ML20032D545
| ML20032D545 | |
| Person / Time | |
|---|---|
| Site: | North Anna  |
| Issue date: | 11/12/1981 |
| From: | Leasburg R VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.) |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20032D546 | List: |
| References | |
| 627, NUDOCS 8111170242 | |
| Download: ML20032D545 (45) | |
Text
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VIRGINIA ELucTItIC AND POWEH CO)(PANY IIIcit>toNn, VIHOINIA 2152G1 g
p November 12, 1981
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7 ro Mr. Harold R. Denton, Director Serial No.:
627 Office of Nuclear Reactor Regulation FR/KLB: plc Attn:
Mr. Robert A. Clark, Chief Docket Nos.: 50-338 Operating Reactors Branch No. 3 50-339 Division of Licensing License Nos.: NPF-4 U.S. Nuclear Regulatory Commission NPF-7 Washington, D.C.
20555 Gentlemen:
AMENDMENT TO OPERATING LICENSES NPF-4 AND NPF-7 NORTH ANNA POWER STATION UNIT NOS. 1 AND 2 PROPOSED TECHNICAL SPECIFICATION CHANGE Pursuant to 10CFR50.90, the Virginia Electric and Power Company requests an amendment, in the form of changes to the Technical Specifications, to Operating License Nos. NPF-4 and NPF-7 for the North Anna Power Station Unit Nos. 1 and 2.
A LOCA-ECCS reanalysis for Nor;h Anna Unit Nos. 1 and 2 has been performed using the NRC approved February, 1978 version of the Westinghouse LOCA-ECCS Evaluation Model. The analysis has been conducted in compliance with Appendix K to 10CFR50 and meets the acceptance criteria delineated in 10CFR50.46. This analysis was performed by Vepco under supervision of Westinghouse, and the results will support continued full power opers. tion for both North Anna Units at steam generator tube plugging levels of up to 7 pere w. The results of this reanalysis also support a new Fq limit of 2.14 after consideration of current fuel rod burst and blockage penalties. These results are provided in Attachment 1.
Proposed changes to the Technical Specifications, consistent with the reanalysis, are provided in Attachment 2.
This request has been reviewed by the Station Nuclear Safety and Operating Committee an the Safety Evaluation and Control staff.
It has been determined that t request does not involve any unreviewed safety questions as defined "0.59.
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VIMotMIA Etretare 450 Powtm Cowraxy to k'e have evaluated this requert in accordance with the criteria in 10CFR170.22.
It has been determined that this request requires a Class III amendment fee. Accordingly, a voucher check in the amount of
$4400.00 is enclosed in payment of the required fee.
Very t uly yours, I
R. H. Leasburg Attachments (1) LOCA-ECCS Safety Evaluation for North Anna Unit Nos. 1 and 2 (2) Proposed Technical Specifications Changes (3) Voucher Check for $4,400.00 cc:
Mr. James P. O'Reilly, Director Office of Inspection and Enforcement Region II i
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. V COMMONWEALTH OF VIRGINIA )
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CITY OF RICHMOND
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h The foregoing document was acknowledged before nie, in and for the City ind t.
Commonwealth aforesaid, today by R. H. Leasburg, who is Vice President-Nuclear Operations, of the Virginia Electric and Power Company. He is duly authorized.-
to execute and file the foregoing doen.wnt in behalf of that Company, and the statements in the document are true to the best of his knowledge and belief.
Acknowledged before me this /[
day of DWas, 19
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k Attachront 1 LOCA -ECOS Safety Evaluation for i
North Anna Unit Nos. 1 and 2 9
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1.0 INTRODUCTION
A reanalysis of the ECCS performance for the postulated large break Loss of Coolant Accident (LOCA)* has been performed which is in compliance with Appendix K to 10 CFR 50.
The results of this reanalysis are presented herein and are in compliance with 10 CFR 50.46, Acceptance criteria for Emergency Core Cooling Systems for Light Water Reactors.
This analysis was performed with the NRC approved (Ref.
2 11, 12, 13) February 1978 version of the Westinghousa LOCA-ECCS evaluation model.
The analytical techniques used are in full compliance with 10 CFR 50, Appendix K.
3 As required by Appendix K of 10 CFR 50, certain conservative assumptions were made for the LOCA-ECCS analysis.
The assumptions pertain to the conditions of the reactor and associated safety system equipment at the time that the LOCA is assumed to occur and include such items as the core peaking factors, the containment pressure, and the performance of the emergency core cooling system (ECCS).
All assumptions and initial operating conditions used in this reanalysis were the same as those used in the previous LOCA-ECCS analysis (Ref. 3) with the following exceptions:
- 1) the limiting value.of the heat flux hot channel factor was increased from 2.10 to 2.20; 2) more accurate data for several containmenc parameters were used; 3) 7% of the steam generator tubes were assumed to be plugged; 4) the 17 x 17 generic fuel The reanalysis of the small break LOCA is not necessary and therefore the analysis of this accident submitted by Reference 1 remains applicable.
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fuel parameters were updated to reflect the current values, such as removal of the previously required inclusion of a 65'r uncertainty in pellet temperature; 5) a previous requirement of analysis employing a spectrum of fuel heatup rates has been eliminated: 6) a burst and blockage adjustmsnt penalty of 0.06 (as explained in Appendix A) must be subtracted from the value for the heat flux hot channel factor.
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o PAGE 3
2.0 DESCRIPTION
OF POSTULATED MAJOR REACTOR COOLANT PIPE RUPTURE (LOSS OF COOLANT ACCIDENT - LOCA)
A LOCA is time result of a rupture of the reactor coolant systen CRCS) piping or of any line connected to the system.
The system boundaries considered in the LOCA analysis are defined in the FSAR.
Sensitivity studies (Reference 4) have indicated that a double-end cold leg guillotine (DECLG) pipe break is limiting.
In the unlikely event of a DECLG break, a rapid depressurization of the RCS will reruit.
The reactor trip signal subsequently occurs when the pressurizar low pressure trip setpoint is reached.
A safety injection system (SIS) signal is actuated when the appropriate setpoint is reached and the high head safety injection pumps are activated.
The actuation and subsequent activation of the ECCS, which occurs with the SIS signal, assumes the most limiting single failure event.
These countermessures will limit the consequences of the accident in two ways:
1.
Reactor trip and borated water injection complement void formation in causing rapid reduction of power to a l
residual level corresponding to fission product decay I
heat.
(It should be noted, however, that no credit is j
taken in the analysis for the insertion of control rods to l
shut down the reactor).
2.
Injection of borated water provides heat transfer from the core and prevents excessive clad temperatures.
Before the break occurs, the unit is in an equilibrium condition, i.e.,
the i
heat generated in the core is being removed via the secondary system.
During blowdown, heat from decay, hot internals and the vessel continues to be transferred to the reactor coolant system.
At the beginning of the blowdown phase, the entire RCS contains subcooled liquid vhich transfers heat from the core by forced convection with some fully developed nucleate boiling.
After the break develops, the time to departure from nucleate boiling is calcuated, l
PAGE 4
consistent with *.ppendix K of 10 CyR 50.
Thereafter, the core heat transfer is cased on local conditions with transition boiling and forced convection to steam as the major heat transfer mechanisms.
During the refill period,.it is assumed that rod-to-rod radiation is the only core heat transfer mechanism.
The heat-transfer between the reactor coolant system and the' secondary side j
may be in either direction depending on the relative temperatures.
For the case of continued heat addition to the secondary side, secondary side pressure increases and the main safety. valves may actuate to reduce the pressure.
Makeup to the secondary side is automatically provided by the auxiliary feedwater system.
Coincident with the safety injection signal, normal feeduater flov is ctopped by closing the main feedwater control valves and tripping the main feedwater pumps.
Emergency feedwater flow is initiated by starting the auxiliary feedwater pumps.
The secondary side flow aids in the reduction of reactor coolant system pressure.
- ~nen the reactor coolant system depressurizes to 600 psia, the accumulators begin to inject borated' i
unter into the reactor coolant loops.
A conservative assumption is then made j
that the injected accumulator water bypasses the core and goes out through the break until the termination of bypass.
This conservatisn is again consistent with Appendix K of: 10 CyR 50.-
In addition, the reactor coolant pumps are assumed to be tripped at the initiation of the accident and effacts-t of pump constdown are included in the blowdown analysis.
4 The unter injected by the accumulators cools the core and subsequent operation of the low head safety injection pumps supplies water for long term i
cooling.
When the RWST is nearly empty, long term cooling of the core is a;complished by switching to the recirculation mode of cor~e cooling, in which the spilled borated water is drawn from the containment sump by the low head i
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PAGE 5
safety injection pumps and returned to the reactor vessel.
The containment spray system and the recirculation spray system operates to return the containment environment to a subatmospheric pressure.
The large break LOCA transient is divided, for analytical purposes, into three phases bloudown, refill, and reflood.
There re three distinct transients analyzed in each phase, includihg the thermal-hydraulic transient in the RCS, the pressure and temperature transient within the conteinment, and the fuel clad temperature transient of the hottest fuel rod in the core.
Based on these considerations, a system of inter-related computer codes has been developed for the analysis of the'LOCA.
Th2 description of the various aspec,ts of the LOCA analysis methodology is given in WCAP-8339(Ref. 5).
This document describes the major phenomena modeled, the interfaces among the computer codes, and the features of the codes which ensure compliance with 10 CFR 50, Appendix K.
The SATAN-VI, WREFLOOD, COCO, and LOCTA-IV codes, which are used in the LOCA analysis, are described in detail in WCAP-8306 (Ref, 6). WCAP-8326(Raf. 7), WCAP-817)(Ref.
8), and WCAP-8305(Ref.
9), respectively.
These codes are able to assess whether sufficient heat transfer geometry an'd core amenability to cooling are preserved during the time spans applicable to the blowdown, refill, and reflood phases of the LOCA.
The SATAM-VI computer code analyzes the thermal-hydraulic transient in the RCE during blowdown and the COCO computer code is used to calculate the containment pressure transient during all three phkses of the LOCA analysis.
Similarly, the LOCTA-IV computer code is used to compute the thermal transient of the hottest fuel rod during the three phases.
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PAGE 6
SATAM-VI is used to determine the RCS pressure, enthalpy, and density, as well as the mass and energy flow rates in the RCS and steam generator secondary, as a function of time during the blowdown phase of the LOCA.
SATAM-VI also calculates the accumulator mass and pressure and the pipe break mass and energy flow rates that are assumed to be vented to the containment during blowdown.
At the end of the blowdown, the mass and energy release rates during blowdown are transferred to the COCO code for use in the determination of the containment pressure response during this first phase of the LOCA.
Additional SATAM-VI output data from.the end of blowdown, including the core inlet flow rate and enthalpy, the core pressure, and the core power decay transient, are input to the LOCTA-IV code.
With input from the SATAM-VI code, WREFLOOD uses a system thermal-hydraulic model to determine the core flooding rate (i.e.,
the rate at which coolant enters the bottom of the core), the coolant pressure and temperature, and the quench front height during the refill and reflood phases of the LOCA.
WREFLOOD also calculates the mass and energy flow rates that are assumed to be vented to the containment.
Since the mass flow rates to the containment depends upon the core pressure, which is a function of the containment backpressure, the WREFLOOD and COCO codes are interactively linked.
WREFLOOD I
is also linked to the LOCTA-IV code in that thermal-hydraulic parameters from WREFLOOD are used by LOCTA-IV in its calculation of the fuel temperature.
LOCTA-IV is used throughout the analysis of the LOCA tr'ansient to calculate the fuel and clad temperature of the hottest rod in the core.
The input to LOCTA-IV consists o# appropriate thermal-hydraulic output from SATAM-VI and l
WREFLOOD and conservative.'y selected initial RCS operating conditions.
These wr er
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PAGE 7
initial conditions are summarized in Table 1 and Figure 1.
(The axial power j
shape of Figure-1 assumed for LOCTA-IV is a cosine curve which has been previously veri fied t Ref. 10) to be the shape that produces the maximum peak clad temperature).
4 The CCCO code, which is also used throughout the LOCA analysis, calculates the containment pressure.
Input to COCO is obtained from the mass and energy flow rates assum_d to be vented to the containment as calculated by the SATAM-VI and WREFLOOD codes.
In addition, conservatively chosen initial containment conditions and an assumed mode of operation for the containment cooling system are input to COCO.
These initial containment conditions and assumed modes of operation are provided in Table 2.
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1 PAGE 8 I 3.0 DISCUSSION OF SIGNITICANT INPUT Significant differences in input between this analysis and the current 1=
applicable analysis are delineated in Section 1.0 and discussed in more detail below.
The changes made in the analysis reflect the operational conditions and limits necessary to allow full power operation at steam generator tube plugging levels of up to 7%.
The notable chtage for this analysis is the increase in assumed steam generator tube plugging.
The currently applicable analysis allowed for 5%
tube plugging.
This plugging level was increased slightly to 7% for this analysis.
A core inlet temperature of 548.6'T was used in the analysis.
This value was adjusted from operational data to encompass this steam generator tube plugging range.
Several changes were made to the containment parameters.
The thickness of one of the heat sinks in Table 2 was corrected to better represent the as-built plant containment.
In addition, the previous generic value for the high-containment pressure setpoint was lowered to 18.5 psia to agree with the value in the North Anna Technical Specifications.
The calculation was performed assuming conservative generic 17 x 17 fuel parameters consistent with the current methodology.
The previously required 65'T uncertainty in pellet temperature has been removed.
l A previous requirement of analysis employing a spectrum of fuel heatup rates has been removed.
This conforms with the NRC methodology for current ECCS analysis.
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When the above changes were incorporated into the analysis, it was found that the assumed heat flux hot channel factor could be increased from 2.10 to 2.20 i
and still ensure compliance with the 10 CFR 50.46 acceptance criteria.
This allowable increase in the assumed heat flux hot channel factor is primarily the result of the change in the generic fuel parameters, the elimination.of the fuel heatup rate spectrum calculations and the higher peak clad temperature result of this analysis.
A worksheet evaluating the potential impact of using fuel rod models presented in the draft MUREG-0630 is included as Appendix A.
The resulting adjustment penalty of 0.06 must be applied to the overall heat flux hot channel factor, resulting in an adjusted overall heat flux hot channel factor of 2.14.
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PAGE 10 i
4.0 RESULTS and 2 and Figure 1 present the initial conditions and modes of-Tables 1 operation that were assumed in the analysis.
Table 3 presents the time sequence of events and Table 4 presents the results for the double-ended cold leg guillotine break (DECLG) for the CD=0.4 discharge coefficient.
The DECLG has been determined to be the limiting break size and location based on the sensitivity studies reported in Reference 4.
Further, all previous LOCA-ECCS submittals for the North Anna units have resulted in the CD=0.4 discharge coefficient being the limiting break size.
The applicability of this conclusion (i.e. CD=0.4 is the limiting break size) for this analysis was explicitly verified.
Consequently, only the results of the most limiting break size are presented in the figures and remaining tables in this submittal.
The current analysis resulted in a limiting peak clad temperature of 2180.2*T, a maximum local cladding oxidation level of 7.75%, and a total core metal-water reaction of less than 0.3%.
The detailed results of the LOCA reanalysis are provided in Tables 3 through 6 and Figures 2 through 18.
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PAGE 11
5.0 CONCLUSION
S For breaks up to and including the double-ended severance of a reactor coolant pipe and for the operating conditions specified in Table 1 and 2, the Emergency Core Cooling System will meet the Acceptance Criteria as presented in 10 CFR 50.46.
That is 1.
The calculated peak fuel rod clad temperature is below the requirement of 2200*F.
2.
The amount of fuel element cladding that reacts chemically with water or steam does not exceed 1 percent of the total amount of Zircaloy in the reactor.
3.
The clad temperature transient is terminated at a time when the core geometry is still amenable to cooling.
The localized cladding oxidation limits of 17% are not exceeded during or after quenching.
4.
The core remains amenable to cooling during and after the break.
5.
The core temperature is reduced and the long-term decay heat is removed for an extended period of time.
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PAGE 12
6.0 REFERENCES
f.
Final Safety Analysis Report, North Anna Power Station, Units 1 and 2, Virginia Electric and Power Company.
2.
Letter from J.
F.
Stolz(MRC) to T.
M.
Anderson (Westinghouse), dated August 29, 1978.
3.
Letter from C.
M.
StallingsCVepco) to H.
R.
Denton(MRC), Serial No. 986C, January 2, 1980.
4.
Buterbaugh, T.
L.,
- Johnson, W.
J.,
and Kopalic, S.
D.,
" Westinghouse ECCS Plant Sensitivity Studies " WCAP-8356, July 1974.
5.
- Bordelon, T.
M.,
et.
al.,
" Westinghouse ECCS Evaluation Model-Summary,"
WCAP-8339, July, 1974.
6.
- Bordelon, F.
M.,
et.al., "SATAM-VI Program:
Comprehensive Space-Time e
Dependent Analysis of Loss-of-Coolant," WCAP-8306, June 1974.
7.
- Bordelon, T.
M.,
and Murphy, E.
T.,
" Containment Pressure Analysis Code (COCO)," WCAP-8326, June 1974.
8.
- Kelly, R.
D.,
et.
al.,
" Calculation Model for Core Reflooding after a Loss-of-Coolant Accident (WREFLOOD Code) " WCAP-8171, June 1974.
9.
- Bordelon, T.
M.,
et.
al.,
"LOCTA-IV Program Loss-of-Coolant Transient Analysis," WCAP-8305, June 1974.
10.
Letter from C.
M.
Stallings(Vepco) to E.
G.
Case (MRC), Serial No. 092, February 17, 1978.
11.
Eicheldinger C.,
" Westinghouse ECCS Evaluation Model - February 1978 Version," WCAP-9220-P-A (Proprietary Version), WCAP-9221-A (Mon-Proprietary Version), February, 1978.
12.
Letter frem T.
M.
Anderson (Westinghouse) to J.
F.
Stoir(:RC), Serial No.
MS-TMA-1981, November 1,
1978.
PAGE 13 i
13.
Letter from T.
M.
Anderson (Westinghouse) to Rs Tedesco(HRC), Serial No.
MS-TMA-2014, December 11, 1978.
1.
PAGE 14 TABLE 1 IMITIAL CORE CONDITIONS ASSUMED FOR THE DOUBLE-ENDED COLD LEG GUILLOTIME BREAK (DECLG)
CALCULATIONAL IMPUT Core Power (MWt, 102% of) 2775 Peak Linear Power (ku/ft, 102% of) 11.98 2.20 Heat Flux Hot Channel Factor'(Fg)
Enthalpy Rise Hot Channel Factor (F
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1.55 H
Accumulator Water Volume (ft3 each) 1025 Reactor Vessel Upper Head Temperature Equal to Thot LIMITIMG FUEL REGION AND CYCLE CYCLE REGION Unit 1 ALL ALL Regions Unit 2 ALL ALL Regions 6
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CONTAINMENT DATA NET FREE VOLUME 1.916 x 105 ft3 INITIAL CONDITIONS 1 Pressure 9.5 psia Temperature 90*F RWST Temperature 35'r Outside Temperature
-10*F SPRAY SYSTEM 1 Number of pumps Operating 2
Runout Flow Rate (per pump) 2000 gym Time in which spray is effective 59 secs 1
STRUCTURAL HEAT SINKS Thickness (In)
Area (Ft2), w/ uncertainty 6 Concrete 8,393 12 Concrete 62,271 18 concrete 55,365 24 concrete 11,591 27 Concrete 9,404 36 Concrete 3,636
.375 Steel, 54 Concrete 22,039
.375 Steel, 54 Concrete 28,933
.500 Steel, 30 Concrete 25,673 26.4 Concrete,.25 Steel, 120 concrete 12.110
.407 Stainless Steel 10,527
.371 Steel 160,328
.882 Steel 9,894
.059' Steel 60,875 i See the response to Comment 56.106 of the FSAR for a detailed breakdown of the containment heat sinks and for justification of the other input parameters used to calculate containment pressure.
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-o PAGE 16 TABLE 3 TIME SE2UENCE OF EVENTS DECLG CD=0.4 (Sec) 0.0' Start 0.71 Reacter Trip 2.07 S.
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Signal 16.28 Acc. Injection 26.41 End of Bypass 27.07 Pump Injection End of Blowdown 29.60 39.85 Bottom of Core Recovery 52.34 Acc. Empty 9
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'T 2180.2 Peak Clad Location, Ft.
7.5 Local Zr/H20 RXM (max), X 7.75 Local Zr/H20 Location, Tt.
7.5 Total Zr/H2O RXN, %
<0.3 37.00 Hot Rod Burst Time, sec.
Hot Rod Burst Location. Tt.
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PAGE 18 TABLE 5 REFLOOD MASS AND EMERGY RELEASES DECLG (CD= 0.4)
TIME (SEC)
TOTAL MASS TOTAL ENERGY FLOWRATE (LB/SEC)
FLOWRATE (105 BTU /SEC) 39.85 0.0 0.0 40.6 0.785 0.0099 46.3 35.87 0.4675 55.9 221.0 1.421 70.8 253.9 1.439 89.7 266.3 1.406 111.2 274.4 1.361 l
134.9 280.9 1.312 f
189.6 293.2 1.209 257.6 307.2 1.099
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PAGE 19 TABLE 6 BROKEN LOOP ACCUMULATOP. FLOW TO CONTAINMENT DECLG, CD=0.4 TIME (SEC)
MASS FLOWRATE* (LBM/SEC) 4010 0.0 3622 1.0 3104 3.0 2761 5.0
.2509 7.0 2226 10.0 1895 15.0 1674 20.0 1523 25.0 i
1415 30.0
- For energy flourate multiply mass flourate by a constant of 59.60 BTU /L3M.
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APPENDIX A
' A.
Evaluation of the potential impact of using fuel rod models pre-sented in draft hUREG-0630 on the Loss of Coolant Accident (LOCA) analysis for North Anna Unit 1 (VRA).-
Th*
cvaluation is based on 1.he limiting break LOCA analysis identified as follows:
BREAK TYPE - DOUELE ENDED COLD LEG GUILLOTINE BREAK DISCEiRGE COEFFICIENT 0.4 WESTINGHOUSE ECCS EVALUATION MODEL VERSION Februarv '78 model CORE PEAKING FACTOR 2.20 l
HOT ROD MAXIMUM TEMPERATURE CALCULATED FOR THE BURST REGION OF THE CLAD -
1693.91 F = 1 CTB' ELEVATION -
6.00 Feet.
HOT ROD MAXIMUM TEMPERATURi. CLACULATED FOR A NON-RUPTURED REGION OF THE CLAD -
2180.2 F = PCTN
~
ELEVATION -
7.50 Feet l
CLAD STRAIN DURING BLOWDOWN AT THIS ELEVATION 2.47 Percent MAXIMUM CLAD STRAIN AT THIS ELEVATION -
9.50 Percent i
Maximum temperature for this non-burst node occurs when the core reflood rate is i
(LESS) than 1.9 inch per second and reflood heat transfer is based on 4
the (STEAM COOLING) calculation.
AVERAGE HOT ASSEMBLY ROD BURST ELEVATION -
7.25 Feet HOT ASSEMBLY BLOCKAGE CALCULATED -
27.2 Percent
1.
BURST NODE The maximum potential inr act on the ruptured clad node is expressed in letter NS-TMA-2174 n terms of the change in the peaking factor limit (FQ) required to maintain a peak clad temperature (PCT) of 2200 F and in terms of a change in PCT at a constant FQ.
Since the clad-water reaction rate increases significantly at temperatures above 2200. F, individual effects (such as APCT due to changes in several fuel rod models) indicated here may not accurately apply.
over large ranges, but a simultaneous change in FQ which causes the PCT to remain in the neighborhood of 2200 F justifies use of this evaluation procedure.
From NS-TMA-2174:
For the Burst Node of the clad:
- 0.01 AFQ -* a- ~.50 F BURST NODE APCT
- Use of the URC burst model and the revised Westinghouse model could require an PQ reduction of 0.027
- The minimum es*irated impact of using the NRC strain model is a required FQ reduction of 0.03.
Therefore, the maximum penalty for the Hot Rod burst node :.s:
(.027 +.03) (150 F/.01) = 855 F APCT
=
1 Margin to the 2200 F limit is:
2 = 2200.#F - PCTB=
506.09 APCT The FQ reduction required to maintain the 2200 F clad temperature limit is:
AFQB = (APCT1 - APCT )
("
)
2 o
--r, r-e.~,
AFQg = ( 85}.- 506.09)
(
)
= 0.023 (but not less than zero).
2.
NON-BURST NODE The maximum temperature calculated for a non-burst section 'f clad typically occurs at an eleration above the core mid-plane during the core reflood phase of the LOCA transient. The potential impact on that maximum clad temperature of using the NRC fuel rod models can
~
be estimated by examining two aspects of the analyses. The first aspect is the change in pellet-clad gap conductance resulting from a difference in clad strain at the non-burst maximum clad temperature node elevation. Note that clad strain all along the fuel rod stops after clad burst occurs and use of a different clad burst model can change the time at which burst is calculated. Three sets of LOCA analysis results were studied to establish an acceptable sensitivity to apply generically in this evaluation. The possible PCT increase resulting from a change in strain (in the Hot Dod) is +20. F per percent decrease in strain at the maximum clad temperature locations.
Since the clad strain calculated during the reactor coolant system blewdown phase of the accident is not changed by the use of NRC fuel rod models, the maxinum decrease in clad strain that must be considered here is the differet.ee between the " maximum clad strain" and the
" clad strain at the end of RCS blowdown" indicated above.
Therefore:
APCT3=(0 strain)
(MAX STRAIN - BLOWDOWN STRAIN)
=(
0)
(.0950
.0247)
= 140.6 F i
1 l
e
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~
The second aspect of the analysis that can increase PCT is i f ow blockage calculated. Since the greatest value of blockage indicated by the NRC blockage model is 75 percent, the maximum PCT increase can be estimated by assuming that the current level of blockage in the analysis (indicated above) is raised te 75 percent and then applying an appropriate sensitivity formula shown in NS-TMA-2174.
Therefore, 4 = 1.25 F (50 - PERCENT CURRENT BLOCKAGE)
APCT
+2.36 F (75-50)
= 1.25 (50 - 27.2) + 2.36 (75-50)
= 87.5 F If PCTN occurs when the core reficod rate is greater than 1.0 inch per second APCT4 = 0.
The total potential PCT increase for the non burst node is then APCT5 = APCT3 + APCT4 = 228.1 F Margin to the 2200 F limit is APCT6 = 2200 F - PCTN = 19.8 F The FQ reduction required to maintain this 2200 F clad temperature limit is (from NS-TMA-2174)
AFQN= (APCT5 - APCT )
([ F PCT}
6 AFQN = 0.208 but not less than zero.
The peaking factor reduction required to maintain the 2200*F clad temperature limit is +.herefore the greater of AFQB and AFQNe
- 3 0 OPETA'EY = 0.21
~
o.
B.
The effect on LOCA analysis results of using improved analytical and modeling techniques (which are currently approved for use in the Upper Head Injection plant LOCA analyses) in the reactor coolant system blowdown calculatien (SATAN computer code) has been quanti-fied via an analysis which has : x;stly been submitted to the NRC for review. Recognizing that re' w or that analysis is not yet complete and that the benefits associated with those model improve-ments can change for other plant designs, the NRC has established a credit that is acceptable for this interim period to help offset penalties resulting from application of the NRC fuel rod models'.
That credit for two, three and four loop plants is an increase in the LOCA peaking factor limit of 0.12, 0.15 and 0.20 respectively.
C.
The peaking factor limit adjustment required to justify plant opera-tion for this interim period is determined as che appropriate AFQ credit identified in section (B) above, minus the AFQ
- 81~
PEN /~TY culated in section (A) above (but not greater than zero).
FQ ADJUSTMENT = 0.15 - 0.21
- -0.06 4
e D
a.