accumulator Discharge NI70 Accumulator Discharge NI71 Accumulator Discharge NI81 Accumualter Disenarge NI82 Accumulator Discharge NI93 Accumulator Discharge NI94 Accumulator Discharge NI124 Safety Injection (Hot Leg)
NI125 Residual Heat Removal (Hot Leg)
NI126 Safety Injection (Hot Leg)
NI128 Safety Injection (Hot Leg)
NI129 Residual Heat Removal (Hot Leg)
NI134 Safety Injection (Hot Leg)
NI156 Safety Injection (Hot Leg)
NI157 Safaty Injection (Hot Leg)
NI159 Safety Injection (Hot Leg) 1 a
NI160 Safety Injection (Hot Leg) l NI165 Safety Injection / Residual Hest Removal (Cold Leg)
{
NI157 Safety Injection / Residual Heat Removal (Cold Leg)
NI169 Safety Injection / Residual Heat < Removal (Cold Leg)
NI171 Safety Injection / Residual Heat Removal (Cold Leg) i NI175 Safety Injection / Residual Heat Removal (Cold Leg)
{
NI176 Safety Injection / Residual Heat Removal (Cold Leg)
NI180 Safety Injection / Residual Heat Removal (Cold Leg)
NI181 Safety Injection /Resioual Heat Removal (Cold Leg)
"NI248 Upper Head Injection NI249 Upper Head Injection NI250 Upper Head Injection g
NI251 Upper Head Injection
> Note i NI252 Upper Head Injection NI253 Upper Head Injection u
a Note 1: Upon the disconnection of the UHI System from the Reactor Coolant System, this specification is no longer applicable CATAW8A - UNITS 1 & 2 3/4 4-22
1 3/4.5 EMERGENCY CORE COOLING SYSTEMS _
j 3/4.5.1 ACCUMULATORS COLD LEG INJECTION LIMITING CONDITION FOR OPERATION 3.5.
1.{ ach cold leg injection accumulator shall be OPERABLE with:
a.
The discharge isolation valve open, b.
A contained borated water volume of between 7853 and 8171 gallons, A boron concentration of between 1900 and 2100 ppm, c.
d.
A nitrogen cover pressure of between 385 and 481 psig, and e.
A water level and pressure channel OPERABLE.
APPLICABILITY:
MODES 1, 2, and 3*. ( CHI system operable)
ACTION:
a 2
With one cold leg injection accumulator inoperable, except as a result.
a.
of a closed isolation valve, restore tne inoperable accumulator to OPERABLE status within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or be in at least HOT STANDBY within
^
the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN withir. the following.6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, b.
With one cold leg injection accumulator inoperable due to the isolation valve being closed, either immediately open the isolation valve or be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTOOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
JURVEILLANCE REOUIREMENTS l
4.5.1.'.1d ach cold leg injection accumulator shall be demonstrated OPERAB ::
a.
At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> by:
1)
Verifying, by the absence of alarms, the contained borated water volume and nitrogen cover pressure in the tanks, and 2)
Verifying that each cold leg injection accumulator isolation valve is open.
" Pressurizer pressure above 1000 psig.
CATAWBA - UNITS 1 & 2 3/4 5-1
l i
3/4.5. EMERGENCY CORE COOLING SYSTEMS _
3 l
3/4.5.1 ACCUMULATORS COLD LEG INJECTION LIMITING CONDITION FOR OPEPATIOPl l
3.5.
.'1.1 ach cold leg injection accumulator shall be OPERABLE with:
1 a.
The discharge isolation valve open,
.b.
A contained borated water volume of betwe n 28Sf s JtTf allons, 4
c.
A baron concentration of between 1900 2100 m,
d.
A nitrogen cover-pressure of betwe n
sig, and e.
A water level and pressure channel OPERABLE.
j APPLICABILITY:
MODES 1, 2, and 3*.
(UHI physically disconnected; Cold Leg
)
Accumulators and discharge paths suitably ACTION:
modified)
With one cold leg injection accumulator inoperable, except as a respit a.
P of a closed isolation valve, restore the inoperable accumulator to q
OPERABLE status within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or be in at l' east HOT STAN0BY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
b.
With one cold leg injection accumulator inoperable due to the l
isolation valve being closed, either immediately open the isolation valve or be in at least HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
SURVEILLANCE REOUIREMENTS m
4'
- 4. 5. '.1j.1 tach cold leg injection accumulator shall be demonstrated OPER/ Lt:
)
t a.
At least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> by:
1)
Verifying, by the absence of alarms, the contained borated water volume and nitrogen cover pressure in the tanks, and 2)
Verifying that each cold leg injection accumulator isolation valve is cpen.
- Pressurizer pressure above 1000 psig.
CATAWBA - UNITS 1 & 2 3/4 5-1
I R
EMERGENCY CORE COOLING SYSTEMS l
SURVEILLANCE REQUIREMENTS (Continued) b.
At least once per 31 days and within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> aftt:' Jach solution volume increase of greater than or equal to 75 gallons by verifying j
the boron concentration of the accumulator solution; c.
At least once per 31 days when the Reactor Coolant System pressure 4
is above 2000 psig by verifying that power is removed from the isolation valve operators on Valves NI54A, NI65B, NI76A, and NI88B and that the respective circuit breakers are padlocked; and d.
At. least once per 18 months by verifying.that each cold leg injection accumulator isolation valve opens automatically under each of the following conditions:*"
j 1)
When an actual or a simulated Reactor Coolant System pressure j
signal exceeds the P-11 (Pressurizer Pressure Block of Safety Injedtion) Setpoint, and
{
Upon receipt of a Safety Injection test signal.
4.5.1.123 ch cold leg inject' ion accumulator water level and pressure channel
)
t shall b
.onstrated OPEP.ABLE-At least once per 31 days by the performance of an ANALOG CHANNEL j
a.
OPERATIONAL TEST, and b.
At least once per 18 months by the performance of a CHANNEL CALIBRATION.
"" This surveillance need not be performed until prior to entering HOT STANCSY following the Unit I refueling.
CATAWBA - UNITS 1 & 2 3/4 5->g:
l
f.
EMERGENCY CORE COOLING SYSTEMS UPPER HEAD INJECTION (Deleted upon the physical disconnection of the WI System from the Reactor Coolant System)
}
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TABLE 3.8-1A (Continued)
UNIT 1 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES DEVICE NUMBER & LOCATION SYSTEM POWERED J
2.
600 VAC MCC (Continued) 1EMXD-F02C Drimary Bkr Pressurizer Power-Operated Relief' Backup Fuse Isol Viv 1NC318
.1EMXD-F05A Primary Bkr Pressurizer Power-Operated Relief Bat:kup Fuse Isol Viv 1NC35B 1EMXD-F05B Primary Bkr Rx Bldg Drain Hdr Inside Cont Backup Fuse Isol Viv 1KC429B 1EMXD-F05C.
Primary Bkr NCDT Hx Cing Water Return Inside-
{
Backup Fuse Isol Viv 1KC332B 1EMXD-F06A Primary Bkr NC Pump 1B Thermal Barrier Outlet 4
Backup Fuse Isol Viv 1KC3648 l
1EMXD-F06B Primary Bkr NC Pumps Rtn Hdr Inside Cont-4 Backup Fuse Isol Viv IKC424B
~1EMXK-F01A Primary Bkr UHI Check Viv Test Line Inside
,Sete 1 b Backup Fuse Cont Isol Viv 1N1266A 1EMXK-F01B Primary Bkr Upper Cont Vent Units Return Backup Fuse Cont Isol Viv 1RN429A 1EMXK-F01C to PORV INC34A From Primary Bkr Backup N2 Backup Fuse Accum Tnk 1A Viv 1NI43BA 1EMXK-F02A Primary Bkr NC Pump 1A Thermal Barrier Backup Fuse Outlet Isol Viv 1XC394A 1EMXK-F02B l
Primary Skr Lower Cont Vent Units Return Backup Fuse Cont Isol Viv 1RN484A f
l'pon removal of cable from power source associated with the deletion Note 1: of UHI, this specification is no longer applicable CATAWBA - UNITS 1 & 2 3/4 B-23 j
TABLE 3.8-1A (Continued) 1.l-UNIT,1 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT HOTECTIVE DEVICES -
DEVICE NUMB:.R LOCATION SYSTEM POWERED 2.
600 VAC MCC (Continued) 1EMXK-F02C Primary Skr NV Supply to Pressurizer Viv Backup Fuse INV037A.
1EMXK-F03A Primary Bkr S/G C Blowdown 8.ine Sample Backup Fuse Inside Cont 1501 V1v 1NM210A I
Primary Bhr S/G A Upper Shell Sample Inside l
Backup Fuse Cont Isol Viv 1NM187A 1EMXK-F048 Primary Skr S/G A Blowdown Line Sample Backup Fuse Inside Cont Isol Viv 1NM190A a
Primary Bkr S/G C Upper Shell Sample Backup Fuse Inside Cont Isol V1v 1NM207A
{
1EMXK-F06A Primary Skr Hydrogen Skimmer Fan 1A Backup Fuse Inlet Viv IVX1A 1EMXK-F07C Primary Bkr Electric Hydrogen Recombiner l
Backup Fuse Power Supply Panel 1A 1EMXK-F09A Primary Bkr Accumulator..lA Discharge Isol Backup Fuse Viv 1NI54A i
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Primary Bkr UHI Check Viv Test Line Inside
,Sote 1 L Backup Fuse Cont Isol Viv 1NI267A 1EMXK-F09C Primary Bkr NC Pump Oil Fill Header Backup Fuse Cont Isol viv 1NC196A Note 1: Upon removal of cable from power source associated with the deletion of Lill, this specification is no longer applicable l-1 CATAWBA - UNITS 1 & 2 3/4 B-24
TABLE 3.8-1B (Continued)
UNIT 2 CONTAINMENT PENETRATION CONDUCTOR OVERCURRENT PROTECTIVE DEVICES DEVICE NUMBER & LOCATION SYSTEM POWERED 2.
600 VAC MCC (Continued) 2EMXD-F02C Primary Bkr Pressurizer Power-Operated Relief Backup Fuse Isol Viv 2NC31B 2EMXD-F05A Primary Bkr Pressurizer Power-Operated Relief Isol Viv 2NC35B Backup Fuse 2EMXD-F05B Primary Bkr Rx Bldg Drain Hdr Inside Cont Backup Fuse Isol Viv 2KC429B 2EMXD-F05C Primary Bkr NCDT Hx Cing Water Return Inside
~
Backup Fuse Isol Viv 2KC332B F
2EMXD-F06A Primary Bkr NC Pump 2B Thermal Barrier Outlet Backup Fuse Isol Viv 2KC364B 2EMXD-F06B Primary Bkr NC Pumps Rtn Hdr Inside Cont Backup Fuse Isol Viv 2KC424B
~ EMXK-F01A 2
Primary Bkr UHI Check Viv Test Line Inside, gate t
_ Backup Fuse Cont Isol Viv 2NI266A 2EMXK-F01B Primary Bkr Upper Cont Vent Units Retu n Backup Fuse Cont Isol Viv 2RN429A 2EMXK-F01C to PORV 2NC34A From Primary Bkr lackup N2 Backup Fuse Accum Tnk 2A Viv 2NI438A 2EMXK-F02A Primary Bkr NC Pump 2A Th7rmal Barrier Backup Fuse Outlet Icol V)v 2KC394A 2EMXK-F02B Primary Bkr Lower Cont Vent Units Return ockup Fuse Cont Isol Viv 2RN484A Upon removal of cable form power source associated with the deletion of OfI, Note 1: this specification is no longer applicable CATAWBA - UNITS 1 & 2 3/4 8-46
o l
TABLE 3.8-18 (Continued)
UNIT 2 CONTAINMENT PENETRATION CONOUCTOR OVERCURRENT DROTECTIVE DEVICES i
DEVICE NUMBER & LOCATION SYSTEM POWERED 2.
600 VAC MCC (Continued) 2EMXK-F02C Primary Bkr NV Supply to Pressurizer Viv Backup Fuse 2NV037A 2EMXK-F03A Primary Bkr S/G C Blowdown Line Sample Backup Fuse Inside Cont Isol Viv 2NM210A 2EMXK-F04A Primary Bkr S/G A Upper Shell Sample Inside Backup Fuse Cont Isol Viv 2NM187A 1
]
2EMXK-F04B 5/G A Blowdown Line Sample I
Primary Bkr Backup Fuse Inside Cont Isol Viv 2NM190A
- ~
2EMXK-F04C Primary Bkr S/G C Upper Shell Sample Backup Fuse Inside Cont Isol Viv 2NM207A 2EMXK-F06A Primary Bkr tiydrogen Skimmer Fan 2A Backup Fuse Inlet Viv 2VX1A 2EMXK-F07C
(
Primary Bkr Electric Hydrogen Recombiner l
Backup Fuse Power Supply Panel 2A j
i 2EMXK-F09A Primary Bkr Accumulator 2A Discharge Isol Backup Fuse Viv 2NI54A DEMXK-F098 i
Prirnary Bkr UHI Check Viv Test Line Inside
, sote 1 L Backup Fuse Cont Isol Viv 2NI267A 2EMXK-F09C Primary Bkr*
NC Pump 011 Fill Header Backup Fuse Cont Isol Viv 2NC196A Note 1: Upon removal of cable from power source associated with the deletion of Gil, this specification is no longer applicable l
CATAWBA - Ub?TS 1 & 2 3/4 B-47 J
1l
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]
3/4.5 EMERGENCY CORE COOLING SYSTEMS BASES
.l i
k I
3/4.5.1 ACCUMULATORS i
The OPERABILITY of each Reactor Coolant System accumulator ensures i
that a sufficient volume of borated water will be immediately forced into the
.i reactor core through each of the cold legs from the cold leg injection accumulators and directly into the reactor vessel from the upper head !njection accumulators in the. event the Reactor Coolant System pressure falls belu< the pressure of the accumulators.
This initial surge of water into the cor 3rovides the initial cooling mechanism during large pipe ruptures.
The limits on accumulator volume, boron concentration and pressure ensure that the assumptions used for accumulator injection in the safety analysis are met.
(
The accumulator power operated isolation valves are considered to be
" operating bypasses" in tha context of IEEE Std. 279-1971, which requires that i
bypasses of a protective function be removed automatically whenever permissive conditions are not met.
In addition, as these accumulator isolation valves a 1
f ail to meet single failure criteria, removal of power to the valves is require'd.
i The limits for operation with an accumulator inoperable for any reason l
except an isolation valve closed minimizes the time exposure of the plant to a LOCA event occurring concurrent with failure of an additional accumulator which may result in unacceptable peak cladding temperatures.
If a closed i
isolation valve cannot be immediately opened, the full capability of one umulator is not available and prompt action is required to place the a
INSERT Cea or in a mode where this capability is not required.
A 3/4..2 and 3/4.5.3 ECCS SUBSYSTEMS The OPERABILITY of two independent ECCS subsystems ensures that sufficient emergency core cooling capability will be available in the event of a LOCA assuming the loss of one subsystem through any singla failure consideration.
Either subsystem operating in conjunction with the accumulators is capable of supplying sufficient core cooling to limit the peak cladding temperatures within acceptable limits for all postulated break sizes ranging from the double ended break of the largest cold leg pipe downward.
In addition, each ECCS subsystem provides long-term e. ore cooling capability in the recirculation mode during the accident recovery period.
With the coolant temperature below 350*F, one OPERABLE ECCS subsystem is acceptable without single failure consideration on the ils of the stable reactivity condition of the reactor and the limited c:
.coling requirements.
CATAWBA - UNITS 1 & 2 B 3/4 5-1
R
.-i INSERT A-l The original licensing basis for the Catawb'a' Nuclear Station assumes'both the UHI System and Cold Leg Accumulators function to. mitigate postulated accidents.
1 i
Subsequent analyses have demonstrated that the UHI System is not required provided minor changes to the Cold Leg Accumulator parameters and discharge paths are.
f implemented. Accordingly, Specification 3/4.5.1 has been modified to address two i
possible plant configurations:
(1) UHI Operable (Original plant configuration)
I (2) UHI Disconnected ' (UHI penetrations to the Reactor Vessel are cut and cappea, Cold Leg Accumulator level and cover pressure changed, Cold Leg Accumulator flow restricting orifice replaced) l
~
4 i
l 1
ATTACHMEhT 2 DISCUSSION AND SAFETY ANALYSIS a
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i 1
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1 l
l i
i The proposed revision to the Catawba Units 1 and 2 Technical Specifications would allow operation with the Upper Head Injection (UHI) System removed.
The present plant design for Catawba, Units 1 and 2 incorporates an ice condenser containment and UHI System. The ice condenser containment was introduced as a less costly alternative to the large dry containment. The ice beds were designed to efficiently condense the steam from the postulated design basis large break loss-of-coolant accident (LOCA), limiting the pressure excursion in the containment. The low containment pressure contributed to increased steam binding in the primary system, thereby delaying reflood of the core and leading to higher calculated peak clad temperatures.
UHI was introduced in some Westinghouse plants to enhance core cooling during blowdown and to generally compensate for core cooling problems associated with low containment pressure. The UHI system may be visualized as a pressurized accumulator tank containing at least 1850 cubic feet of borated water which is connected through a system of shut-off and check valves to the upper head of the reactor vessel. A nitrogen gas system keeps the UHI system pressurized to over 1200 psi. When the primary system is above 1900 psi, the UH1 is isolated only by check valves. UHI plant upper head internals are designed to assure good mixing of the injected flow with the steam from the hot legs during & cold leg LOCA.
Experience has demonstrated that the UHI system adds to the complexity of plant
~
l
]
operation, requires additional maintenance and generally reduces plant availability. Licensee event reports have identified numerous problems with the* P UHI systems. Duke Power Company began discussions with the NRC early in 1985 while 1
evaluating the possibility of eliminating the UHI systems from the McGuire Station, j
Units 1 end 2.
After a series of interchanges with the staff and the ACRS, it was j
generally agreeJ that there were adequate incentives to consider removal.
The UHI removal analyses include licensing basis transient analysis of a spectrum of large and small break LOCAs, the steamline break accident, and containment response to LOCA and steamline break transients.
In order to expedite the review process and allow an easier comparison of existing analyses to the UHI removal confirmatory analyses, the details of the analytical results are provided in the form of markups of the applicable FSAR section (see Attachment 4).
ECCS ANALYSIS UHI was installed at Catawba Units 1 and 2 to increase operational flexibility through a higher total core peaking factor. The pumped safety injection / cold leg accumulator capability of Catawba was not decreased when UHI was added. Rather, Catawba, without URI, has the same installed ECCS capability as the standard RESAR-3S 4-loop Westinghouse plant. Advances in ECCS techr.clogy and experimental information gathered over the past decade permit more accurate but still conservative 10 CFR 50 Appendix K ECCS computations to be performed. Use of this advanced technclogy demonstrates that Catawba, being equipped with the standard Westinghouse ECCS, complies with the requirements of 10 CFR 50.46.
Thus, peaking factors can be obtained with UHI removed that allow full operational flexibility.
The burden of maintaining this complex equipment should be lifted to allow Duke Power to concentrate efforts on items that will improve plant safety.
Implementation of UHI removal requires new FSAR large and small break ECCS performance analyses. The completed LOCA reanalysis 4 fort includes large break cases which utilize the BASH evaluation models and small break cases which utilize the NOTRUMP evaluation model.
LARGE BREAK LOCA j
l A large break LOCA spectrum consisting of three double-ended cold-leg guillotine breaks with different discharge coefficients is presented in Attachment 4, utilizing the NRC approved BASH evaluation model.
In this model, the NRC-approved 1981 Evaluation Model versions of SATAN-V1, LOTIC, WREFLOOD and LOCTA-IV are employed. The NRC-approved BART code, which models the reflood core heat transfer, has been combined with improved reflood thermal-hydraulic models to create BASH.
Both minimum and maximum safeguards cases have been analyzed for the limiting discharge coef ficient identified.
As shown by the analytical results detailed in Attachment 4, the large break LOCA analysis performed demonstrates that Catawba will satisfy all Appendix K requirements af ter the removal of the UHI System and minor modifications to the Cold Leg Accumulators.
SMALL BREAK LOCA A spectrum of three small breaks performed with NOTRUMP is presented in Attachment 4.
The NOTRUMP computer code has been approved following extensive'NRC staff review. The results are well within all acceptance criteria limits of 10 CFR P
50.46.
NON-LOCA TRANSIENT The only FSAR non-LOCA transients for whfch the Reactor Coolant System (RCS) is predicted to depressurize to the UHI Systee actuation range (1200-1300 psia) are the steamline break events. Therefore, the analysis of all other accidents are unaffected by the proposea UHI System elimination.
The Inadvertent Opening of a Steam Generator Relief or Safety Valve (FSAR Section 15.1.4) and Steam System Piping Failure (FSAR Section 15.1.5) analyses assume actuation of the UHI System.
These accidents have been reanalyzed without UHI System actuation. The results are presented in the FSAR page markups included in j
The DNB analysis has concluded that the DNB design basis is met in j
all cases.
]
CONTAINMENT RESPONSE ANALYSIS i
The containment response has been evaluated for comparison to peak temperature (steamline break transient) and peak pressure (LOCA) criteria.
1 STEAMLINE BREAK - CONTAINMENT ANALYSIS The steamline break inside containment analysis for temperature and pressure response is presented in Section 6.2 of the Catawba FSAR. However, this analysis did not account for heat transfer to steam from the uncovered portion of the steam generator tube bundle in the calculation of mass / energy releases. The effect of this additional heat transfer is to cause superheating of the steam, resulting in an increase in the calculated containment temperature for ice condenser type l
containments.
To support the NRC Containment Systems Branch review of this issue, Westinghouse submitted Topical Report WCAP-8822-P-S1, " Mass and Energy Releases Following a Steam Line Rupture".
This report summarizes the results of the limiting cases from sensitivity studies which were performed to determine the mass / energy blowdowns to be used in the analysis for ice condenser containments.
A review of the sensitivity studies performed in support of the WCAP has determined that the RCS is predicted to depressurize to the UHI System actuation range only
'i for cases at low power levels (and therefore low decay heat levels). The peak tm enthalpies of these low power cases is well below that of the limiting cases ented in the WCAP.
Therefore, elimination of the UHI System for Catawba vill have no impact on the conclusions of the containment response steamline break analysis since UHI does not affect the calculated limiting mass / energy releases.
LOCA-CONTAINMENT ANALYSIS The removal of UHI does not impact the existing analysis since the mass / energy release model employed has conservatively neglected the UHI contribution.
i
SUMMARY
The detailed analyses performed to support the removal of the UHI System and provided via this report show adequate and acceptable performance of the ECCS and,,
that Catawba continues to satisfy all applicable safety and regulatory requirements?
with UHI removed.
BASH RELIABILITY ENHANCEMENTS Some changes were made in the BASH methodology as a result of the identification of the need to improve the methodology in some areas. As described in WCAP-10266, Revision 2, "The 1981 Version of the Westinghouse ECCS Evaluation Model Using the BASH Code, Addendum 2:
BASH Methodology Improvements and Reliability Enhancements", (transmitted to NRC via letter dated March 26, 1987 from Mr. W. J.
Johnson, Westinghouse, to Mr. James Lyons, NRC/NRR) there are three basic steps in the BASH methodology, in addition to those performed for normal LOCA analyses:
(1)
Calculate the flooding rate after bottom of core recovery (BOCREC) using the BASH code.
(2)
Smooth the resulting oscillatory flooding rate into a series of constant flooding rate intervals. This process was carried out in a subroutine in the i
BASH code, following the completion of the reflood calculation.
(3)
Calculate the hot rod peak clad temperature (PCT) using the LOCBART coda, and the smoothed flooding rate obtained from the previous two steps.
BASH performs a mass balance on the core at each time step by comparing the mass within the core volume with the net difference of the inlet and outlet flows at the core boundaries. During the calculation of some PWR cases with BASH, it was found that the mass balance was poorer than those found in the analyses presanied in WCAP-10266, Rev. 2.
This was a result of difficulties encountered by the numerics in cases where the core temperature and/or pressure were relatively low..
1 I
An enhancement to the mass balance logic was made to the BASH code. The logic addition did not change the basic modeling, but enhanced the reliability of the results by ensuring that the mass balance was within acceptable limits, consistent with other finite difference computer codes used for safety analyses. During incorporation of the mass balance enhancement, it was decided to also incorporate more accurate fluid property tables for the low pressure range to further assure accuracy in the mass balance calculation.
Modifications were also made to the flooding rate smoothing routines used in the BASH methodology to enhance the reliability of the code logic in the choice of the proper flooding rate line segments.
As described in Appendix C of WCAP-10266, Rev.
2, two or three line segments were to be calculated from the oscillatory BASH core flooding rate for use in the calculation of the PCT by the LOCBART code. While the routines were generally quite successful, in some unusual instances, the BASH code would choose inappropriate points requiring intervention by the analyst who followed written guidelines 7 determine the flooding rate information.
Improvene..ts were made to the flooding rate smoothing routines to enhance the reliability over the full range of expected transient conditions.
The following Table presents the results of sensitivity studies performed to assess the effect of the mass balance enhancement for reliability improvement. The impact on PCT varies somewhat depending on specific plant characteristics. Consequently, it was concluded that calculations with the previous BASH model where the mass balance difference did not exceed '150 lb. during the initial flooding rate surge,
- would not be affected by the BASH modifications.
l !
3 t'
BASH SENSITIVITY STUDIES-' SHOWING THE-EFFECT OF THE MASS BALANCE MODIFICATId'l CORE MASS CHAF GE. IN CASE BALANCE PCT
-1CT:
(LBM)-
( F)
.( F) 1 4200 1579.
la 60-1603
+24 2
900' 1765 2a 130 1754
-11.
3 450-1823 3a 100 1871
+48.
4 230 2004 s., ~
4a 61 2014
+10 1
Cases la, 2a, 3a, and 4a contain BASH methodology reliability enhancements.
-]
2 The mass balance is determine by integrating inlet and outlet core ' flow and i
-i comparing with calculated core mass:
)
U 2 Mass Balance Difference =
(W - W )dt - [M(t ) '- M(t )]
g 2
y t g Where W = Core Inlet Flowrate g
W,= Core Outlet Flowrate M (t ) = Core Mass at time t g 2 2
M (t )' = Core Mass at time t g y g
)
1 n:
CONCLUSIONS The BASH methodology modifications represent improvements to currently approved models.. The results of recalculations have shown that the modifications overall have had a small impact, and do not significantly. alter the degree of conservatism
-inherent in the BASH methodology. The revised analysis results are expected to remain below the' acceptance egiteria of 2200 F with no alterations in the total number core peaking factor F 9
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~6-
i ATTACHMDIT 3 i
SIGNIFICANT HAZARDS CONSIDERATION ANALYSIS I
i e
4, -;-
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'i ANALYSIS OF NO SIGNIFICANT HAZARDS CONSIDERATIONS 0
I 10 CFR 50.92 states that a proposed amendment involves no significant hazards considerations if-operation in accordance with the proposed amendment would not:
(1)
Involve a significant increase in the probability or consequences of an accident previously eraluated; or (2) Create the possibility of a new or different kind of accident from any accident previously evaluated; or (3)
Involve a significant reduction in a margin of safety.
The proposed amendment does not involve an increase in the probability or consequences of any previously evaluated accident. Deletion of the URI System has been addressed for impact on all applicable Design Basis Accidents presented in FSAR Chapters 6 and 15.
All applicable acceptance criteria will continue to be l
met.
The results of the reanalysis are presented as marked-up FSAR pages in l
The proposed amendment will not create the possibility of a new or different kind
of accident from any accident previously evaluated. The changes to the design agd,~
operation of the plant have been evaluated.
Upon approval and implementation of
?
the proposed amendment, no unanalyzed conditions will exist as a result of this amendment.
The proposed amendment does not involve a significant reduction in a margin of safety. All applicable DBAs have been addressed with' acceptable results obtained.
These changes have already been reviewed and approved by the NRC Staff for the McGuire Nuclear Station, Units 1 and 2 (see letter dated May '13,1986 from Mr. Darl Hood, NRC/NRR, to Mr. H. B. Tucker transmitting amendments 57/38 to Facility l
Operating Licenses NPF-9 and NPF-17).
k For the above reasons, Duke Power concludes that this proposed amendment request does not involve Significant Hazards Considerations.
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ATTACHMENT 4 PROPOSED FSAR REVISIONS l
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1 1
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CNS 1l b.
Steam Plant Pipina Blowdown The contributtori to the mass and energy releases from the secondary plant steam piping is included in the mass and energy release calculations.
For all ruptures, the steam piping volume blowdown begins at the time of the break and continues at a uniform rate until the entire piping inventory is released. The flow rate is determined using the Moody correlation, the pipe cross-sectional area, and the initial steam pressure.
Following the piping blowdown, reverse flow from the intact steam generators continues until the times presented in Table 6.2.1-68 to simulate the reverse steam generator flow prior to steamline isolation.
The blowdown model is further discussed in References 18 and 19.
6.2.1.4.3 Single Failure Analysis The following single failures were evaluated to determine the limiting set of conditions for this analysis:
(
l Failure of a main steam isolation valve increases the volume of a.
steam piping which is not isolated from the break. When all valves operate, the piping volume capable of blowing down is located between the steam generator and the first isolation valve.
If this.?
valve fails, the volume between the break and the isolation valves in the other steamlines including safety and relief valve headers and other connecting lines will feed the break.
b.
Failure of a diesel generator would result in the loss of one containment safeguards train resulting in minimum heat removal capability.
Failure of a feedwater isolation valve could only result in c.
additional inventory in the feedwater line which would not be a
I isolated from the steam generator. The mass in this volume can flash into the steam generator and exit through the break.
Both the feedwater isolation valve and the feedwater regulating valve close in no more than 5 seconds, precluding the pumping of any additional feedwater into the steam generator.
The additional line volume available to flash into the steam generator is that between the feedwater isolation valve and the feedwater regulating valve, including all headers and connecting lines.
The resultant mass and energy release rates for the limiting steam pipe break are presented in Table 6.2.1-61.
]
6.2.1.5 Minimum Containment Pressure Analysis for Perfomance Capability Studies of Emergency Core Cooling System The containment pressure analysis is performed with the LOTIC-2 code (Reference 3). The transient pressure computed by the LOTIC-2 code can be input to the.WREftee9 code for the purpose of computing the reflood transient.
.)'
BASH 6.2-36 Rev. 14 I
CNS L
The containment backpressure used for the 'f..it' ; rr C - C' '
^A'""
D
'.. it ECCS analysis presented in 'Section 15.6.5 is presented in Figure 6.2.1-76.
The containment backpressure is calculated using the methods and assumptions described in Reference 3.
Input parameters, including the containment initial y
J conditions; containment volume; passive sink materials, thicknesses, and surface areas; and starting time and number of containment cooling systems used in the analysis, are described below.
6.2.1.5.1 Mass and Energy Release Data The mass / energy releases to the containment during the blowdown and reflood The lportionsofthelimitingbreaktransientarepresentedinTable6.2.1-62.
mass and energy releases from the broken loop accumulator are given in Table 6.2.1-64.
The mathematical models which calculate the mass and energy releases to the con-l tainment are described in Section 15.6.5.
Since the requirements of Appendix K of 10 CFR 50 are very specific in regard to the modeling of the RCS during, blow-l down and the models used are in conformance with Appendix K, no alterations tir those models have been made in regard to the mass and energy releases.
A break spectrum analysis is performed (see references in 5ection 15.6.5) that considers various break sizes, break locations, and Moody discharge coefficients for the double-ended cold leg guil10 tines, which do affect the mass and energy released to the containment.
This effect is considered for each case analyzed. During
(
reflood, the effect of steam-water mixing between the safety injection water and the steam flowing through the RCS intact loops reduces the available energy re-s leased to the containment vapor space and therefore tends to minimize contain-ment pressure.
6.2.1.5.2 Initial Containment Internal Conditions Containment data and initial conditions used in the analysis are presented in Table 6.2.1-65.
j 6.2.1.5.3 Containment Volume 1,19Q498 The volume used in the analysis is 2,T,'M ft8 1
6.2.1.5.4 Active Heat Sinks The containment spray system and the containment atmosphere ice condenser sys-tem with the air recirculation system operate to remove heat from the Containment.
Pertinent data for these systems, which were used in the analyses,'are presented in Table 6.2.1-66.
The sump temperature was not used in the analysis because the maximum peak clad-l ding temperature occurs prior to initiation of the recirculation phase for the containment spray system.
In addition, heat transfer between the sump water and the containment vapor space was not considered in the analysis.
(
6.2-37 Rev. 13 l
)
6.2.1.5.5 Steam-Water Mixing Water spillage rates from the broken loop accumulator are determined as part of the core reflooding calculation and are included in the containment LOTIC code calculation model.
3 6.2.1.5.6 Passive Heat Sinks The passive heat sinks used in the analysis and their thermophysical properties are given in Table 6.2.1-67.
6.2.1.5.7 Heat Transfer to Passive Heat Sinks cbe o.lo, m w m s~ f* swr ds, i
The containment pressure transient for the -;.;t ;;.;; CD - 0 !, 7.cf;;t Q is presented in Figure 6.2.1-76. " i; -........'...
u,..
A L. a,,..;; r; ' n
'.d;;d ;;.......... fe. ;... u.= e > op p.. W.
Figure 6.2.1-77 shows the heat re-moval rates of the lower compartment heat sinks; the heat transfer model used 1
-is described in Reference 3.
Figure 6.2.1-78 gives the flowrate exiting the ice condenser drains Figure 6.2.1-79 presents the temperature transients in both the upper and lower compartments.
Figures 6.2.1-80 and 6.2.1-81 illustrate the heat removal rates by the sump and the ice condenser drains. The total heat removal rate in the lower compartment is the summation of the rates given in Figures 6.2.1-77, 6.2.1-80 and 6.2.1-81.
)
6.2.1.5.8 Other Parameters All parameters having a substantial effect on the minimum containment pressure analysis have been discussed or referenced in the preceding sections.
6.2.1.6 Testing and Inspection Preoperational and periodic inservice tests are conducted to ensure the func-tional capability of the Containment and associated systems.
The tests are discussed in the following Sections:
Section Test 14.2 Ice condenser bypass leakage 14.2, 16.0 Containment Air Return System j
l 16.0 Ice condenser drains 14.2, 16.0 Ice condenser testing 5.4.7, 14.2, 16.0 Residual Heat Removal System 6.2.2.4, 16.0 Containment Spray System 6.2.6, 16.0 Containment leakage testing
)I 6.2-38 Rev. 13
I MOW 4. A, / ~ (, *A D e It44, ad
$er k t e. M'n new a t N A s/
1 Table 6.2.
(Page 1) l l
DECL Blowdown Mass and Energy lease CD = 0.6 MASS FL ENERGY FLO (SEC (LB/SEC)
(BTU /SEC)
.1024E+05 5637E+07 l
O.
.2000E+01
.5290E+05 E+08 4000E+01
.3308E+05
.1895
.6001E+01
.2942E+05
.1753E+08
.8001E+01
.2518E+05
.1546E+08
.1000E+02 19E+05
.1320E+08
.1200E+02
.1
+05
.1068E+08
.1240E+02
.1613E
.1068E+08 1400E+02
.1219E+05
.8576E+07 OE+02
.8654E+04
.6304E+07
.180 +02
.5898E+04
.4222E+07
.2000E+h
.4230E+04
.2701E+07 i
4
.2200E+02
.3615E+04 1983E+07
.2400E+02
.3247E+04 1E+07
.2500E+02
.4002E+04
.162 7
.2600E+02
.3903E+04
.1557E+
.2700E+02
.3986E+04
.1547E+07
.2800E+02
.4453E+04
.1648E+07
.2900E+02
.4693E+04
.1588E+07 1
.3000E+02 4721E+04
.1438E+07 1
.3100E+02 4E+04
.1307E+07
.3600E+02
.3862Eg04
.9802E+06
.4100E+02
.8898E+
.1314E+06
.5100E+02
.8615E+03
.1551E+06 600E+02
.3673E+04
.5950E+06 E+02
.8099E+03
.1642E+06
.3996E+04
.5943E+06
.7100 2
.7600E+0
.4044E+04
.6016E+06
.8600E+02
.6236E+04 678E+06
,9100E+02
.3637E+04 3E+06
.1010E+03
.1033E+04
.219 06
.1060E+03
.3004E+03
.1443E+
e C
Rev. 13 l
h e,\\ d o A M hep l4C 4 Table C.2.1-62 (Pa 2)
)
DECL Ref1 d Mass and Energy Re1 e
0 = 0.6 MASS FL ENERGY FL TIME (SEC)
(LB/SEC)
(BTU /SEC) 80E+03
.1848E+03
.7022E+04
.1 4E+03
.2259E+03
.2443E+05
.10
+03
.1946E+03
.1965E+05
.1089 03 006E+03
.2742E+05
.1093E+
2E+03 1125E+06
,i
.1127E+03
.253 +03 31E+05
.1181E+03
.2506E+
.86 E+05
.1246E+03
.2507E+03
.8688 5
.1340E+03
.2526E+03
.8911E+
.1464E+03
.4349E+03
.1416E+06
.1747E+03
.5377E+03
.1617E+06
.2101E+03
.5680E+03
.1568E+06
.2525E+03
.5812E+03
.1460E+06 i
I l
1 i
b, Rev. 13 New Page
l Table 6.2.1-62 (Page 1)
DECL Blowdown Mass. and Enercrv Release CD = 0.6 i
TIME MASS FLOW ENERGY FLOW (SEC)
(LB/SEC)
(BTU /SEC) 0.050 6.3778E+04 3.5544E+07 1.001 6.3043E+04 3.5324E+07 2.001 5.4561E+04 3.1079E+07 3.001 4.2733E+04 2.4733E+07 4.001 3.5300E+04 2.0890E+07 5;004 3.0936E+04-1.8667E+07 6.002 2.8986E+04 1.7837E+07 7.007 2.5565E+04 1.6561E+07 8.002 2.2103E+04 1.4962E+07 P
i 9.001 2.0792E+04 1.3805E+07 10.002 1.9714E+04 1.2762E+07 3
11.004 1.8151E+04 1.1737E+07 12.005 1.6182E+04 1.0666E+07 12.505 1.5256E+04 1.0188E+07 13.005 1.4361E+04 9.7367E+06 14.006 1.2515E+04 8.8412E+06 15.005 1.0101E+04 7.6244E+06 16.006 8.2147E+03 6.5592E+06-17.001 6.7195E+03 5.6640E+06 18.003 5.5874E+03 4.9654E+06 19.004 4.4684E+03 4.2132E+06 20.001 3.0045E+03 2.8970E+06 21.007 7.2176E+03 3.8160E+06 22.000 7.4753E+03 3.5673E+06 4
23.003 7.3579E+03 3.1500E+06 24.001 6.9598E+03 2.7060E+06 25.002 6.6299E+03 2.4345E+06 26.003 5.6767E+03 1.9287E+06 27.001 4.6297E+03 1.4368E+06 28.003 3.6508E+03 1.0290E+06 29.001 3.2045E+03 8.0846E+05 30.004 3.0164E+03 6.9198E+05 i
31.001 3.8384E+02 1.0710E+05 31.730 1.5287E+03 3.0996E+05 I
.,e
..,a,-..
1[
I 1
'l Table 6.2.1-62 (Page 2)
DECL Reflood Mass and Enerov Release CD = 0.6 TIME MASS FLOW-ENERGY FLOW (SEC)
(LB/SEC)
(BTU /SEC) 44.78 0.0
' 0. 0 i
45.35 3.250 4.2270E+03 46.10 4.402 5.7253E+03 f
63.22 2.749E+02 2.3178E+05 4
81.82 2.968E+02 2.3462E+05 102.92 3.0690+02 2.3239E+05 149.72 3.196E+02 2.2454E+05 l
j 165.07 3.601E+02 2.3739E+05 a 194.62 3.787E+02 2.4142E+05 250.62 3.901E+02 2.4163E+05 i
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TABL 2.1-64
(
Broken Loop Accumulator Flow Containment TIME MASS (sec)
OW RATE *
(
see) 0.0 0.0 3
0.02 235.0
.00 1235.4 3.
1146.4 l
- 5. 0 1074.5 1015.6 7.02 g
10.02 943.4 15.02 848.9 20.02 779.7 25.02 725.2 35.02 647.6 45.02 595.0 55.02 0.7 3'
65.02 5
8 75.00 481.
.00 455.2
- 95. 0 481.8
(
96.69 429.6 10000.h 97.0 0.0 0.0 0 %\\%
- Enthalpy of acc mulator water is 38.0 BTU /LBM
TABLE 6.2.1-64 Broken Looo Accumulator Flow to Containment TIME MASS (seei FLOW RATE
- fib /sec) 0.0 0.0.
0.02 358.8 1.02 3531.8 3.02 2880.6 5.02 2490.3 7.02 2218.6 9.02 2012.5 11.02 1838.2 13.02 1676.1 15.02 1532.9 17.02 1632.3 19.02 1632.9 21.02 1560.6 23.02 1496.3 25.02 1438.7 27.02 1391.9 29.02 1350.8 31.02 1314.4-33.02 1282.2 1
35.02 1253.2 36.28 1239.6 36.30 0.0 300.00 0.0
- Enthalpy of accumulttor water is 59.6 BTU /LBM i
I 3
TABLE 6. 2.1-65 Active Heat Sink Data
,ff For Minimum Post-LOCA Containment Pressure I
Containment Spray System Parameters A.
Maximum spray system flow, total 9600 gpm B.
Fastest post-LOCA initiation of spray system Ass'sning off-site power loss at start j)O of LOCA
,7&"sec 11 Containment Air Recirculation System Fan A.
Maximum number of fan coolers operating d
B.
Fastest post-LOCA initiation 600 sec
]
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TABLE 6.2.1-66 (Pace 1 of 3)
STRUCTURAL HEAT SINKS Thickness and Material Arag)
.(ft)
(ft A. Uccer comeartmen'
- 1. Operating Floor, Crane Wall, Refueling Canal, Miscellaneous concrete Slab 1 21142 0.000833 Coating 2 1.34 Concrete Slab 2 5017 0.0156 Stainless Steel-1.5 Concrete
- 2. Containment Vessel Dome, Containment Shell, Polar Crane, Miscellaneous Steel Slab 3 24391 0.00059 Coating 1 F
0.058 Carbon Steel Slab 4 31035 0.00059 Coating 1 0.0290 Carbon Steel Slao 5 801 0.0625 Stainless Steel B.
Lower and Dead Ended Compartments 1.
Operating Floor, Crane 3
1 Wall, Refueling Canal, Miscellaneous Concrete Slab 1 57387 0.000833 Coating 2 1.97 Concrete Slab 2 9019 0.00133 Coating 3 2.04 Concrete f
Slab 1 3541 0.00133 Coating 3 2.50 Concrete Slab 4 2361 0.0156 Stainless Steel 1.50 Concrete slab 5 768 0.00059 Coating 1 0.04207 Carbon Steel.
1.50 Concrete l
i TABLE 6.2.1-66 (Pace 2 of 3)
STRUCTURAL HEAT SINKS Thickness and Material Areg),
(ft (ft)
- 2. Containment Shell, Reactor Coolant Pumps, Supports, and Miscellaneous Steel Slab 6 56551 0.00059 Coating 1 0.0535 Carbon Steel Slab 7 14445 0.00059 Coating 1 0.0625 Carbon Steel Slab 8 9040 0.00059 Coating 1 0.0625 Carbon Steel Slab 9 32640 0.0026 Stainless Steel
- 3. Cooling Coils 4
y Slab 10 51000 0.00042 Copper C.
Ice Condenser 1.
Ice Baskets Slab 1 180628 0.00663 Steel
- 2. Lattice Frames Slab 2 76650 0.0217 Steel 3.
Lower Support Structure Slab 3 28670 0.0267 Steel
- 4. Ice Condenser Floor Slab 4 3336 0.000833 Coating 0.333 Concrete
- 5. Containment Wall Panels and Containment Shell Eigh_}
19100 1.0 Steel and Insulation 0.0625 Steel Shell
i l
1 1
i TABLE 6.2.1-66 (Pace 3 of'3)
STRUCTURAL HEAT SINKS
{
Thickness and Mater.21 Arag)
(ft)
(ft l
- 6. Crane Wall Panels and Crane Wall Slab 6 13055 1.0 Steel and Insulation 1.0 Concrete l
l 0
Coatings (Btu /ft hr F) l
- 1) 2 mils organic, 5 mils inorganic - 0.6'
- 2) 10 mils organic - 0.29
- 3) 16 mils organic - 0.29 Volumetric Heat Capacity (Btu /ft3 _ oy) f
~
Concrete 31.95 Carbon Steel 58.8
- l-Stainless Steel 55.11 f
Inorganic Coating 28.8 Organic Coating 18.2
]
I l
i TABLE 6.2.1-67 Containment Data Required for ECCS Evaluation of Ice Condenser Containment
,e 3
I.
Containment Volume In FT 670.,/01 Upper Compartment
,.wv.
$~13 M !
Ol,,'?.
Lower Compartment Ice Condenser (Includes, lower plenum, 181,400.
upper plenum, and ice bed cooling ducts) p t,7 M 2 2 ', 0 '.
Dead Ended Compartments 3
1,16c, ?'.
FT II.
Initial Conditions A.
....t-Operation Containment Pressure 14.7 psia I
B.
"';t ;t-Operational Containment Temperature for 100 F UC the upper, lower, and dead ended compartments
/af 43 F LC
_ L: :. Refueling Water Storage Tank Temperature 70 F D.
L c '-Service Water Temperature 32'F E.
.,. Temperature Outside Containment 10 F j
F. - _ n Initial Spray Temperature 70~F G.
Ice Condenser Temperature 2 7 + % S,f l
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(
LIST OF FIGURES (Continued)
L 1
Ficure No.
Title 4
15.4.7-5.
Loading a Region 2 Assembly'into a Region 1 Position Near Core -
j Periphery 15.4.8-1 Nuclear Power Transient, BOL HFP, Rod Ejection Accident 15.4.8-2
. Hot Spot Fuel and Clad Temperature Versus Time,.BOL HFP, Rod Ejection Accident 15.4.8-3 Nuclear Power Transient, BOL HZP, Rod Ejection Accident.
15.4.8-4 Hot Spot Fuel and Clad Temperatures Versus Time, BOL HZP and Rod Ejection Accident 15.4.8-5 Integrated Break Flow,.RCCA Ejection l
15.5.1-1 Inadvertent Actuation of ECCS During Power Operation 15.5.1-2 Inadvertent Operation of ECCS During' Power Operation 15.5.1-3 Inadvertent Operation of ECCS Ouring Power Operation h
15.6.1-1 Inadvertent Ooening of a Pressurizer Safety Valve 15.6.1-2 Inadvertent Opening of a Pressurizer Safety Valve 15.6.3-1 Break Flow and Injection Flow 15.6.3-2 Reactor Coolant System Pressure 15.6.3-3 Reactor Coolant System Temperatures 15.6.3-4 Steam Generator Pressure 15.6.3-5 Steam Generator Temperature 15.6.3-6 Pressurizer Water Volume 15.6.3-7 Steam Generator Flow 15.6.5-1 Sequence of Events for Large Break Loss-of-Coolant Analysis 15.6.5-2 Code Interface Description for Large Break Model 15.6.5-3 Code Interface Description for Small Break Model 0:0LC (Cf.0 L F;rfect "' * ;
c_o 1.o DCCt G l
15.6.5-4 Fluid Quality, 15.6.5-5 Fluid Qualityp OECLC (CD -
' I2;2 ""I H CCD = c.3 DE C.L G 15-xvii
li i
i bele ed R*e la c4-l LIST OF FIGURES (Continued) l Ficure o.
Title i
15.6.5-6 Fluid Quality - DECLG (C =0 8), Perfect Mixing D
15.6.5-7 luid Quality - DECLG (C =0.6), Perfect Mixing D
15.6.5-8 F1 d Quality - DECLG (C =0.4), Perfect' Mixing 0
15.6.5-9 Mass locity - DECLG (C =1.0), Perfect Mixing D
15.6.5-10 Mass Vel city - DECLG (C =1.0), Imperfect Mixing 0
15.6.5-11 Mass veloc y - DECLG (C =0.8), Perfect Mixing D
15.6.5-12 Mass Velocity - DECLG (C =0.6), Perfect Mixing 0
15.6.5-13 Mass Velocity - eCLG (C =0.4), Perfect Mixing D
e 15.6.5-14 Heat Transfer coeff ient - DECLG (C =1.0), Perfect Mixing 0
15.6.5"15 Heat Transfer Coefficie
- DECLG (C =1.0), Imperfect Mixing 0
15.G.5-16 Heat Transfer Coefficient DECLG (C =0.8), Perfect Mixing D
15.6.5-17 Heat Transfer Coefficient - 0 LG (C =. ),
erfect Mixing D
15.6.5-18 Heat Transfer Coefficient - DECL (C =0.4), Perfect Mixing D
15.6.5-19 Core Pressure - DECLG (C =1.0), Per et Mixing 0
15.6.5-20 Core Pressure - DECLG (C =1.0), Imperfe t Mixing j
D 15.6.5-21 Core Pressure - DECLG (C =0.8), Perfect M.ing j
0 i
15.6.5-22 Core Pressure - DECLG (C =0.6), Perfect Mixi j
0 15.6.5-23 Core Pressure - DECLG (C =0.4), Perfect Mixing 0
15.6.5-24 Flowrate in Lower Half of Core, C =1.0 DECLG, Per et Mixing, D
(Satan Nodes)
)
15.6.5-25 Flowrate in Lower Half of Core, C =1.0 DECLG, Imperfe'
- Mixing, D
(Satan Nodes) 15.6.5-26 Flowrate in Lower Half of Core, 0 :0.8 DECLG, Perfect Mixi 0
(Satan Nodes) l l
15.6.5-27 Flowrate in Lower Half of Core, C =0.6 DECLG, Perfect Mixing,
.Ij D
(Satan Nodes) s i
15-xviii
)
i
{
I.IST OF FIGURES (Continued)
Ficrure No.
Title f
15.6.5-6 Fluid Quality, C =0.5 DECI4 D
15.6.5-7 Fluid Quality, C =0.8 DEC14 (Max. SI) p 15.6.5-8 Mass velocity, C =1.0 DECIA p
15.6.5-9 Hass Velocity, C =0.8 DECI4
]
p 15.6.5-10 Mass Velocity, C =0.6 DECLG D
15.6.5-11 Mass Velocity, C =0.8 DECLG (Max. SI) p 15.6.5-12 Heat Transfer Coefficient, C =1.0 DECI4 p
]
15.6.5-13 Heat Transfer Coefficient, C =0.8 DECI4 a
D p
f 15.6.5-14 Heat Transfer Coefficient, C =0.6 DECLG
{
D
(
15.6.5-15 Heat Transfer Coefficient, C =0.8 DECLG (Max. SI) p 15.6.5-16 Core Pressure, C =1.0 DECI4 q
p l
15.6.5-17 Core Pressure, C =0.8 DECLG D
15.6.5-18 Core Pressure, C =0.6 DECLG D
15.6.5-19 Core Flowrate, C =1.0 DECLG D
15.6.5-20 Core Flowrate, C =0.8 DECLG D
15.6.5 21 Core Flowrate, C =0.6 DECLG p
15.6.5-22 Peak Clad Temperature, C =1.0 DECLG D
15.6.5-23. Peak Clad Temperature, C =0.8 DECI4 p
15.6.5-24 Peak Clad Temperature, C =0.6 DECLG p
15.6.5-25 Peak Clad Temperature, C =0.8 DECI4 (M.ax. SI) p 15.6.5-26 Fluid Temperature, C =1.0 DECLG p
15.6.5-27 Fluid Temperature, C =0.8 DECLG p
q l
R
- e a c. o Delete and LIST OF FIGURES (Continued)
Title ure No.
Flowrate in Lower Hal' of Core, C =0.4 DECLG, Perfect Mixing,.
f 15.6.5 8 D
f (Satan Nodes) 15.6.5-29 Flowrate in Upper Half of Core, C =1.0 DECLG, Perfect Mixing, D
(Satan Nodes) 15.6.5-30 wrate in Upper Half of Core, C =1.0 DECLG, Imperfect Mixing, D
(S an Nodes).
15.6.5-31 Flowr e in Upper Half of Core, C =0.8 DECLG, Perfect Mixing, D
(Satan des) 15.6.5-32 Flowrate i Upper Half of Core, C =0.6 DECLG, Perfect Mixing, D
(Satan Node 15.6.5-33 Flowrate in Up r Half of Core, C =0.4 DECLO, Perfect Mixing, -
D (Satan Nodes) a 15.6.5-34 Void Fraction in Lo r Half of Core, C =1.0 DECLG, Perfect Mixirig, D
(Satan Nodes) 15.6.5-35 Void Fraction in Lower 1f of Core C =1.0 DECLG, Imperfect Mix-D ing, (Satan Nodes) 15.6.5-36 Void Fraction in Lower Half Core, C =0.8 DECLG, Perfect Mixing, D
(Satan Nodes) 15.6.5-37 Void Fraction in Lower Half of Co
, C =0.6.DECLG, Perfect D
Mixing, (Satan Nodes) 15.6.5-38 Void Fraction in Lower Half of Core, C 0.4 DECLG, Perfect D
Mixing, (Satan Nodes) 15.6.5-39 Void Fraction in Upper Half of Core, C =1.0 ECLG, Perfect D
Mixing, (Satan Nodes) 15.6.5-40 Void Fraction in Upper Half of Core, C =1.0 DECL Imperfect D
Mixing, (Satan Nodes) 15.6.5-41 Void Fraction in Upper Half of Core, C =0.8 DECLG, Per e t' D
Mixing, (Satan Nodes) 1 15.6.5-42 Void Fraction in Upper Half of Core, C =0.6 DECLG, Perfect ixing, D
(Satan Nodes) 15.6.5-43 Void Fraction in Upper Half of Core, C =0.4 DECLG, Perfect Mixi D
(Satan Nodes) 15.6.5-44 Peak Clad Temperature-Nodes 6 and 13, C =1.0 DECLG, Perfect Mixing D
15-xix
p LIST OF FIGURES (Continued)
Fiaura No.
Title 15.6.5-28 Fluid Temperature, Cp=0.6 DECLG-C =0.8 DECLG (Max. SI) 15.6.5-29 Fluid Tamparature, p
15.6.5-30 Reflood Mixture Levels, C =1.0 DECLG p
15.6.5-31 Reflood Mixture Invals, C =0.8 DECLG p
15.6.5-32 Reflood Mixture Levels, C =0.6 DECLG p
15.6.5-33 Reflood Mixture Levels; C =0.8 DECLG (Max. SI)
D 15.6.5-34 Core Inlet Velocity, C =1.0 DECLG D
15.6.5-35 Core Inlet Velocity, C =0.8 DECLG D
15.6.5-36 Core Inlet Velocity, C =0.6 DECLG 3
D 15.6.5-37 Core Inlet Velocity, C =0.8 DECLG (Max. SI)
D 15.6.5-38 Accumulator Injection, C =1.0 DECLG D
15.6.5-39 Accumulator Injection, C =0.8 DECLG D
15.6.5-40 Accumulator Injection, C =0.6 DECLG D
15.6.5-41 Compartment Pressure, Minimum SI 1F.6.5-42 Compartment Pressure, Maximum SI 15.6.5-43 Small Break Power Distribution Assumed for LOCA Analyses 15.6.5-44 Core Power after Reactor Trip (Applies to all Small Breaks) 15.6.5-45 Safety Injection Flowrate vs. Pressure 15.6.5-46 3" Cold Leg Break RCS Pressure vs. Time 15.6.5-47 3" Cold Leg Break Core Mixture Height vs. Time 15.6.5-48 3" Cold Leg Break Hot Spot Clad Temperature vs. Time 15.6.5-49 4" Cold Leg Break RCS Pressure vs. Time
D., i +e maal Rep I = c.e LIST OF FIGURES (Continued)
.)
Figure o.
Title 15.6.5-45 Peak Clad Temperature-Nodes 9 and 14, C =1.0 DECLG, Imperfect D
Mixing 15.6.5-46 k Clad Temperature-Nodes 6 and 7, C =0.8 DECLG, Perfect Mixing D
15.6.5-47 Peak lad Temperature-Nodes 7 and 14, C =0.6 DECLG, Perfect Mixing D
~
15.6.5-48 Peak Cia Temperature-Nodes 8 and 7, C =0.4 DECLG, Perfect Mixing D
15.6.5-49 Fluit. Tempe ture-Nodes 13 and 6, C =1.0 DECLG, Perfect Mixing D
15.6.5-50 Fluid Temperat e-Nodes 9 and 14, C =1.0 DECLG, Imperfect Mixing D
~
15.6.5-51 Fluid Temperature odes 6 and 7, CD = 0.8 DECLG, Perfect Mixing s
15.6.5-52 Fluid Temperature-No s 7 and 14, C =0.6 DECLG, Perfect Mixing C
D 15.6.5-53 Fluid Temperature-Noce 8 and 7, C =0.4 DECLG, Perfect Mixing D
15.6.5 54 Reflood Transient, C =1.0 ECLG, Perfect Mixing D
15.6.5-55 Reflood Transient, C =1.0 DE G Imperfect Mixing D
15.6.5-56 Reflood Transient, C =0.8 DECL Perfect Mixing D
15.6.5-57 Reflood Transient, C =0.6 DECLG, efect Mixing D
15.6.5-5B Reflood Transient, C =0.4 DECLG, Per et Mixing D
15 6.5-59 Reflooding Rate, C =1.0 DECLG, Perfect
'xing D
IL 6.5-60 Reflooding Rate, C =1.0 DECLG, Imperfect M ing 0
15.6.5-61 Reflooding Rate, C =0.8 DECLG, Perfect Mixing D
15.G.5-62 Reflooding Rate, C =0.6 DECLG, Perfect Mixing D
15.o.5-63 Reflooding Rate, C =0.4 DECLG, Perfect Mixing D
15.6.5-64 Accumulator Flowratu, C =1.0 DECLG, Perfect Mixing D
15.6.5-65 Accumulator Flowrates, C =1.0 DECLG, Imperfect Mixir.2 D
I 15.6.5-56 Accumulator Flowrates, C =0.8 DECLG, Perfect Mixing D
15.6.5-67 Accumulator Flowrates, C =0.6 DECLG, Perfect Mixing
]
D 15.6.5-68 Accumulator Flowrates, C =0.4 DECLG, Perfect Mixing D
9Bann
l-9; I
i-
\\
\\
(
.{
j s
i i
IJST OF TIGURES (Continued)
Fierure No.
- Title e
8 15.6.5-50 4" Cold Leg Break Core Mixture Height vs. Time 15.6.5-51 4" Cold Leg Break Hot Spot Clad.TempArature vs. Time 15.6.5-52 6" Cold Leg Break RCS Pressure vs. Time 15.6.5-53 6" Cold Leg Break Core Mixture Height vs. Time i
15.6.5-54 6" Cold Leg Break' Hot Spot Clad Temperature vs. T
15.6.5-55 4"ColdLehBreakCoreSteamFlowvs. Time l
l 15.6.5-56 4*'Celd Leg. Break Core Heat Transfer coeffecient vs. Time l
16.6.5-57
.4" Cold tao streak Hot' Spot Fluid Camperature vs. Time j
l j
e i
f f
b LIST OF FIGURES (Continued)
F ure No.
Title l
15.6.5 9 SI + Accumulator Flow, C =1.0 DECLG, Perfect Mixing D
15.6.5-7 SI + Accumulator Flow, C =1.3 DECLG, Imperfect Mixing D
15.6.5-71 SI + Accumulator Flow, C
.41 CECLG, Perfect Mixing D
15.6.5-72 SI + Accumulator Flow, C =0.6 DECLG, Perfect Mixing D
15.6.5-73 5 + Accumulator Flow, C =0.4 DECLG, Perfect Mixing D
15.6.5-74 Ves 1 Mass Inventory, C =1.0 DERG, Perfect M g D
15.6.5-75 Vesse Mass Inventory, C =1.0 DECLG, Imperfect M xing D
15.6.5-76 Vessel ss Inventory, C =0.8 DECLG, Perfect Mixing D
15.6.5-77 Vessel. Mas Inventory, C =0.6 DECLG, Perfect Mixing D
s a.
15.6.5-78 Vessel Mass ventory, C =0.4 DECLG, Perfect Mixing D
15.6.5-79 Small Break Po r Distribution Assumed for LOCA Analyses C.
15.6.5-80 RCS Pressure, 8 i h Small Break 15.6.5-81 Core Mixture Height, inch Small Break 15.6.5-82 Hot Spot Clad Temperat e, 8 inch Small Break 15.6.5 83 Core Power After Reactor ip (Applies to all Small Breaks) 15.6.5-84 Core Steam Flowrate, 8 inch ell Break 15.6.5-85 Core Heat Transfer Coefficient, inch Small Break 15.6.5-86 Hot Spot Fluid Temperature, 8 inch all Break 15.6.5-87 RCS Pressure, 6 inch Small Break 15.6.5-88 RCS Pressure, 4 inch Small Break 15.6.5-89 Core Mixture Height, 6 inen Small Break 15.6.5-90 Core Mixture Height, 4 inch Small Break 15.6.5-91 Hot Spot Clad Temperature, 6 inch Small Break 15.6.5-97 Hot Spot Clad Temperature, 6 inch Small Break 15-xxi
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MM 15.6.5 LOSS-0F-COOLANT ACCIDENTS 15.6.5.1 Jdentification of Causes and Frecuency Classification e
A loss-of-coolant accidetit (LOCA) is the result of a pipe rupture of the reactor coolant system (RCS) pressure boundary.
For the analyses reported here, a major pipe break (large break) is defined as a rupture with a total cross sectional area equal to or greater than 1.0 ftz. This event is con-sidered an ANS Condition IV event, a limiting fault, in that it is not ex-pected to occur during the lifetime of the plant but is postulated as a conservative design basis.
See Section 15.0.2.
A minor pipe break (small break), as considered in this section, is defined as a rupture of the reactor coolant pressure boundary (Section 5.2) with a i
2 in which the normally operating total cross-sectional area less than 1.0 ft charging system flow is not sufficient to sustain pressurizer level and pres-This is considered a Condition III event in that it is an infrequent sure.
fault which may occur during the life of the plant.
The Acceptance Criteria for the loss-of-coolant accident is described in 10 f
CFR 50.46 as follows:
j The calculated peak fuel element cladding temperature is below I
a.
the requirement of 2200*F.
]
s b.
The amount of fuel element cladding that reacts chemically with
{
water or steam does not exceed 1% of the total amount of Zircaloy in the reactor.
(
The cladding temperature transient is terminated at a time when c.
the core geometry is still amenable to cooling. The localized i
cladding oxidation limits of 17% are not exceeded during or after quenching.
d.
The core remains amenable to cooling during and after the break.
The core temperature is reduced and decay heat is removed for an e.
extended period of time, as required by the long-lived radioac-tivity remaining in the core.
These criteria were established te provide significant margin in ECCS per-formance following a LOCA.
i In all cases, small breaks (less than 1.0 fts) yield results with more margin to the Acceptance Criteria limits than large breaks.
15.6.5.2 Sequence of Events and Systems Operations i
Should a major break occur, depressurization of the RCS results in a pressure decrease in the pressurizer. The reactor trip signal subsequently occurs when the pressurizer low pressure trip setpoint is reached.
A safety injection sig-nal (5:5) is gererated when the appropriate setpoint is reached.
The counter-measures will limit the consequences of the accident in two ways:
(
15.6-9
~
I be Ieke s g M-e9 anc L Reactor trip and borated water injection complement void femation a.
in causing rapid reduction of power to a residual level correspond-
)
ing to fission product decay heat. However. no credit is taken in the LOCA analysis for boron content of the injection water.
In ddition, the insertion of control rods to shut down the reactor neglected in the large break analysis.
b.
Inje ion of borated wcter provides for heat transfer from the core and pr ents excessive clad temperatures.
Description of Laro Break LOCA Transient The sequet:e of events o11owing a large break LO A are presented in Figure 15.6.5-1.
j j
Before the break occurs, t unit is in an equilibrium condition, i.e., the
=
heat generated in the core being removed via the secondary system.
During
]
blowdown, heat from fission F duct decay, hot internals and the vessel-
~
continues to be transferred to he reactor coolant.
At the beginning of the
~
blowdown phase, the entire RCS ntains subcooled liquid which transfers heat from the core ty forced convectio with some fully developed nucleate boiling.
Thereafter, the core heat transfer Cs based on local conditions with transition boiling and forced convection to sta m as the major heat transfer mechanisms.
The heat transfer between the Reactor olant System and the secondary system a
may be in either direction depending on he relative temperatures.
In the case of continued heat addition to the s ondary, secondary system pressure increases and the main steam safety valve may actuate to limit the pressure.
}
Makeup water to the secondary side is auto tically provided by the auxiliary feedwater system. The SIS actuates a feedw er isolation signal which isolates normal feedwater flow by closing the main fee ater isolation vah, and also initiates emergency feedwater flow by starting the auxiliary feedwater pumps, The secondary flow aids in the reduction of Rea or Coolant Systee' pressure.
When the Reactor Coolant System pressure falls be appr3ximately 1250 psia i.he upper head injection accumulators begin to inj t borated water directly into the reactor upper head region. This water is rected from the upper head directly to all but 8 peripheral assemblies in t core via the RCC guide tubes and UHI support columns. This flow provides add ional core cooling during the blowdown phase of the transient.
A detailed escription of the interactions of UHI water and those effects on the blowd and subsequent reflood transients is given in Reference 4.
When the Reactor Coolant Systes depressurizes to approximate. 400 psia, the accumulators begin to inject borated water into the reactor co ant loops.
Since the loss of offsite power is assumed, the reactor coolant p s are assumed to trip at,the inception of the accident. The effects of p coastdown are included in the blowdown analysis.
i 15.6-10 Rev.
S) s b
\\
Reactor trip and borated water injection complement void formation in l
causing rapid reduction of power to a residual level corresponding to a.
However, no credit is taken in the LOCA l
fission product decay heat.
ana?ysis for boron content of the injection water aiding in shutdown.
l In addition, the insertion of control rods to shut down the reactor is neglected in the large break analysis.
1 i
Injection of borated water provides for heat transfer from the core l
b.
and prevents excessive clad temperatures.
f l
Description of Laroe Break LOCA Transient 1
The sequence of events following a large break LOCA are presented in Figuro i
15.6.5-1.
f Before the break occurs, the unit is in an equilibrium condition, i.e., the During heat generated in the core is being removed via the secondary system.
blowdown, heat from fission product decay, hot internals and the vessel continues to be transferred to the reactor coolant. At the beginning of the l
blowdown phase, the entire RCS contains subcooled liquid which transfers heat from the core by forced convection with some fully developed nucleate Thereaf ter, the core heat transfer is based on local conditions with boiling'.
transition boiling and forced convection to steam as the major heat transfer,
mechanisms.
The heat transfer between the Reactor Coolant System and the secondary system In the may be in either direction depending on'the relative temperatures.
case of continued heat addition to the secondary, secondary system pressure increases and the main steam safety valves may actuate to limit the pressure.
Makeup water to the secondary side is automatically provided by the auxiliary The SIS actuates a feedwater isolation signal which feedwater system.
isolates normal feedwater flow by closina the main feedwater isolation valves andalsoinitiatesemergencyfeedwaterflowbystartingtheauxiliary The secondary flow aids in the reduction of. Reactor Coolant feedwater pumps.
System pressure.
When the Reactor Coolant system depressurizes to approximately 600 psia, the accumulators begin to inject borated water into the reactor coolant loops.
Since the less of offsite power is assumed, the reactor coolant pumps are assumed to trip at the beginning of the accident. The effects of pump coastdown are included in the blowdown analysis.
l The blowdown phase of the transient ends when the RCS pressure (initially assumed at 2280 psia) falls to a value appro.ching that of the containment Prior to or at the end of the blowdown, the mechanisms that are atmosphere.
responsible for the bypassing of emergency core cooling injection water into the RCS are calculated not to be effective. At this time (called end of Refill is complete bypass) refill of the reactor vessel lower plenum begins.when em vessel, which is bounded by the bottom of the fuel rods (called bottom of core recovery time).
15.6-10
)
m._:
._a.
h golCh f
CNS The blowdown phase of the transient ends when the RCS pressure (initially.
. (
ssumed at 2280 psia) falls to a value approaching that of the containment sphere. Prior to or at the end of the blowdown, the mechanisas that are l
onsible for the bypassing of emergency core cooling water injection into re the S are calculated not to be effective. At this time (called end of j
bypass refill of the reactor vessel lower plenum begins.
Refil) is complete j
when e ency core cooling water has filled the lower plenum of the reactor vessel wh h is bounded by the bottom of the fuel rods (called bottom of core recovery t
, BOC).
The reflood p se of the transient is defined as the time period lasting from
)
the end of refi until the reactor Nssel has been filled with water to-the extent that the re temperature rise has been terminated.
From the later
.l stage of blowdown d then the beginning of reflood, the safety injection accumulator tanks r. idly discharge borated cooling water into the RCS, con-l tributing to the fill g of the reactor vessel downcomer. The downcomer water elevation head p vides the driving force required for the reflooding-of the reactor core.
The low' head and high head afety injection pumps aid in the filling of the
~
downcomer and subsequently s ply water to maintain a full downcomer and complete the reflooding proce~. The safety injection pumped flow as a func,
tion of pressure is given in Ta le 15.6.5-6 for the large and small break Cases.
0 l
Continued operation of the ECCS pum supplies water during long-term cooling.
Core temperatures have been reduced long-term steady-state levels associated
('
l with dissipation of residual heat.
A er the water level of the refueling water storage tank (RWST) reaches a min um allowable value, coolant _for long-term cooling of the core is obtaine by switching to the cold leg re-circulation phase of operation in which s 'lled borated water is drawn from the containment n mps by the low head safet injection (RHR) pumps and re-turned to the RC5 cold legs.
The Containmen Spray System continues to operate to further reduce containment pressur Approximately 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> after i
initiation of the LDCA, the ECCS is realigned supply water to the RCS hot i
legs in order to ccitrol the boric acid concent tion in the reactor vessel.
Description of Small Break LOCA Transient Ruptures of small cross section will cause expulsion f the coolant at a rate which :an be accommodated by the charging pumps. Thes pumps would maintain an operational water level in the pressurizer peMttin the operator to execute an orderly shutdown.
The coolant which would be leased to the con-l tainment contains the fission products existing at equilib 'um.
l The maximum break size for which the normal makeup system can aintain the pressurizer level is obtained by comparing the calculated flow or the Reactor Coolant System through the postulatei break against the arging pump 1
makeup flow at normal Reactor Coolant System pressure, i.e., 2250
.fa..A 1
makeup flow rate from one centrifugal charging pump is typically ad ate to j
sustain pressurizer level at 2250 psia for a break through a 0.375 in J
diameter hole. This break results in a loss of approximately 17.25 lb c.
i 15,6-11 Rev.
1
h reflood phase of the transient is defined as the time period lasting from the and of refill until the reactor vessel has been filled with water to From the later extent that the core temperature rise has been terminated.
stage of blowdown and then the beginning of reflood, the safety injection accumuletor tanks rapidly discharge borated cooling water into the RCS, The downcomer contributirg to the filling of the reactor vessel downcomer.
water elevation head provides the driving force required for the reflooding of The low head and high head safety injection pumps aid in the filling cf the downcomer and subsequently supply water to maintain a full the reactor. core.
I The safety injection pumped downcomer and complete the reflooding process. flow as s'f m. tion of pr cases.
Continued operation of the ECCS pumps supplies water during l After the water level in the cooling.
associated with dissipation of residual heat.
refueling water storage tank (RWST) reaches a minimum allowable value, coolant for long-term cooling of the core is obtained by switching to the cold leg recirculation phase of operation in which spilled borated w The Containment Spray System cantinues to operate to to the RCS cold legs.
Approximately 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> after initiation -
further reduce containment pressure.
of the LOCA, the ECCS is realigned to supply water to the RCS hot legs in
- F order to control the beric acid concentration in the reactor vessel.
Description of Small Break LOCA Transient Ruptures of small cross section will cause expulsion of the coolant at a rate These pumps would maintain which can be accommodated by the charging pumps.
an operational water level in the pressurizar permitting the operator toT execute an orderly shutdown; containment contains the fission products existing at equilibrium.
The maximum break size for which the normal makeup system can maintain the
~
pressurizer level is obtained by comparing the calculated f A
makeup flow at normal Reactor Coolant System pressure, i.e., 2250 psia.
makeup flow rate from one centrifugal charging pump is typically adequate to i
sustain pressurizer level at 2250 psia for a break through a 0.375 inch This break results in a loss of approximately 17.25 lb/sec. -
diameter hole.
Should a larger break. occur, depressurization of the Reactor Coolant System causes fluid to flow into the loops from the pressurizer resulting in aReactor trip pressure and level decrease in the pressurizer, During the earlier part of low pressurizer pressure trip setpoint is reached.the small brea to overcome the flow maintained by the reactor coolant pumps' through the core Therefore, upward flow as they are coasting down following reactor trip.The Safety Injection System is ac f
through the core is maintained.
The consequences of the accident are l
the appropriate setpoint is reached.
limited in two ways:
15.6-11 t
Dek4e.
ed m
uld a larger break occur, depressurization of the Reactor Coolant System S
es fluid to flow to the Re6ctor Coolant System from the pressurizer result-t ca ing a pressure and level decrease in the pressurizer.
Reactor trip occurs then e pressurizer low pressure trip setpoint is reached.
The Safety In-The con-jection ' ystem is actuated when the appropriate setpoint is reached.
sequence 'of the accident a e limited in two ways:
\\
1.
Reactor' rip and borated water injection complement void formation in causing r id reduction of nuclear power to a residual level correspond-ing to the layed fission and fission product decay.
N 2.
Injection of b ated water ensures sufficient flooding of the core to prevent excessi clad temperatures.
N Before the break occurs e plant is in an equilibrium condition, i.e., the heat generated in the cor,is t,eing removed via the secondary system. During blowdown, heat from decay, ot internals and the vessel continues to be trans-ferred to the Reactor Coolan System. The heat transfer between the Reactor Coolant System and the second ry system may be in either direction depending on the relative temperatures.
In he case of contint.ed heat addition to the sec-ondary, system pressure increase'.and steam dump may occur. Makeup to the sec-ondary side is automatically prov ed by the auxiliary feedwater pumps.
The safety injection signal stops norma feedwater flow by closing the main feed-cater line isolation valves and init tes emergency feedwater flow by starting b
auxiliary feedwater pumps. The secon ry flow aids in the reduction of Reactor Coolant System pressure. When the Reac or Coolant System depressurizes to the upper head accumalator setpoint pressure,,the upper head accumulator begins in-
}
jecting borated water into the reactor ves el upper head.
A description of the
/
operation of upper head injection during smh I break transients can be found in UCAP 8479 Rev. 2 (Reference 4) and WCAP 9639 Reference 14).
When the RCS depressurizes to 400 psia, the col leg accumulators begin to in-ject water into the reactor coolant loops. The reactor coolant pumps are as-sumed to be tripped at the initialization of the a' ident and effects of pump coastdown are included in the blowdown analyses.
\\
15.6.5.3 Core and System Performance x
15.5.5.3.1 Mathematical Model j
The requirements of acceptable ECCS Evaluation Model are pres ted in Appendix K N
of 10CFR50.
N\\
Large Break LOCA Evaluation Model
\\
4 The analysis of a large break LOCA transient is divided into three ph'ses:
j (1) blowdown, (2) refill, and (3) reflood. There are three distinct t n-
\\
sients analyzed in each phase, including the themal-hydraulic transient (n the RCS, the pressure and temperature transient within the containment, a l
the fuel and cladding temoerature transient of the hottest fuel rod in the 15.6-12 l
l
I 1.. Reactor trip and borated water injection complemnt void formation in the
]
core and cause a rapid reduction of nuclear power to a residual level corresponding to the delayed fission and fission product decay, l
i Injection of borated water ensures sufficient flooding of the ' core to J
2.
prevent excessive clad temperatures.
8efore the break occurs the plant 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, het internals, and the vessel continues to be transferred to the Reactor Coolant System. The heat transfer between the 4
Reactor Coolant System and the secondary system may be in either direction
)
In the cese of continued heat depending on the relative temperaturas.
addition to the secondary, system pressure increases and steam dump may.
Nakeup to the secondary side is automatically provided by the occur.
auxiliary feedwater pumps. The safety injection signal stops normal feedwater I
flow by closing the main feedwater line isolation valves and initiates auxiliary feedwater flow by starting auxiliary feedwater pumps. The secondary flow aids in the reduction of Reactor Coolant System pressures.
When the RCS depressurizes to 600 psia, the cold leg accumulators begin to inject water into the reactor coolant loops. Due to the loss of offsite power assumption, the reactor coolant pumps are assumed to be tripped at the time of ;_
reacter trip during the accident and the effects of pump coastdown are-(
included in the blowdown antlyses, j
15.6.5.3 Analysis of Effects and Consequences Nethods of Analysis The requirements of an acceptable ECCS Evaluation Model are presented in The requirements of Appendix K Appendix K of 10 CFR 50 (Reference 3).
regarding specific model features were met by selecting models which provide a significant overall conservatism in the analysis. The assumptions made pertain to the conditions of the reactor and associated safety system equipment at the time that the LOCA occurs and include'such items as the core peaking factors, the containment pressure, and the performance of the ECCS Decay heat generated throughout the transient is also conservatively system.
The thermal-hydraulic j
. calculated as required by Appendix K of 10 CFR 50.
analyses reported in this section were performed with an upper head fluid J
temperature of Teold' Large Creek Evaluation Model lne analysis of a large break LOCA transient is divided into three phases:
There are three distinct (1) blowdown, (2) refill, and (3) reflood.(1) the thermal-hydraulic transient in the transients analyzed in each phase:
RCS, (2) the pressure and temperature transient within the containment, (3) and the fuel and cladding temperature transient of the hottest fuel rod in the Based on these considerations, a system of interrelated' computer codes core.
has been developed for the analysis of the LOCA.
The description of the various aspects of the LOCA analysis methodology is
't These documents describe the major given in References 4, 10, 13 and 14.
phenomena modeled, the interfaces among the computer codes, and the features nem
I bM AAd bf
-l CNS l'
ore. Based on these considerations, a system of. inter-related computer l
des has been developed for the analysis of the LOCA.
The scription of the various aspects of the LOCA analysis methodology is given 'n WCAP-8339 (Reference 5) and WCAP 8480 (Reference 4).
These documents describ the major phenomena modeled, the interfaces among the computer codes, i
and the tures of the codes which ensure compliance with the Acceptance Criteria.
he differences between the approved non-UHI Westinghouse Appendix K Model and he model used for there analyses are reported in WCAP-8479, Rev-ision 2 (Refe nce 4).
The thermal analyses reported in this section were per-formed with an per head fluid temperature of T The UHI accumulator pres-l cold.
sure set point en res UHI activation prior to upper head fluid flashing. The l
~
SATAN-VI, POWLOCTA, WREFLOOD, LOTIC, and LOCTA-IV codes, which are used in the LOCA analysis, are cribed in detail in References 4 through 9.
These codes are used to assess th core heat transfer geometry and to determine if the core remsins amenable to coo ing thoughout and subsequent to the blowdown, refill, I
i and'reflood phases of th LOCA. The SATAN-VI computer code analyzes the thermal-hydraulic, transient in the CS during blowdown and refill while POWLOCTA cal-culates the average channel xial temperature distribution during this periiod,of the transient. The WREFLOOD mputer code is used to calculate the thermal '
hydraulic transient during the flood phase of the accident. The LOTIC computer code is used to calculate the co ainment prt. 6ure transient during all three phases of the LOCA analysis.
Simi rly, the LOCTA-IV computer code is used to I
compute the thermal tt6nsient of the hottest fuel rod during the three phases, SATAN-VI is used to calculate the RCS ssure, enthalpy,. density and the mass and energy flow rates in the RCS, as wel as steam generator-energy transfer between the primary and secondary systems a function of time during the blow-down phase of the LOCA.
SATAN-VI also cale ates the accumulator water mass and internal pressure and the pipe break mass an energy flow rates that are assumed to be vented to the containment during blowdo At the end of the blowdown and refill phases, these data are transferred to th WREFLOOD code.
Also at the end of blowdown and refill phases, the mass and energ release rates are transferred
)
to the LOTIC code'for use in the determination of e containment pressure res-ponse during these phases of the LOCA.
Additional S AN-VI output data from the end of blowdown and refill, including the core pressu and the. core power decay transient, are input to the LOCTA-IV code.
With input from the. SATAN-VI and POWLOCTA codes, WREFLOOD s a system thermal-l hydraulic model to determine the core flooding rate), the co ant pressure and j
temperature, and the quench front height during the reflood ph es of the LOCA, j
WREFLOOD also calculates the mass and energy flow rate to the co ainment through 1
the break.
Since the mass flow rate to the containment depends u the ccre flooding rate and the local core pressure. which is a function of t containment y
backpressure, the transient pressure computed by the LOTIC code is in t to the WREFLOOD code.
For the analyses presented in this report, the contain nt pres-sure was conservatively assumed to be 0.0 psig, except for both 1.0 DECL breaks and the 0.8 DECLG perfect mixing break, which utilize a constant ba pressure conservatively assumed to be 0.5 psig. WREFLOOD is also linked to l
g the LOCTA-IV code in that thermal-hydraulic parhmeters from WREFLOOD are use i
I 15.6-13 Rev. 5 l
i P
of the codes which ensure compliance with the Acceptance Criteria. The SATAN-VI (Reference 5), WREFIDoD (Reference 6), IDTIC (Referencs 7), BART' (Reference 13), BASH (Reference 14) and IACTA-IV (Reference 8) codes are l
used to assess the core heat transfer geometry and to determine if the core remains amenable to cooling throughout and subsequent to the '
blowdown, refill, and reflood phases of the IACA. The SATAN-VI computer code. analyzes the thermal-hydraulic. transient in the RCS during blowdown. The WRITIACD and BASH computer codes are used.to. calculate the thermal-hydraulic transient during the reflood phase of the accident.
The IATIC computer code is used to calculate the containment pressure transient during all three rhases of the 14CA analysis..The BART computer code is used to calculate the fluid and heat transfer conditions in the core during reflood, and the 14CTA-IV computer code is used to compute the. thermal transient of the hottest fuel rod during the three phases.
Fuel parameters input to the IACTA-IV code were taken from a new version of the PAD. code (Reference 9).
SATAN-VI is used to calculate the RCS pressure, enthalpy, density, and-mass and energy flow rates, as well as steam generator heat transfer-between the primary and secondary systems, as a function of time during.
the blowdown phase of the IACA. SATAN-VI also calculates the accumulator water mass and internal ~ pressure and the pipe break mass and energy flow rates that are assumed to be vented to the containment during blowdown.
i At the end of the blowdown phase these data are transferred to the
{
J WREFIDOD code. The mass and energy release rates during blowdown and reficod are transferred to the LOTIC code for use in the determination of the containment pressure response during these phases of the IDCA.
Additional SATAN-VI output data from the end of blowdown and refill, e
including the core pressure and the core power decay transient,' are input of the 14CTA-IV code.
\\
BASH is an integral part of the ECCS evaluation model which provides a more realistic thermal-hydraulic simulation of the reactor core and RCS during the reflood phase of a LOCA.
Instantaneous values of accumulator conditions and safety injection flow at the time of completion of lower plenum refill are provided to BASH by WREFLOOD. Figure '15.6.5-2 illustrates how BASH has been substituted for WREFLOOD in calculating transient values of core inlet flow, enthalpy, and pressure for the detailed fuel rod model, LOCTA. A more detailed description of the BASH code is available in Reference 14. The BASH code provides a much more' sophisticated treatment of stear / water flow phenomena in the reactor coolant system during core reflood. A more dynamic
)
interaction between the tore thermal-hydraulics and system behavior is expected, and recent experiments have Dorne this out.
In the BASH code reflood model, BART provides the entrainment rate for a given flooding rate, then a system model determines loop flows and pressure drops in response to the calculated core exit flow. An updated inlet flow is used to ecleulate a new entrainment rate. This system will produce a more dynamic flooding transient, which reflects the close coupling between core thermal-hydraulics and loop behavior, a
The LOTIC code is a mathematical model of the ice condenser containment.
LOTIC is described in detail in Reference 7.
LOTIC is run using output from SATAN and WREFLOOD, which provide the necessary mass and energy releases to the containment.
In this analysis.the WREFLOOD/LOTIC system is used only-to provide containment boundary conditions required by BASH.
15.6-13
hghW evvY
$4C'L CNS j
by OCTA-IV in its calculation of the fuel temperature.
LOCTA-IV is used thro hout the analysis of the LOCA transient to calculate the fuel clad-ding t perature and metal-water reaction of the hottest rod in the core.
Dynamic team cooling is included in the LOCTA-IV calculation as described j
in Refere e 10.
]
A schematic r resentat'an of the computer code interfaces is given in Figure 15.6.5-2.
The large break an lysis was performed with the NRC approved Westinghouse UHI ECCS Evaluations Mo 1 (Reference 4).
Small Break LOCA Evalu tion Model The WFLASH program used i ' the analysis of the small break _ loss-of-coolant accident is an extension o the FLASH-4 (Reference 11) code developed at the y
Westinghouso Bettis Atomic P er Laboratory.
The WFLASH program permits a detailed spatial representatio.of the Reactor Coolant System.
~
The Reactor Coolant System is no lizedintovolumesinterconnectedbyflow-paths.
The broken loop is modeled xplicitly with the intact loops lumped a
y
?
into a second loop.
The transient b' avior of the system is determined from the governing conservation equations mass, energy and momentum applied throughout the system. A detailed desc iption of WFLASH is given in References 12 and 13.
}
s The use of VFLASH in the analysis involves,'among other things, the represen-tationofthereactorcoreasaheatedcontrh{volumewiththeassociated bubble rise medel to permit a transient mixtu g height calculation.
The multinode capat;ility of the program enables an applicit and detailed spatial representation cf various system components.
InNarticular, it enables a i
proper calculation of the behavior of the loop sea during a loss-of-coolant transient.
Cladding thermal analyses are performed with the LOCTA V (Reference 9) code j
which uses the RCS pressure, fuel rod power history, st m flow past the un-covered part of the core and mixture height history from e WFLASH hydraulic calculations as input.
A schematic representation of the computer code interfaces is iven in Figure
- 15. 6. 5,- 3.
4 The small break analysis was performed with the approved We:tingh se ECCS Small i
Break Evaluation Model (References 4, 9, 12, and H).
i 15.6.5.3.2 Input Parameters and Initial Conditions Table 15.6.5-1 lists important inpat parameters and initial conditions u d in the analysis, s
'i 15.6-14 Rev. 5
)
i
l l
i i
The LOCTA-IV code.is a computer program that evaluates fuel, cladding and
]
1 coolant. temperatures during a LOCA. A more complete description than is presented here can be found in Reference 8.
During refill and reflood, the ECCS model-uses a code consisting of LOCTA coupled with BART to yield a.
significant improvement in fuel rod behavior prediction.
In the BART/LOCTA l
detailed fuel rod model, for the calculation of local heat transfer coefficients, the empirical FLECHT correlation is replaced by the 8 ART code.
8 ART employs rigorous mechanistic models to generate heat transfer 1
coefficients appropriate to the actual flow and heat transfer regimes experienced by the LOCTA fuel rods. This is considered a more dynamic realistic approach than relying on a static empirical correlation.
Small Break LOCA Evaluation Nodel The NOTRUMP computer code is used in the analysis of loss-of-coolant accidents
'i due to small breaks in the reactor coolant system. The NOTRUMP computer code is a state-of-the-art one-dimensional general network code consisting of a number of advanced features. Among these features are the calculation of thermal non-equilibrium in all fluid volumes, flow regine-dependent drift flux calculations with counter-current flooding limitations, mixture level tracking logic in multiple-stacked fluid nodes, and regime-dependent heat transfer correlations. The NOTRUMP small break LOCA emergency core cooling system (ECCS) evaluation model was developed to determine the RCS response to design l
basis small break LOCAs and to address the NRC concerns expressed in NUREG-0611, " Generic Evaluation of Feedwater Transients and Small Break I
Loss-of-Coolant Accidents in Westinghouse Designed Operating Plants."
j f
In NOTRUMP, the RCS is nodalized into volumes interconnected by flowpaths.
The broken loop is modeled explicitly with the intact loops lumped into a second loop. The transient behavior of the system'is determined from the governing conservation equations of mass, energy and momentum applied throughout the system. A detailed description of NOTRUMP is given in References 11 and 15.
The use of NOTRUMP in the anclysis involves, among other things, the representation of the reactor core as heated control volumes with an associated bubble rise model to permit a transient mixture height calculation. The multinode capability of the program enables an explicit and j
. detailed spatial representation of various system components. In particular, it enables a proper calculation of the behavior of the loop seal during a loss-of-coolant transient.
Cladding thermal analyses are performed with the LOCTA-IV (Reference 8) code which uses the RCS pressure, fuel rod power history, steam flow past the uncovered part of the core, and mixture height history from the NOTRUMP hydraulic calculations, as input.
15.6-14
De)<k d Rop N A m
he bases used to select the numerical values that are input parameters to j
/.(
analysis have been conservatively determined from extensive sensitivity st ies (Refcarence 4). ~ In addition, the requirements of Appendix K regarding spec fic model features were met by selecting models which provide a signifi-cant erall conservatism in the analysis.. The assumptions made pertain to the co itions of the reactor and associated safety system equipment at the time th the LOCA occurs and include such items as the core peaking factors, the conta ent pressure, and the performance of the ECCS system. Decay heat generated roughout the transient is also conservatively calculated as re-quired by endix K, 10CFR50.46.
The worst brea (CD = 1.0) was run with a variation in UHI accumulator volume l
delivery for th erfect (1043 ft3) and imperfect mixing case (877 ft3) assumptions. The delivered volume considered in the analyses encompasses the g
volume delivery ba (156 fts) associated with UHI delivery uncertainties at a 95 percent probab ity level.
Cases presented herei provide the results of a conservative application of this range of values.
n addition,~ UHI volume deliveries for each case-presented herein will di fer somewhat due to variation ir UHI flowrate during the time of isolation val closure and be dependent on oischarge coefficiertt j
assumed.
s af The imperfect mixing case was nalyzed to develop a low delivery volume since the upper head drains earlier <
the transient and subsequently voids the lower plenum and core. The impe' ect mixing case was also run at a higher i
f(
pressure (1300) than the perfect xing case (1200) to allow for uncertainty in the accumulator setpoint pressur' Similarly, the high pressure for the imperfect mixing case represents the.ost conservative case since the smaller accumulator volume would be delivered a shorter amount of time and earlier in the blowdown transient, thereby prov.ing for a longer core heatup time.
i 15.6.5.3.3 Results Large Break Results Based on the results of the LOCA sensitivity stu s (Reference 4), the i
limiting large break was found to be double-ended
.1d leg guillotine (DECLG).
i Therefore, only the DECLG break is considered in the arge break ECCS perfor-mance analysis.
Calculations were performed for a ran' of Moody break discharge coefficients (C ).
The results of these cale ations are summarized D
in Tables 15.6.5-2 through 15.6.5-5.
The mass and energy release data for the break resulting in t highest calculated peak cladding temperature is presented in Section 6.
1.5.
Figures 15.6.5-4 through 15.6.5-63 present the parameters of prin al interest from the large break ECCS analyses.
For all cases analyze 1
l transients of the following parameters tre presented:
(
4 15.6-15 Rev. 1 q
]
1 i
A schematic representation of the computer code interfaces is given in Figure i
15.6.5-3.
3 The small break analysis was performed with the approved Westinghouse ECCS Small Break Evaluation Model (References 8,11 and 15).
Laroe Break Input Parameters and' Initial Conditions Table 15.6.5-1 lists important input parameters and initial conditions used in the large break analyses.
Small Break Input Parameters and Initial Cenditions Table 15.6.5-1 lists important input parameters and initial conditions used in the sm:11 break analyses.
The axial power distribution and core decay power assumed for the small break analyses are shown in Figures 15.6.5-43 and 15.6.5-44.
Safety injection flow rate to the Reactor Coolant System as a function of the system pressure is used as part of the input. The Safety Injection (SI) system was assumed to be delivering to the RCS 30 seconds af ter the generatierr of a safety injection signal.
I For these analyses, the SI delivery considers pumped injection flow which is depicted in Figure 15.6.545 as a function of RCS pressure. This figure represents injection flow from the SI pumps based on performance curves The 30 second delay includes time degraded 5 percent from the design head.
required for diesel startup and loading of the safety injection pumps onto the The effect of flow from the RHR pumps is not considered here emergency buses.
since their shutoff head is lower than RCS pressure during the time portion of the transient considered here. Also, minimum safeguards Emergency Core 4
Cooling System capability and operability has been o sumed in this analysis.
The hydraulic analyses are periormed'with the NOTRUMP code using 102% of the E
licensed NSSS core power. The core thermal transient analyses are performed with the LOCTA-IV code using 102% of licensed NSSS core power.
Large Break Results Based on the results of the LOCA sensitivity studies (Reference 12), the j
limiting large break was found to be double-ended cold leg guillotine (DECLG). Therefore, only the DECLG break is considered in the large break ECCS performance analysis. Calculations were performed for a range of Woody Consistent with the methodology break dischar e coefficients (C ).
described in eference 16 the bheak size which resulted in the worst case for minimum safe +y injection was used in a calculation in which no failures of the The results of these calculations are ECCS were aest,med (Maximum safeguards)15.6.5-5.
summarized in Tables 15.6.5-2 through Figures 15.6.5-4 through 15.6.5-M present the parameters of principal interest from the large break ECCS analyses. Transients of the'following l
parameters are presented for each discharge coefficient analyzed, and where appropriate for the worst break maximurt safegucrds case.
15,6-15 l
I h 4 Ch
@ INS k&f l8L C f e igures 15.6.5-4 The following quantities are presented at the clad brough L'urst location and at the hot spot (location of Fig s 15.6.5-18 aaximum clad temperature) both on the hottest 'uel rod (hot rod):
1.
fluid quality 2.
mass velocity 3.
heat transfer coefficient.
The heat transfer coefficient shown.is calculated i
by the LOCTA IV code.
Figures 15.6.5-1 The system pressure shown is the calculated pres-through sure in the core.
Core flowrates, and core void Figuras 15.6.5-43 fraction are al v presented.
Figures 15.6.5-44 These figures show the hot spot clad temperature through transient and the clad temperature transient at Figure 15.6.5-53 the burst location.
The fluid temperature shown
's also for the hot spot and burst location.
The n dal notation of the figures is defined in Ta le 15.6.5-7.
Figures 15.6.5-54 Thes figures show the core reflood transient.
through b
Figure 15.6.5-63 Figures 15.6.5-64 These fig esshowtheEmergencyCoreCooking through System for 11 cases analyzed. Both UHI and cold T
Figur^e 15.6.5-73 leg accumula rs are incladed in the figures. As J
described ear er the cold leg accumulator deli-very during b1 down is discarded until the end of bypass is calcul ted.
Cold leg accumulator flow, however, is estab shed in refill reflood calcu-lations. The cold eg accumulator flow assumed is j
the sum of that inj ted in the intact cold legs.
Figures 15.6.5*74 These figures show the total cold leg accumulator through mass injection prior to.nd of bypass, accumulator Figure 15.6.5-78 mass spilled out break, 1culated bypass deficit, and vessel inventory.
The maximum cladding temperature calculated for a la e break is 2155'F which is less than the Acceptance Criteria limit of 2200*F o 10 CFR 50.46.
The max-imum local metal water reaction is 6.2 percent, which i well below the embrit-tiement limit of 17 percent as required by 10 CFR 50.46.
The total core metal-water reaction is less than 0.3 percent for all breaks, a compared with 1 per-cent criterion of 10 CFR 50.46, and the cladding temperatu transient is ter-minated at a time when the core geometry is still amenable t cooling. As a result, the core temperature will continue to drop and the ab ity to remove decay heat generated in the fuel for an extended period of ti ill be provided.
Table 15.6.5-12 contains an evaluation of the potential impact o using fuel
-od models presented in NUREG-0630 on the Catawba limiting break c e.
15.6-16
- v. 5 T
0 Figure 15.6.5 The following quantities are presented for the through-location of maximum clad temperature.
)
Figure 15.6.5-15
- 1. fluid quality
- 2. mass velocity
- 3. heat transfer coefficient The heat transfer coefficient shown'is calculated by the IOCTA-IV Code.
Figure 15.6.5-16 The system pressure shown is the calculated through pressure in the core.
Core flowrates are also Figure 15.6.5-21 presented.
Figure 15.6.5-22 These figures show the clad temperature transient-through and the fluid temperature at the maximum clad Figure 15.6.5-29 temperature location.
Figure.15.6.5-30 These figure show the core reflood transient.
through
~
Figure 15.6.5-37 Figure 15.6.5-38 These figures show the cold leg accumulator through delivery during blowdawn.
Figure 15.6.5-40 Figure 15.6.5-41 The calculated containment pressure is presented and for the minimum and maximum Figure 15.6.5-42 safeguards cases.
The maximum cladding tempersture calculated'for the large break is 1703.7 F which is less than the Acceptance criteria limit of-2200 F of 10 CFR 50.46.
The maximum local metal water reaction is 0
+.7t percent, which is well below the embrittlement limit of 17 percent as required by 10 CFR 50.46.
The total core metal water reaction is less than 0.3 percent for all breaks, as compared with the 1 percent criterion of 10 CFR 50.46, and the cladding temperature transient is terminated at a time when the core geometry is still amenable to cooling.
As a result, the core temperature will continue to drop and the ability to remove decay heat generated in the fuel for an extended period of time will be provided.
Small Break Results 1
As noted previously, the calculated peak cladding comparature resulting from a small break LOCA is less than that calculated for a large break.
A range of small break analyses are presented which establishes the-limiting break size.
The results of these analyses are summarized in Tables 15.6.5-7 and 15.6.5-8.
Figures. 15.6.5-45 through 15.6.5-57 present the principal parameters of interest for the small break ECCS analyses.. For all cases analyzed, the following transient parameters.are included:
a.
RCS pressure b.
core mixture height c.
hot spot clad temperature
Insert A
he r-e.
j
(.
e
)
CNS l
1 Saa11 Bre Results As noted pre ugly, the calculated peak cladding temperatur esulting from reak. A range a small break LOCA is less than that calculated for a large, limiting break of small break analy s are presented which establishes t,h'e size. The results of se analyses are summarized in ables 15.6.5-8 and 15.6.5-9.
Figures 15.6.5-79 through 15.. -92 present the ncipal parameters of in-terestforthesmallbreakECCSaM1yses.
For /11 cases analyzed, the follow-ing transient parameters are includ :
/
/
a.
RCS pressure,
]
i b.
core mixture height,
{
c.
hot spot clad temperatyre, and d.
core power after r actor trip.
(Figure 15.. -83 applies for all 3' cases) l
,e
\\
For limiting break a yzed, the following additional trans t parameters
{
are presented:
a.
.or steam flowrate, b.
ore heat transfer coefficient, and
{
)
hot spot fluid temperature.
e maximum calculated peak cladding temperature for the small breaks analyze is 1218'F. These results are well below all Acceptance Criteria limits of f
10 CFR 50.46 and in all cases are not limiting when compared to the results I
presented for large breaks.
15.6.5.4 Environmental Consecuences The postulated consequences of a Design Basis Accident are calculated for
- 1) offsite and 2) control room operators.
15.6.5.4.1 Offsite Dose Consequences The offsite radiological consequences of a Design Basis Accident are calculated based on the following assumptions and parameters.
1.
One hundred percent of the' core noble gases and twenty five percent of the core iodines are released to the containment atmosphere.
2.
Fifty percent of the core iodines are deposited in the sump.
{
15.6-17
$%S&
k For the limiting break analyzed (4 inch), the following additional transient 15.6.511):
parametera are presented (Figures 15.6.5-srthrough a.
core steam flow rate b.
core heat transfer coefficient.
c.- hot spot fluid temperature The maximum calculated peak cladding temperature for the small breaks analyzed is 1304'F. These results are well below all Acceptance Criteria limits of 10 CFR 50.46 and no case is limiting when compared to the results presented for large breaks.
i 4
i i
CNS t.Y
- A Y 0fAO Wl hO fgys EFERENCES FOR SECTION 15.6 1.
Burnett, T. W. T., et. al., "LOFTRAN Code Description", WCAP-7907-P-A Proprietary) and WCAP-7907-A (Non-Proprietary), April,1984.
\\
2.
Ch mer, H., Boman, L. H., Sharp, D. R., " Improved Thermal Design l
Proc ures", WCAP-8567, July 1975.
l 3.
- Young,
. Y., " Westinghouse Emergency Core Cooling System Evaluation Model Appi ation to Plants Equipped with Upper Head Injection",
WCAP-8479,( oprietary), and WCAP-8480 (Non-Proprietary), January, 1975.
i 4.
Bordelon, F. M., _. W. Massie, and T. A. Borden, " Westinghouse ECCS I
Evaluation Model-ary", WCAP-8339, (Non-Proprietary), July 1974.
5.
Bordelon, F. M., et. a
" SATAN-VI Program:
Comprehensive Space Time Dependent Analysis f Loss of Coolant", WCAP-8302 (Proprietary) and WCAP-8303 (Non-Propri ry) June 1974.
l 6.
Kelly, R. D., et. al., "Calcu ted Model for Core Reflooding After a Loss of Coolant Accident (WREFL Code)", WCAP-8170 (Proprietary) anu WCAP-8171 (Non-Proprietary), June 74.
7.
Hsieh', T., and Raymond, M., "Long Te Ice Condenser Containment l
LOTIC Code Supplement 1", WCAP-8355, Su lement 1 (Non-Proprietary) May
~
1975 and WCAP-8354, Supplement 1 (Proprie ry), July 1974.
8.
Bordelon, F. M., et. al., "LOCTA-IV Program:
oss of Coolant Transient Analysis", WCAP-8301 (Proprietary) and WCAP-83 (Non-Proprietary), June
)
1974.
9.
Eiche1dinger, C., " Westinghouse ECCS Evaluation Mode February, 1978 Version", WCAP-9220.(Proprietary) February, 1979 and P-9221 (Non-Proprietary) February, 1978.
10.
Porsching, T.
A., et. al., " FLASH-4:
A Fully Implicit FOR N-IV Program for the Digital Simulation of Transients in a Reacto Plant".
WAPD-TM-84, Bettis Atomic Power Laboratory, March 1969.
11.
Exposito, V.
J., K. Kesavan and B. A. Maul, "WFLASH - A FORTRAN Computer Program for Simulation of Transients in a Multi-Loop PWR",
WCAP-8200, Revision 2 (Proprietary) July 1974, and WCAP-8261, Revisi 1
(Non-Proprietary) July 1974.
12.
Skwarek, R., W. J. Johnson, P. E. Meyer, " Westinghouse Emergency Core Cooling System Small Break October 75 Model", WCAP-8970 (Proprietary) and WCAP-8971 (Non-Proprietary), January 1979.
13.
Docherty, P. J., et al, " Report on Small Break Accidents for Westinghouse Nuclear Steam Supply System (NSSS) With Upper Head Injection (UHI)",
i WCAP-9639 (Non-Proprietary), December 1979.
s 15.6-20 Rev. 14 a
i
l REFERENCES FOR SECTION 15.6 Burnett, T. W. T., et. al., 'LOFTRAN Code Description", WC'AP-7907, June 1.
]
1972.
2.
Chelemer, H., Boman, L. H., Sharp, D. R., " Improved Thornal Design Procedures", WCAP-8587. July 1975.
l
" Acceptance Criteria for Emergency Core Cooling System for Light Water f
3.
Cooled Nuclear Power Reactors, 10 CFR 50.46 and Appendix K of 10 CFR
- 50. Federal Register, Volume 39, Number. 3. January 4,1974.
4.
Bordelon, F. M., Massie, H. W. and Borden, T. A., " Westinghouse ECCS Evaluation Model-Summary", WCAP-8339 -(Non-Proprietary), July 1974.
J 5.
Bordelon, F. M., et. al., ' SATAN-VI Program: Comprehensive Space Time Dependent Analysis of Loss of Coolant". WCAP-8302, (Proprietary) June 1974,andWCAP-8303.(Non-Proprietary), June 1974.
6.
Kelly, R. D., et. al., " Calculated Model for Core Reflooding After a Loss of Coolant Accident (WREFLOOD Code" WCAP-8170 (Proprietary) and WCAP-8171 (Non-Proprietary), June 1974 l
7.
Hsieh, T. and Raymond, W., "Long-Term Ice Condenser Containment LOTIC Code Supplement 1", WCAP-8355 Supplement.1, Nay 1975, WCAP-8354 (Proprietary), it.ly 1974.
8.
Bordelon, F. M., et. al., 'LOCTA-IV Program: Less of Coolant Transient Analysis", WCAP-8301. (Proprietary) and WCAP-8305, (Non-Proprietary),
j June 1974.
9.
Rahe, E. P., Westinghouse letter to Thomas, C. O., U.S.N.R.C., Letter Number NS-EPR-2673, October 27,1982
Subject:
" Westinghouse Revised PAD Code Thermal Safety Model", WCAP-8720 Addendum 2 (Proprietary).
. 10. Westinghouse ECCS Evaluation Model,1981 Version". WCAP-9220-P-A, Rev.1 (Proprietary), WCAP-9221-A, Rev. 1 (Non-Proprietary), February, 1982.
- 11. Lee, H., Tauche, W. D., Schwarz, W. R., " Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code", WCAP-10081-A, August 1985.
- 12. Salvatori, R., "Wastinghouse Emergency Core Cooling System - Plant Sensitivity Studi>as", WCAP-8340, (Proprietary) July 1974.
- 13. Young, W., et. al., 'BART-1A: A Computer Code for the Best Estimate Analyzed Reflood Transients" WCAP-9561-P-A,1984 (Westinghouse Proprietary).
f
i 4
An Integrated Core and RCS Reflood Code Kabadi, J. N., et. al., '8 ASH:
14.
for Analysis of PWR Loss-of-Coolant Accidents", WCAP-10266,1984 (WestinghouseProprietary).
Meyer, P. E. and Kornfilt, J., "NOTRUMP, A Nodal Transient Small Break 15.
j and General Noterk Code", WCAP-10080-A, August 1985.
1 Rahe. E. P. (Westinghouse), letter to Tedesco, R. L. (USNRC), No.
16.
HS-EPR-2538, December 1981.
4 I
l
\\
~~*#
DetA ud Aelu e TABLE 15.6.5-1 Input Parameters Used in the ECCS Analysis
(
3411 Core r* (swt)
Peak Li
.r Power (Includes 102% factor) kW/ft Larg' Break 12.88 Small eak 12.22 Total Peaking ctor, F 2.32 q
Axial Peaking Fac r, F 1.495
]
Z a
Power Shape large Break Chopped Cosine Small Break See Figure 15.6.5-79 Fuel Assembly Array 17 x 17 (optimized)
Cold Leg Accumulator Water lume (nominal) 1050 (Ft / accumulator) 2 Cold Leg Accumulator Tank Volume (nominal) 1350 (Ft / accumulator) 3
,,P Cold Leg Accumulator Gas Pressure (m nimum) (psia) 400
)
(-
UHI Accumulator injected volume (f t3)
Perfect mixing case 1043 Imperfect mixing case 877 Small breaks 860 UHI Accumulator initial pressure Perfect mixing case and small breaks (psia 1200 Imperfect mixing case (psia) 1300 Safety Injection Pumped Flow (> 10 inch)
See Table 15.6.5-6 3
Safety Injection Pumped Flow (i 10 inch)
See Trble 15.6.5-6 Containment Parameters psig (See text for additional d tails)
Initial Loop Flow (1b/sec) 99 Vessel Inlet Temperature ('F) 561.6 Vessel Outlet Temperature (*F) 620.0 Reactor Coolant Pressure, psia 2280 Steam Pressure (psia) 1000 Steam Generator Tube Plugging Level (%)
0 k
Notes:
- Two percent is added to this power to account for calorimetric error.
Rev. 14
TABLE 15.6.5-1 Inout Parameters Used in the ECCS Analysis 3411 Core Power * (mwt)
Peak Linear Power (kw/ft)
(Includes 102% factor)
Large Break 12.88 Small Break 12.22 2.32 Total Peaking Factor, Fg i
Power Shape Large Break Chopped Cosine i
Small Break See Figure 15.6.5-43 i
Fuel Assembly Array 17 x 17 (optimized)'
ulator Nominal Cold Leg Accg/ accumulator)
Water Volume (ft 1050 ator Tank cold Leg Accumug/ accumulator) 8 Volume (ft 1350 l
Minimum Cold Leg Accumulator Gas Pressure (psia) 600 Pumped safety Injection Flow Large Break See Table 15.6.5-6 l
Small Break See Table 15.6.5-45 Steam Generator Initial Pressure (psia) 954 Steam Generator Tube Plugging Level (%)
10 Initial Loop Flow (lbm/sec) 9701.4 1
I Vessel Inlet Temperature (OF) 560.2 Vessel Outlet Temperature (OF) 621.3 Reactor Coolant Pressure (psia) 2280 Esfauit:
- Two percent is added to this power to account for calorimetric uncertainty
1 2
TABLE 15.6.5-2 s
Larce Break LOCA Time Secuence Of Events 3
Perfect Mixing 1.0 C
0.8 C
0.6 C
0.4 1
D =ECLG D =ECLG D =ECLG D =ECLG
]
C D
D D
D (sec)
(sec)
(sec)
(see) i Start 0.0 0.0 0.0 0.0 Reactor Trip Signal
.8 0.8 0.8 0.8 UHI Accumulator Injection 3.1 3.4 4.2 6.3 Safety Injection Signal 4.8 4^
5.0 5.3 1
Cold Leg Accumulator Injection 12.2 14.
, 17.4 22.8 l
UHI Accumulator Injection Complete 21.6 23.3 25.5 29.7 Pump Injection 29.8 29.9
.0 30.3 End of Bypass 48.9 53.2 69.5 66.3-Bottom of Core Recovery 106.5 104.6 104.0 112.0 :_
Cold leg Accumulator Empty 112.3 115.8 US. 6 1
4 hh TA&LE 15, 4. 5 - A 4
l Rev. 5 w-
'l TABLE 15.6.5-2 j
Larce Break LOCA T1me Secuence of Events 1.0 C
0.8 C
0.6 p =ECLG p =ECLG D =CLG-C D
D DE (sec)
(sec)
(sec) 1 i
Start 0.0 0.0 0.0 Reactor Trip Signal 0.447 0.451 0.457
]
Safety Injection Signal 2.5 2.6 2.7 Cold Img Accumulator Injection 10.4 12.3 15.2 Pumped Injection 32.5 32.6 32.7 End of Bypass 23.40 26.45 31.14 End of Blowdown 23.40 26.50 31.73;.
1 Bottom of Core Recovery 36.58 39.85 44.78 Cold Leg Accumulator Empty 51.44 53.88 57.89 i
//
't i
i l
l l
I 1
Table 15.6.5-3 Large Break LOCA Time Seouence of Events Imperfect Mixing i
1.0 C 0.8 C 0.6 C 0.4 C
D =ECLGD =ECLG'D =ECLGD =ECLG D
D D
D 1
(sec)
(sec)
(sec)
(sec)
St. art 0.0 0.0 0.0 0.0 Reactor Trip Signal 0.8 0.778 0.795
-0.823 UHIAccumulatorInjection 2.2 Safety-Injection Signal 4.9 5.0
- 5. 3 '
ColdLegAccumulatorInjection 13.4 15.1 17.7 23.3 UHIAccumulatorInjectionComplete 19.4 PumpInjection 29.8 29.9 30.0 30.$ ;-
End of Bypass 48.8 45.8 7.5 55.7
(-
Bottom of Core Recovery 70.2 57.5 68.
76.4 Cold Leg Accumulator Empty 112.2 115.0 118.8 26.6
- Not available
'l
- bClLka, cnad
$cp{sco wid ggg TMLE
/ 5, f. 5'
~S '
l 1
1 1
1 1
.i 1
/*
Rev. 14 l
L
TABLE 15.6.5-3 Larce Break LOCA Time Secuence of Events Maximum Safeguards i
0.8 Cp =ECI.f, -
D (sec)
I J
0.0 Start Reactor Trip Signal 0.451 Safety Injection Signal
- 2.6
' Cold Leg Accumulator Injection.
12.3 32.6 Pumped Injection 26.45 a
p End of Bypass End of Blowdown 26.50 Bottom of Core Recovery 39.28 55.44 Cold Leg Accumulator Empty I
I 4
1 i
)
t-1
T/.8LE 15,6.5-4 Laroe Break LOCA Results
)
Fuel Claddinc Data
/
1 Perfect Mixing 0.8 C
0.6 C
0.4 D =ECLG_ D =ECLG_
D =ECLG 1.0 C
O CD =ECLG _
D D
O l
}
RESULTS 2155 2111 2103 2000
' Peak Clad Temperature (*F) 5.75 5
- 7. 5 7.5 Peak Clad Temperature Location (ft) 5.7 6.2 5.9 2.5 Local Zr/H O Reaction (max), (%)
2
' 7.5
_ 6. 0
- 7. 5 7.5 Local Zr/H O Location, (ft)
< 0. 3 '
<0.3 2
<0.3
<0.
Total Zr/H O Reaction, (%)
82.8
~
2 73.5 70.8 68.0
' Hot Rod Burst Time, (sec) 6.0 6.0 5.75 5.7
, Hot Rod Burst Location, (ft)
)
D ele 4e.
ud R pl c.e. with new 15.6.6 )
~T~AElE.
i l
l i
]
Rev. 5
p TABLE 15.6.5-4 Larce Break LOCA Results l
Fuel Claddine Data 1.0 C
0.8 CD = 0.6 Cp =ECLG p =ECLG
_ DECLG D
D Paak Clad Temperature (OF) 1666.8 1703.7
-1681.3 Peak Clad Temperature Location (ft) 7.0 7.0 6.5 l
Local Zr/H O Reaction (max), (%)
1.77 2.44 1.82 2
Local Zr/H O Location (ft) 7.0 7.0 7.0 2
Total Zr/H O Reaction- (%)
<0.3
<0.3
< 0. 3, 2
Not Rod Burst Time (sec) 58.0 59'.5 58 7.?
Hot kod Burst Location (ft) 6.25 6.25 6.25 i
i
I 1
Table 15.6.5-5 Large Break LOCA Results
(
Fuel Claddina Data j
Imperfect Mixing D =ECLGD =ECLGC,j =ECLG D =ECLG 1.0 C 0.8 0.6 C 0.4 C
D D
D O
RESULTS l
1 Peak Clad Temperature (*F) 1992 2110 2144 2006 Peak Clad Temperature Location (ft) 7.
7.5 7.5 7.5 Local Zr/H O Neaction (max), (%)
4.0
.9 6.5 4.1 2
Local Zr/H O Location, (ft) 7.5 7.5 7.5 7.5 2
t l
Total Zr/H O Reaction, (%)
<0.3
<0.3
.3
<0.3.
2 Hot Rod Burst Time, (sec) 108.2 64.0 64.0 77.0a
~
Hot Rod Burst Location, (ft) 6.25 5.75 6.5 5.5
(
D e.t a h
^ d AepI<c.e_ witk ees TAsag js, c, 5 - 3 l
l s
a Rev. 14
TABLE 15.6.5-5 Larae Break LOCA Results Fuel Claddine Data Maximum Safeguards 0.8 Cp =ECLG D
Peak Clad Temperature.(CF) 1697.7 6.75 Peak.. Clad Temperature Location (ft)
Local Zr/H O Reaction (max), (%)'
4.71 2
7.0 Local Zr/H O Location (ft) 2
<0.3 Total Zr/H 0' Reaction (%)
2 4
2
' Hot Rod Burst Time (sec) 119.7 7.0 Hot Rod Burst Location (ft)
TABLE 15. 6. 5-6 Safety Injection Pumped Flow Assumed for Breaks Greater Than or Equal to 10 Inches
,)
SI Flow Pre re (psi (Ib/sec) 14.7 493.2 34.7 437.2 54.7 378.6 74.7 315.5 114.7 198.4 214.7 100.2 l
614.7 81.2 1014.7 58.5 3014.7
.0.0 s.
a
,F Assymed for Breaks Le Than 10 Inches
}
Pressure SI Flow (psia) lb/sec) 14.7 4.70 114.7 19 7
414.7 92.
814.7 75.28 1214.7 52.93 J
1614.7 19.18 l
2014.7 2.38 2414.7 0.0 Ot lf,44 4M btp I44 t wi%
new TASLC
/S,(.,f-6 O
1 i
l 1
TABLE 15.6.5-6 Safety injection Pumped Flow j
Assumed for Ersaks Greater Than or Equal to 10 Inches NININUM SAFEGUARDS:
51 Flow l
Pressure (1b/see),
(psia) 436.6 14.7 374.7 1
310.0
{
34.7 54.7 238.6 1
74.7 l
113.3 114.7 96.1 i
214.7 72.4 614.7 NAXINUM SAFEGUARDS:
$1 Flow Pressure (1b/see)
(psia) 1237 14.7 1149 1
34.7 1056 54.7 958-74.7 851 94.7 732 114.7 171 214.7 166-314.7 4
E
[
l 1
4 TABLE 15.6.5-7 N004L REPRESENTATION OF CORE TRANSIENTS goeBreak levation from Elevation from Node Bot of Core (ft.)
" Node" Bottom of Core (ft.)
l 1
0.
11 6.75 2
- 1. 5 12 7.00
)
i 3
3.6 13 7.25 l
4 4.0 14 7.50 f
5 5.0 15 7.75 6
5.5 16 8.0a 7
5.75 17 9.0 8
6.0 1
10.5 I
(
9 6.25 19 12.0 10 6.50
=
" Applicable to the nodes in Figures 15.6.5-4 through 15.6.5-53 6L b
(
4
TABLE 15.6.5-8 Small Break LOCA Time Sequence of Events j
4 in 6 in 8 in (Sec)
(Sec)
(Sec)
Start 0.0 0.0 0.0 Reactor Trip 19.22 12.22 10.41 UHI Accumulator Injection 4
160.
80.
Pumped Injection 50.
45.
40.
Cold Leg Accumulator Injection 1600.
50, 210.
UHI Accumulator Injection Complete 600.
260.
150.
Top of Core Uncovered
- 423, 164.5 74.9-a d
Peak Clad Temperature Occurs 499.16 239.92 1
03 Top of Core Covered 511.
580.
196.3
~
Nt w.
ud ge.e lue nev>
TABLE /5, 6,5 - 7 I
i h
I TABLE 15.6.5-7 Small Break LOCA Time Sequence of Events 1
3 in 4 in-6 in (See)
Mej (See) 0.0 0.0 0.0 Start Reactor Trip 8.57 4.89 2.58 Top of Core Uncovered 1243 658 260 Cold Leg Accumulator Injection 1985 874 368 Peak Clad Temperature Occurs 1637 948 443 Top of Core Covered 2240 1440 473 e
TABLE 15.6.5-9 Saall Break LOCA Results k
4 in 6 in 8 in
.i RESULTS Peak Clad Temperature, 'F 1009 1139.
1218.
Peak Clad Location, Ft 11.25 10.75 11.0
)
J Local Zr/H 0 Reaction,(max)%
.0713 0.0726 0.0773 l
2 l
Local Ir/H O Reaction Location, Ft.
11.25 10.75 11.0 2
Total Zr/H 0 Reaction, %
<0.3
<0.3
<0.3 Hot Rod Burst Time, see Hct Rod Burst Location, Ft.
=
De wre ud Aq kce_ w. %
~
ne-s TM&E
/f. 6. f-8 l
(
i l
\\
e
(
i L
.i
. i TABLE 15.6.5 -
Small Break LOCA Results Fuel Cladding Data 3 in 4 in 6 in RESULTS J
Peak Clad Temperature (*F) 1134 1340 1067 Peak Clad Location (ft) 12 12 12 Local Zr/H O Reaction (max), (%)
0.1037
-0.1436 0.0736 l
2 Local Zr/H O Reaction Location Ft.
12 12 12 2
Total 2r/H O Reaction, (%)
<0.3
<0.3
<0.3 2
Hot Rod Burst Time, (sec)
N/A N/A-N/A -
HotRodBurstLocation,(ft)
N/A N'/A N/A * :-
1
1 l
l l
)
TABLE 15.6.5-12 (Page 1) k.
A.
luation of the potential impact.of using fuel red models presented in raft NUREG-0630 on the Loss of Coolant Accident (LOCA) analysis for Cata a Units 1 and 2.
This eva ation is based on the limiting break LOCA analysis identified as follows.
BREAK TYPE -
UBLE ENDED COLD LEG GUILLOTINE
)
BREAK DISCHARGE EFFICIENT CD = 1.0 PERFECT MIXING WESTINGHOUSE ECCS E LUATION MODEL VERSION 1978 UHI f
CORE PEAKING FACTOR
.32 HOT ROD MAXIMUM TEMPERATU CALCULATED FOR THE BURST REGION OF THE
. CLAD -
1671
'F = PCT B ELEVATION -
6.0 Feet HOT R00 MAXIMUM TEMPERATURE CALCUL TED FOR A NON-RUPTURE 0 REGION OF b
THE CLAD -
2155
'F = PCTN ELEVATION -
5.5 Feet
(
CLAD STRAIN DURING BLOWDOWN AT THI LEVATION 10 PERCENT 10 PERCENT MAXIMUM CLAD STRAIN AT THIS ELEVATIO Maximum temperature for this non-burst node ccurs'when the core reflood rate is greater than 1.0 inen per second and eflood heat transfer is based on the calculation.
AVERAGE HOT ASSEMBLY ROD BURST ELEVATION -
5.
Feet HOT ASSEMBLY BLOCKAGE CALCULATED -
38.6 Perce t 1.
BURST NODE The maximum potential impact on the ruptured clad de is expressed in letter NS-TMA-2174 in_ terms of the change in the aking factor limit (FQ) required to maintain a peak clad temperatu' (PCT) of 2200'F and in terms of a change in PCT at a constant F Since the clad-water reaction rate increases significantly at mperatures above 2200'F, individual effects (such as APCT to changes \\'n several fuel rod models) indicated here may not accurately apply o r large ranges, but a simultaneous change in FQ which causes the PC 'Nto re-main in the neighborhood of 2200*F justifies use of this eval tion procedure.
4 k
b Rev. 5 L/ O,6 YO New Page w
TABLE 15.6.5-12 (Page 2)
Y From NS-TMA-2174:
/
For the Burst Node of the clad:
0.01 AFQ +
- 150*F BURST NODE APCT
.Use'of the NRC' burst model and the revised Westinghouse burst model could require an FQ reduction of 0.027-
,j The maximum estimated impact of using th'~NRC strain
~
e odel is a required FQ reduction of'O 03.
.j Therefore, he maximum penalty for the Hot Rod burst node is:
APCT3=
.027 +.03) (150*F/.01)'= 855'F Margin to the 2 0*F limit is:
APCT2 = 2200* - PCTg=
529
'F The FQ reduction requ ed to. maintain the 2200'F clad tempera-ture limit is:
AFQ * (0
~0
(.01 AFQ) 3 B
1 2
150"F-
)
= (855 - 529) (&)
.022 150 2.
NON-BURST NODE The maximum temperature calculated for non-burst sectio'n of clad occurs durin( the :: ore reflood phase of t e LOCA transient.
The potential impact on that maximum clad temp ature of.using.the NRC fuel rod models can be estimated by examini two aspects of the analyses.
The first aspect is the change in ellet-clad gap con-ductance resulting from a difference in clad s ain at the non-burst maximum clad temperature node elevation.
ote that clad strain all along the fuel rod stops after clad bt st occurs and ae 4
of a different clad burst model can change the tim at which burst is calculated.
Three sets of LOCA analysis results ere studied to establish an acceptable sensitivity to apply gene cally in this evaluation. The possible PCT increase resulting from he. change in strain (in the Hot Rod) is +20. AF per percent decrease strain at the maximum clad temperature locations.
The clad strain alcualted during the reactor coolant system blowdown phase of the a ident is not changed by the use of. NRC fuel rod models.from the " max um clad strain" vnlue of 10 percent as demonstrated below:-
To account explicitly for the impact of NUREG 063'O curves on th non-burst node,:it is necessary to evaluate the effect on hot to burst time as it relates to strain at the PCT-location (non-burst)
Rev. 5 l
New Page h6 b
TABLE 15.6.5-12 (Page 3)
This is done by determining the diff,noce in strain at I
The Cat.awba t when burst is predteced according to the curves. sis cas l
s s the reference point for evaluation of burst temperat f
an 7
This but the strain at the 5.5 ft elevation has reached 10 Hot N l
analysi percent.
pellet temperature margin for Catawba that was no i
There exist
/
certainty on hot rod burst time (-65'F stored energy).
I accounted to of 65"F model ted for in the existing Catawba analysis, hot rod This result If this was acco than that currently calculated.
burst would be lat by application of the sensitivity study of PCT from WCAP-9180.
The above consideration has been establishe From accounts for a total i itial stored energy reduction of 65*F.a red to initial stored ene WCAP-9180 this results j
nalysis exhibits a 6.26 *F/sec heatup The existing Westinghouse
) clad temperature elevation at 6.07 3
rate at a 1543 (maximum val struction of a NUREG-0630 curve leads C
kpsi stress at 6 seconds.
to a burst temperature of 819. 2*C = 1507.5'F; the excess in cal-
/
culated temperature above the 1507.5 = 35.5 F.
uld result in 6 decrease in clad Since the effect of stored energy effect more than compensates temperature of 38 F, the stored ener rst at 6 seconds still is for postulated burst curve changes.
lad strain value of 10 per-not predicted to occur, so the maximum tion.
cent will remain in effect at 5.5 ft ele Therefore:
APCT =(
)
5 AIN - RE G E0 BL M SHAW 3
.01 strain
= ( J ) (.10
.10)
.01
/
=0 the flow The second aspect of the analysis that can increase PCTccurs when t blockage calculated.
Since PCTN
= 0.
The total pot tial is greater than 1.0 inch per second APCT 4 j
PCT increase for the non-burst node is then Rev 5
& CM New Page
~
TABLE 15.6.5-12 (Page 4)
APCT5 = APCT3 + APCT4=0
.i gin to the 2200*F limit is PCT 6 = 2200'F - PCTN = 45' l
The FQ eduction required to maintain this 2200'F clad temperature limit is from NS-TMA-2174)
AFQN = (AP
- APCT ) (
0
)
_1 6
i 10*F APCT AFQg =.045 bu not less,than zero The peaking fact reduction required to maintain the 2200'F clad-i and AFQ '
temperature limit 's therefore the greater of AFQB N
or; AFQPENALTY
.022 t
B.
The NRC has recently approve the removal of 65'F uncertainty on the hot rod fuel pellet temperatur for ECCS analysis. The effect of re-ducing pellet temperature on the alculated PCT has been determined based on previously established s sitivities performed to quantify a
this effect (WCAP-9180).
From thes it is estimated that this re-duction in applied model uncertainty ill result in a decrease in calculated PCT of 25.6'F for Catawba, plying the same sensitivity used in calculating AFQ '
N AFQCREDIT = 25.6 F (.01aF0
) =.0256 10 F APCT C.
The peaking factor limit adjustment required to stify plant operation for this interim period is determined as the appro iate AFQ credit calculated in identified ii Section (B) above, minus the AFQPENALT Section (A) above (bu not greater than zero).
FQ ADJUSTMENT =.0256
.022 > 0.
V UT~ C,/
~)
Rev. 5 New Page l
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1 i
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Delete all Figures from Section 15.6.5 and replace with these.
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SMEAX OCCURE REACTOR TRIP (COW'ENSATLD PRESSJL!ZER PES 3URE) SISNAL PUWpED SAFETY INJECT!DN SIGNAL (H1 1 CONT. PRESS OR LD PREssJt! zen p)
L PM SAFETY INJECTION KOINE ( ASSUWING OUSITE PDeCR AVAILASJ) o w
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Core Power after Reactor Trip
( Applies to all Small Breaks)
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(335/WB 1) MO73 IS Figure 15.6.5-45:
Safety Injection Flowrate vs. Pressure 4
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Figure 15.6.5-46: 3" Cold. Leg Break-RCS Pressure vs. Time L
12.
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I i
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3" Cold Leg Break Hot Spot Clad Temperature vs. Time 1
i l.
p_,_
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fs 1
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Figure 15.6.5-51: 4" Cold Leg Break Hot Spot Clad Temperature vs. Time
[
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sa.
st.
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TIMC ($CCI Figure 15.6.5-53: 6" Cold Leg Break
)
Core Mixture Height vs. Time i
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Q
5620.
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I i
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.1
135 el!
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Figure 15.6.5-56: Core Heat Transfer coef fecient vs. Tir.e
(s;;
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l
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11MC (SCCI 4" Cold 14cr Break Figure 15.6.5-57: Hot Spot ~ Fluid Temperature vs. Time 3
4 i
b.
p 1
ATTACHMENT 5 RADIOLOGICAL ASPECTS OF UHI DELETION i
1 l
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i i
p.
y
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$g i
ATTACHMENT 5
)
RADIOLOGICAL ASPECTS OF UHI DELETION Dose Estimate Manhour Subtask (Person-Rem)
Estimate
- 1) Cut and Cap Auxiliary a) Duke 12.8 32 Head Adaptors.
b) Vendor 5.0 10
- 2) Removal of Piping, 30.8 1650 Valves and Hangers
- 3) Cut and Cap Containment 0.6 200 Penetrations 1
- 4) Exposing Head Adaptors 22.8 700 (i.e., Remove DRPI Coil Stacks, CRDM Ventilation, i
Shield Installation /
Removal, Insulation Removal, etc.)
TOTAL 72 (per 2592 (per unit) unit /.;-
i Estimated general area dose rates are as follows:
j i
a) 2-20 mr/hr along Auxiliary Building UHI piping b) 20-30 mr/hr along lower containment UHI piping c)20-100 mr/hr along reactor head UHI piping j
d) maximum projected dose rates range from 1000 to 1500 mr/hr at the auxiliary
)
head adaptors A " gross estimate" of.the curies of beta and gamma radioactivity contained on the UHI components to be removed is 6.0 curies per unit.
NOTE: The above estimates are based heavily on the McGuire UHI deletion estimates.
More specific Catawba data will be provided at a later date.
i o_