ML20094N811
| ML20094N811 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 03/31/1992 |
| From: | John Marshall TEXAS UTILITIES ELECTRIC CO. (TU ELECTRIC) |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| TXX-92173, NUDOCS 9204070089 | |
| Download: ML20094N811 (8) | |
Text
.
4 lL".If503 Log Il TXX-92173 OC File # 10010 1-.
915
~
7UELECTRIC March 31, 1992 William J. Cahill, Jr.
Group Yke Presdent U. S. Nuclear Regulatory Commission Attn: Document Eentrol Desk Washington D. C.
20555
SUBJECT:
COMANCHE PEAK STEAM ELECTRIC STATION (CPSES)
DOCKET NOS. 50-445 AND 50-446 REQUEST FOR ADDITIONAL INFORMATION ON RXE-91-002
" REACT IVITY Atl0MALY EVEllTS MI "iODOLOGY" REF:
1)
Letter from W. J. Cahill, Ji to the NRC logg;d TXX 92093 dated, February 14, 1992.
2)
Letter f rom the NRC to Mr. William J. Cahill, Jr. dated January 14, 1992, Requesting Additional Information regarding lopical Report RXE-91-002.
Gentlemen:
Reference 1 transmitted TU Electric's response to 26 of the 28 questions provided in Reference 2.
Attached, please find TV Electri:'s responses to the two remaining questions provided in Reference 2.
Should clarification or additional information regarding responses to the referenced letter be required to enable the Stt f f to complete its review, contact Mr. Jimmy D. Seawright at 214-812-4375.
Since ely, William J. Cahili, Jr.
h,p
&b{
J. S. Marshall Generic Licensing Manager JDS/gj Attachment c - Mr.
R.D.Me3, Region IV
]
Resident Inspectors CPSES (2)
Mr, T. A. B e r g"Ta n. NRR Mr. B. E. Hol l a n, NRR g
,r i
)
I 9204070089 920331 PDR ADOCK 0500044S
',/
1 P
PDR 400 N. Olive Street LD bl Dallas,Teus 752nl
Attachment to TXX-92173 page 1 of 7 0
4 TOPICAL REPORT RXE-91-002 REACTIVITY ANOMALY EVENTS METHODOLOGY Note:
The references, figures, tables, and nomenclature quotmi in this response correspond to those provided in Topical Report
~
RXE-91-002.
Q_ufJt i oll 5
24.
u i
Provide the nothodology, predictions and sensitivity studies for the control rod ejection DNBR analyses.
Ans. WAE Note:
The tables within the text of this response that are not found within RXE-91-002 are identified by alphabetic character, and are located at the end of the response to this question.
The core T-H model is used to evaluate each CRE event scenario with respect to DNB.
A statepoint analysis is performed using boundary conditions for core power, system pressure, core inlet temperature, and core inlet flow rate from the system T-H analysis.
The system boundary conditions selected for use in the statepoint analysis correspond to the most limiting event conditions with respect to DNB.
A chopped cosine axial power shape is used unless the hot channel axial power shape obtained from the core physics calculations is more limiting.
_,___,e--------
~ ~ ~ ' ~ ' ~ ~ ~ '
w Attachment to TXX-92173 Page 2 of 7 I
The DilB analysis results for each case are evaluated against the appropriate base case by utilizing the relationship of the MDtJBR to the product of the F and the core pow 2r.
3n This relationship is unique for a given combination of temperature, pressure, flow rate, and axial power shape, i.e., a given base case.
The mathematical relationship is given in Equation 4.3-1, and is restated below for completeness.
(Posi,) * (fin) aH 11M psr where:
P Core power at a statopoint s7
=
P DlIBR limited core power at F$n, m
=
T and g,
PR r 3
F$u Design limit hot channel peaking factol.-
=
T Core inlet temperature at a statepoint n
=
prs 7 RCS pressure at a statepoint
=
F Dt.tR limited hot enannel peaking factor
=
an.uu at Pu For demonstration purposes, the following conditions worn used in the core T-H analyses to establish the base cases F$n
=
. 55 T
565.5*F for HFP u
=
557'F for HZP PR 2208 psia
=
n D11BR DtiBR design limit = 1.35
=
uu m
100%, Thermal Design Flow for HFP
=
cou.
46% of Thermal Design Flow for HZP The pressure is held constant throughout the core T-H analysic. bt ause the initial value is the most restrictive
Attachment to TXX-92173 Page 3 of 7 with respect to DNB.
The core. inlet temperature remains constant throughout the core T-H analysis because the time j
to reach MDNBR ir much less than the loop transit time.
The core inlet flow ate is maintained constant at the minimum expected flow rate for the scenario of interest.
The core T-H analysis is performed using the above stated conditions, in conjunction with a limiting axial power shape, to douermine thn DNBR limited core power (Pmn) for the ccre exposure and initial power level of interest, i.e.,
the base case.
The peak core average thermal power (Pn) for a given event scenario is then used to determine the DNBR limited hot channel peaking factor (Fanu).
The Fauu is used to perform a fuel pin census to determine the number of fuel pins which experience DNB.
Any fuel pin having an augmented Fai greater than or equal to Fauau is assumed to experience DNB.
The peak core average thermal power from the system T-H analysis and the results of the core T-H analysis / fuel pin census for the cases presented in Section 5.5 are provided in Table A.
The corresponding results for the RXE-91-002 Appendix B sensitivity studies are provided in Tables B through D.
TABLE A Control Rod Ejection DNBR Results h
Case Pea' Thermal Power Pins in DNB 4'
(MW)
% Total HFP BOC 3962 0.01 HFP EOC 3902 0.02 HZP BOC-696 0.3 HZP EOC 1013 4.1
1 TABLE B
.3
^~
.HFP Neutronics Parameters Sensitivity' Study DNBR Results
~m>.
$N ea o
Pe?k P1ns Peak Pins A:7 Thermal in DNB Thermal in DNB o$
Power Power Parameter Variation MW
% Totql MW
% Total-'
4-tt i
O.
- Nominal 4032 0.01 4015-0.02 4
-x DTC
-10%
4057 0.0~
4032-0.02-Y'
+10%
4009
-0.01 4001 0.02 MTC, pcm/*F
+3 4058 0.01 4029 0.02 d
-3 4008 0.01 4004 0.02 f*,
psec
+5 4032 G.01 4011 0.02
-5 4031 0.01 4014 0.02 l
S,g
-5%
4047 0.01 4016 0.02
+5%
4017 0.01 4015 0.02-Ejected Control Rod' Worth
+10%
4089 0.01 4075 0.16
-10%
3976 0.01 3961 0.02 Ejection. Time Doubled 4030 0.01 4013 0.02 Halved 4G32 0.01 4016 0.02 3
Reactor Trip Delay'Ti.me, seconds
+0.5 4078 0.01 4015 0.02
-0.5 3962 0.01 4016-0.02 1
Scran Worth
+10%
4030 0.01 4015 0.02
-10%
4034 0.01 4015 0.02 l
Time Step Size, seconds.
0.01 4035 0.01 401G 0.02 Reload-S's & A's 4039 0.01 4020 0.02 l
i l
3 s
= _ _ - ~, _
TABLE C I
HZP.Neutronics Parameters Sensitivity Study DNBR Results
~m>
,$ N oao Peak Pins Peak Pins in s' Thermal in DNB Thermal in DNB o$
Power
. Power
. Parameter Variation MW
% Total
__ MW
% Total-4, o
Nominal 680 0.1 1009 4.1 gx DTC
~10%
765 1.3 1138 7.2-Y
+10%
612 0.1 907 2.9 U
MTC, pca/*F
+3 732 0.6 101L 5.7 o
-3 637 0.1 1003 4.1
+5 680 0.1 999 4.1
(*,
ysec
-5.
681 0.1 1030 5.7
-St 715 0.4 1047 5.7
- g
+5%
644 0.1 971 4.1 Ejected Control Rod Worth
+10%
813 2.1 1203 7.2
-10%
539 0.1 821 0.8 Ejection Time Doubled 680 0.1 1009 4.1 Halved 680 0.1 1009 4.1 Reactor Trip Delay Time, seconds
+0.5 700 0.3 1009 4.1.
-0.5 655 0.1 1009 4.1 Scram Worth
+10%
678 0.1 1009 4.1
(-5% for EOC HZP)
-10%
682 9.2 1009 4.1 Tiue Step Size, seconds 0.01 698 0.3 692 0.1-Reload S's & X's 684 0.2 1010 4.1 r -
.i TABLE D Thermal-Hydrau'lic Parameters Sensitivity Study DNBR Results
- *o > -
^oa Peak Pins Pcak Pins e fr.
Thermal in DNB Thermal in DNB:
gg Power' Power-m :s i
Parameter Variation
_ MW
% Total MW
% Total a e
+
~ Nominal 4032 0.01 4015 0.02 i
Hx Fuel Temperature (U
+50%
4081 0.01' 4058 0.10 7[
-50%
3931 0.01 3955 0.02.
o w
Fuel Pellet, / mesh pts.
10 4028 0.01 4011 0.02, P
j Inlet Temperature,
- F
+5.5 4032' O.01-4017 0.02
-5.5 4031 0.01 4014 0.02 RCS Pressure, psi
+30 4031 0.01 N/A N/A
-30 4032 0.01 N/A N/A t
i
{
Nominal 680 0.1 1009 4.1' l
t Fuel Temperature (U
+50%
753 0.9 1234 7.2 l
-50%
533 0.1 668 0.1 Fuel Peller, # eesh pts.
10 678 0.1 976 4.1 l
Inlet Temperature,
'F
+5.5 682 0.2 1013 5.7
-5.5 678 0.1 1005 4.1 I
i l
RCS Pressure, psi
+30 680 0.1 1009 4.1
(
-30 680 0.1 1009 4.1 l
(U Fuel rod gap conductance used for variance j
l l
1 i
I
_m
__... ~ _.._ _. _ _..
Attachment to TXX-92173 Page 7-of:7'
- 25. 'Ouestion
~In the rod ejection accident analysis, the use of a film boiling heat transfor correlation is conservative for fuel enthalpy calculations, but is nonconservative for heat flux predictions in'DNBR analyses.
How is the heat transfer calculation perforred in the DNBR analysis?
Answer The hot spot model uses a film boiling heat transfer correlation for the fuel enthalpy calculations.
The hot spot model is not used for the DNB analyses.
The heat transfer calculation for-the DNB analysis is baced on the core average thermal power response of the system T-H analysis.. The system T-H model uses a subcooled nucleate boiling correlation to predict the fuel pin heat flux.
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