ML20114D681
| ML20114D681 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 09/03/1992 |
| From: | TEXAS UTILITIES ELECTRIC CO. (TU ELECTRIC) |
| To: | |
| Shared Package | |
| ML20114D680 | List: |
| References | |
| NUDOCS 9209090250 | |
| Download: ML20114D681 (49) | |
Text
. . .
MARKED-UP TEEHNICAL SPEClflCAT10NS PAGE5 (NUREG 1399)
PAGE f;-25 ATTACHMENT 3 TO TXX-92323 PAGE 1 0F 3 9209090250 DR 920903 ADOCK 05000443 PDR
A act m t 3 to Txx-92323 .
EMINISTRATIVE CONTROLS 6
WCAP-10216-P-A, " RELAXATION OF CONSTANT AXIAL OFFS:T 6 i CONTROL FQ SURVE!LLANCE TECHNICAL SPECIFICATION", June 1983 I (W Proprietary).
(Methodolo Factor (W(gy for Specification
- 2) surveillance 3.2.2-Hee.
requireents for FQFlux Hot Channel 6 Methodology).)
~n -
b ele.tc % WCAP-8200, "WFLASH, A FORTRAN-IV COMPUTER PROGRAM FOR 6 SIMULATION OF TRANSIENTS IN A MULTI-LOOP PWR." Revision 2.
June 1974 (W Proprietary).
(Methodology for Spacification 3.2.2 - Heat Flux Hot 6 ChannelFactor.) _ _
WCAP-9220-P-A, " Westinghouse ECCS Evaluation Model, 6 February 1978 Version," February 1978 (W Proprietary).
(Methodology for Specification 3.2.2 - Heat Flux Hot 6 Channel Factor.)
='
The core operating limits shall be determined so that all applicable 6
lin.its (e.g., fuel thermal-mechanical limits, core thermal-hydraul'.c limits, ECCS limits, nucicar limits such as shutdown margin and transient and accident analysis limits) of the safety analysis are met.
The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or 6 supplements thereto, shall be provided upon issuance, for each reload cycle, to the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector.
SPECIAL REPORli 6.9.2 In addition to the applicable reporting requirement of Title 10, Code of Federal Regulations, spef.ial reports shall be submitted to the Regional Administrator of the Regionai Office of the NRC within the time period specified for each report.
COMANCHE fCAK - UNIT 1- 6-25 AMENDMENT 6 NOVEMBER 27, 1991
Attachment 3 to TXX-92323 Ppge 3 pf-3 g _ g,gg E
l INSERT A WCAP 10079 P A, 'NOTRtmP, A N00AL TRANSIENT SMALL BREAK An u GENERAL NE1VORK CODE " August 1985 (M Proprietary). F (Methodology for Specification 3.2.2 - Heat Flux Hot Channel el Factor.)
WCAP 10054 P4 , 'VESTINGHOUSE SMALL BREAK ECCS EVALUATION MODEL USING THE NOTRUMP CODE *, August 1985, M Proprietary).
(Methodology for Specification 3.2.2 - Heat Flux Hot Channel Factor).
WCAP-11145 P A,
- WESTINGHOUSE SMALL BREAK LOCA ECCS EVALUATION MODEL GENERIC STUDY V?% THE NOTRUMP CODE", October 1986, y Proprietary).
(Methodology for Specification 3.2.2 - Heat Flux Hot Channel Factor). _
WESTINGHOUSE LETTER, WPT-14670 TV ELECTRIC COMPANY COMANCHE PEAK STEAM ELECTRIC STATION UNIT NU BER 1 SMALL BREAK LOCA USING W FLASH PEAK CLAD TEMPERATURE (PCT)
JULY 13,1992 ENCLOSURE i TO TXX-92323
~
(TOTAL T GES - 3) s
WT-14670
% ET-NSL-0PL-II-92-ci9 Westinghouse Energy Systems Flectric Corporation sa 355
"'"stu*gh Penessvan:a 15230 0355 July 13, 1992 Mr. W. J. Cahill, Jr., Executive Vice President t'uclear Engineering & Operations S.O. No. TBX-4708 TU Electric Company 1
- 0. Box 1002 4 n Rose, Texas 76043 Ref: 1. WPT-13635 L 2. WPT-14479 ention: W. Choe (No Response Required)
TV ELECTRIC COMPANY COMANCHE PEAK STEAM ELECTRIC STATION UNIT NUMBER 1 S$ LL BREAK LOCA USING W-FLASH PEAK CLAD TEMPE
Dear Mr. Cahill:
As recently discussed with Mr. Whee Choe of your organization, the curre Comanche Peak Unit open issues previously I small provided break LOCA via Reference 1. analysis has been evaluated Our investigation of PI-91-005, "Small Break LOCA Burst and Blockege Considerations", ha in a Peak Clad Temperatun (PCT) over the 22004 criteria of 10CFR50.46 when applied to the original Unit 1 analysis which utilizes the W-FLASH ,
be reportable to the NRC as we have recently sup
- 2) an engineering assessment based on application of the NOTRUMP Sm LOCA methodology which demonstrates compliance with the 2200'F criteria .
Please find attached the results of our evaluation in this matter.
l if there Roy Owocare any questions, on 412/374-4037. please contact Craig Thompson on 412/374-4409 or Very truly yours, C.< -W U fL J. L. Vota, Manager Comanche Peak Projects R. H. Owoc Attachment
WPT-14670 ET-NSL-0PL-II-92-319 COMANCHE PEAK STEAM ELECTRIC STATION UNIT SMALL BREAK LOCA LICENSING BASIS PEAK CLA OVER THE 2200'F 10CFR50.46 CRITERIA INTRODUCTION 1
.annual Westinghouse provided reporting _ requirement TU Electric with text (Reference 1) as pa of 10CFR50.46.
Attachment 2 to Reference (1)
' were considered.to the Westinghouse ECCS Evaluation model. be too new to require repo The investigation for "SMALL BREAK
' LOCA BURST AND BLOCKAGE CONSIDERATIONS", with conside
- limiting time in life, has been completed, and Westinghouse has de that this concern, when applied to CPSES-11 current licensing bas -
criteriao)f100FR50.46.SBLOCA EM , results in a Peak Cladding Tem reportable to the NRC as a substantial safety hazard, s recently supplied TU Electric with an engineering assessment (Re based on application of.the NOTRUMP Small Break LOCA evaluatio CPSES-li which demonstrated compliance with the 2200*F criteria .
ILCHNICAL DISCUSSION discussion, an abbreviated discussion is provided pointed out that the small break'LOCA burst model.could go-outside theR high pressure rods or failure to predict burst wh pressure rods.
When this-condition was corrected, burst occurred for lower pressure _ rods at. higher temperatures, which caused the Zirconium water reaction to
-Temperatures. occur very rapidly, leading to higher calculated Peak C W-F' ASH Small Break LOCA Evaluation Model a psig backfill pressure at beginning of life cond the corrected burst model, an increase in PCT occurred.
combined with the series of: previous 10CFR50.59 safety evaluations w in-PCT above the 2200'F criteria for the CPSES-1 s licensing basis.
-CONCLUSl_03 model, has been. performed for CPSES-1.A new small break.L This new analysis, using all changes showing a large : amount of margin to the 10CFR5 This result indicates that NOTRUMP calculates improved core coolin .
3, .
I-
WPT-14670 ET-NSL-0PL-II-92-319 compared to the older W-FLASH model.
evaluated to have a PCT over the 2200*F criteria, the NOTR ameliorates 10CFR50.46 any criteria. concern with regard to safe operation or for com
{
quality assurance program and is therefore cons licensing basis analysis.
l l
17x17 Standard analysis. Fuel THRIVE data with the values used !
Differences core regions. were limited to pressure drops in the downcomer a:
Because the values for pressure drops used in the NOTRUMP analysis are higher in the downcomer and only 0.1% lower in the core
, these differences would have negligible effect on the calculated PCT .
REFERENCES 1.
WPT-13635, J. L. Vota (H) to Mr. W. J. Cahill, Jr. TV Steam Electric Station, ECCS Evaluation Model Change " Comanche(s"),
Peak 2.
, , 1991, June 20 Steam Electric Station Unit Number 1, NOTR s-Engineering Assessment in Support of Continued Operation", , 1992. April i
WESTINGHOUSE LETTER. WPT-14387 TV ELECTRIC COMPANY COMANCHE PEAK STEAM ELECTRIC STATION UNIT 2 SMALL BREAK LOCA ECCS REANALYSIS FEBRUARY 26, 1992 ENCLOSURE 2 TO TXX-92323
-(TOTAL PAGES - 28)
I:
1
WPT-14387 QIT \
Westinghouse Energy Systems Electric Corporation $$ mnmana tmc cns February 26, 1992 ET-NSL-0PL-II-92-102 Mr. W. J. Cahill, Jr., Executive Vice President Nuclear Engineering & Operations 'i.0. No. TCX-4708 TV Electric Company Ref: 1) CPSES-9122081 P. O. Box 1002 2) WPT-14131 Glen Rose, Texas 76043
- 3) CPSES-9130542 Attention: A. Tajbakhsh No Response Required TU ELECTRIC COMPANY COMANCHE PEAK STEAM ELECTRIC STATION UNIT 2 SMAtt BREAK LOCA ECCS REANALYSIS
Dear Mr. Cahill:
In response to your request of Reference 1, and in accordance with our commitment of Reference 2, please find attached the Comanche Peak Unit 2 Small Break LOCA ECCS reanalysis.
The Small Break LOCA ECCS analysis for Unit 2 was performed at a core power level of 3411 Hwt. Other pertinent analysis assumptions include 5 per :nt steam generator tube plugging level, 17x17 Optimized Fuel Assembly (0FA) fuel design, and 27S psig fuel rod helium backfill pressure. The-analysis was performed with the NRC-approved Westinghouse ECCS Small Break Evaluation Model which utilizes NOTRUMP and is described in WCAP-10081-A.
Some of the assumptions used in the analysis were provided by TV Electric in the partially filled out Accident Analysis Checklist (ACC) per above Reference
- 3. TV Electric should assure that these assumptions remain valid.
The Unit 2 Small Break LOCA section FSAR updates are pros :ded in Attachment A which include the results of the 2 , 3 , and 4-inch break analysis. The 3-inch
-~+ .'
.c WP1-1A387- i Mr. W. J. Cahill,-Jr.- .
Page 2:
February 26, 1992
-is the worst case break resulting in a Peak Clad Temperature (PCT) of 1434*F. These results demonstrate-conformance with 10 CFR 50.46 requirements for Small Break LOCA ECCS Analysis for Comanche Peak Unit 2. i This. closes ~open; item 10179-3.
If'there are any questions on the above, please contact Mr, Roy Owoc at 412/374-4037.
Very truly yours, WESTINGHOUSE ELECTRIC CORPORATION RH0/lgi C-J. L. Vota, Manager
& Y fst-Comanche Peak Projects
-Attachment cc: W.J.hCahill,Jr. .
IL,- 1A
-CCS . IL, lA, I AR
-S.'C. Wood- _
IL, IA VETIP Coordinator- -IL, IA L. Terry =. . IL, lA
- J. B. Roberts - I L ', - 1 A
-T.iA.-Hope IL, IA
-W; G. Guldemond : ll,-- 1 A W. Choe- IL, IA A. Tajbakhsh- IL,'1A D.-Bize. IL, IA
'A
, y , - ~ "
g 4 ATTACHMENT A Comanche Peak Steam Electric Station Unit No. 2 .
FSAR Updates for Small Break LOCA
7, - .
CPSES-FSAR 15.6.5.3 CORE AND SYSTEM PERFORMANCE 15.6.5.3.1 MATHEMATICAL MODEL Small Break LOCA Evaluation Model The Coiaanche Peak Steam Electric Station Unit No. 2 small break LOCA analysis was perfonned using the Westinghouse ECCS_ Small Break Evaluation modell4 which utilizes the NOTRUMP12,13 and LOCTA-IV8 computer codes. These computer codes are used to perform the analysis of Loss-Of-Coolant Accidents due to small breaks in the Reactor Coolant System (RCS). The NOTRUMP computer code, approved for this use by the Nuclear Regulatory Commission (NRC), is used to calculate the transient depressurization of the RCS as well as to describe the mass and enthalpy of the flow through the reactor core and break. This code is a state-of-the-art one-dimensional general network code incorporating a number of advanced features.- Ar.ong these new features are the' utilization of nonequilibrium thermal calculation in all fluid volumes, flow regine-dependent drift flux calculations with counter-current flooding limitatiuns, mixture level-tracking logic in multiple-stack fluid nodes and-regime-dependent heat transfer correlations. The NOTRUMP small break LOCA emergency core cooling system evaluation model was developed to
' determine the RCS response to-design basis small break LOCAs and to address.the NRC concerns expressed in NUREG-0611, " Generic Evaluation of Feedwater Transients and Small Break Loss-of-Coolant Accidents in l Westinghouse-Designed Operating Plants."
j' .In NOTRUMP12,13, the RCS is subdivided into fluid filled control volumes l-
.(fluid nodes) and metal nodes interconnected by flowpaths and heat transfer llaks. The transient behavior of the system is determined from j the governing conservation: equations of mass, energy, and momentum applied to these nodes. The_ broken loop is modeled explicitly, and the intact loopsDare lumped into a second loop. - A detailed description of the NOTRUMP code -is provided in References 12 and 13.
In the NOTRUMP model l4, the reactor core is represented as a vertichi l_ stack of heated control volumes with an associated bubble rise model to
CPSES-FSAR permit a transient mixture height calculatinn. The multi-node capability l
of the program enables the explicit and 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 accident.
Clad thermal analysis are performed with the LOCTA-IV, Reference 8, computer code which uses as input 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. For all computations, the H0 TRUMP and LOCTA-IV calculations were terminated slightly after the time the core mixture level returned to the top of the core following core uncovery.
A schematic representation of the computer code interfaces is given in Figures 15.6-5 and 15.6-6.
15.6.5.3.3 RESULTS Small Break Results As noted previously, the calculated peak clad temperature resulting from a small break LOCA is less than calculated for a large break. A range of small break analyses are presented which establishes the limiting break size as-3 inches. The results 'of these analyses are sumarized in Tables 15.6-1 and 15.6-7.
Figures 15.6-34 through 15.6-47 present the principal parameters of interest for the small break ECCS analyses. For all cases analyzed the following transient parameters are presented:
- a. RCS pressure. (Figure 15.6-34,15.6-41,15.6-42)
- b. Core mixture height. (Figure 15.6-35, 15.6-43, 15.6-44)
- c. Hot spot clad ^,emperature. (Figure 15.6-36, 15.6-45, 15.6-46)
cU '~
CPSES-FSAR t
d.-Cora Power after trip. (Figure 15.6-37) e.-Pumped-safety injection. (Figure 15.6-47)
For the limiting 3 inch break, the following additional transient parameters are presented, a.. Core steam flow rate. (Figure 15.6-38)
.b. Core heat transfer coefficient. (Figure 15.6-39)
- c. Hot- spot fluid temperature.- (Figure 15.6-40)
Peak clad temperature for the limiting break (3-inch)-was 1433.8'F. The e
maximum local zirconium oxidation was 0.60% and the core wide oxidation was less than the 1% critoria. These results indicate that a-coolable geometry was maintained for small break LOCAs and therefore, long-term core cooling is assured by continued operation of the ECCS. These results
- are_wsil below all Acceptance Criteri3 limits of 10CFR50.46 and in all cases are not limiting when compared to the results presented for large
- breaks.-
I
< CPSES-FSAR TABLE 15.6-1 (sheet 3 of 4)
TIME SEQUENCE OF EVENTS FOR INCIDENTS WHICH DECREASE IN REACTOR COOLANT INVENTORY Accident fJutal Time hect 3.
DECLG CD - 0.4 Start (Min SI) 0.0 Reactor trip signal 0.53 Safety injection signal 1.62 Accumulator injection begins 19.6 End-of-bypiss 35.73 End-of-blowdown 35.73 Pump injection begins 26.62
-Bottom of core recovery 48.25 Accumulctor empty 54.08 Small break LOCA
-1. 2 inch Start 0.0 Reactor trip signal 62.9 Safety injection signal 73.9 Top of core uncovered
-2381.2-Accumulator injection begins N/A Peak clad temperature occurs 4062.6
- Top of core covered 5512.5 2.: 3 inch -Start 0.0 Reactor trip signal- 21'.6 Safety injection signal- -31.6 Top of core uncovered 990.5 Accumulator injection begins 1999.8 Peak clad temperature occurs 1841.8 Top of core covered 3263.9
~, ,,_
T M CPSES-FSAR
. TABLE 15.6-1
-(sheet 4 of 4)-
-TIME SEQUENCE OF EVENTS FOR INCIDENTS WHICH CAUSE A
- DECREASE IN REACTOR COOLANT INVENTORY
' accident Event Time (sec)
- 3. 4Linch ' Start 0.0 Reactor trip signal 12.7 Safety injection signal 21.6' Top of co're uncovered 623.5
' Accumulator injection begins 8S7.6 Peak clad temperature occurs 348.0
-Top of core covered 1342.2 c ._
'b s-I'
[ ..
L
.3t u
, . CPSES-FSAR TABLE 15.6-5 INPUT PARAMETERS USED IN THE ECCS ANALYSIS Licensed core power (a) , (MWt) 3411 Peak linear power, includes 102 % factor (KW/ft) 12.87 Total peaking factor, Fg 2.32(b)
Axial peaking factor, F 7 1.497 Power shape large break Chopped cosine Small break See Figure 15.6-48 Fuel assembly array Optimized 17x17 Accumulator water volume, nominal (ft 3/accum) 850 3
Accumulator tank volume, nominal (ft /accum) 1350 Accumulator gas pressure, minimum (psia) 600 Safety injection pumped flow See Figures 15.6-21 and 15.6-47 Containment parameters See Sec 6.2 Initial loop flow (lb/sec) 9868 Large Small Vessel inlet temperature (OF) 558.3 564.1 Vessel outlet temperaturs (CF) 618.7 623.3 Average reactor coolant pressure (psia) 2280 2280 Steam pressure (psia) 994.7 1000 Steam generator tube plugging level (%) 0 5 (a) Two percent is to be added to this power to account for calori.netric error.
(b) " Envelope" for small break.
E
-.-- . CPSES-FSAR TABLE 15.6-7 SMALL BREAK LOCA RESULTS FUEL CLADDING DATA Results 2_ inch 3 inch 4 inch Peak clad temperature (OF ) 1005.3 1433.8 1290.9 Peak clad temperature location (ft) 11.5 11.75 11.5 Local Zr/H 2O reaction, maximum (%) 0.05 0.60 0.11 Local Zr/H2 O reaction location (ft) 11.5 11.75 11.5 Total .
/H2O reaction (%) <l.0 <1.0 <l.0 Hot rod burst time (sec) N/A N/A N/A H
rod burst location (ft) N/A N/A N/A l
l-l-
z, _,.
.CPSES-FSAR. >
- i '
15.
6.7 REFERENCES
r
- 12. Meyer,3P. E., "NOTRUMP, A Nodal Transient Small Break and General
_ Network Code," WCAP-100079-P-A (Proprietary), and WCAP-10080-P-A-(Non-Proprietary), August 1985.
- 13. Rupprecht, S. D., et al, " Westinghouse Small-Break LOCA ECCS Evaluation Model Generic: Study with the NOTRUMP Code,"
WCAP-11145-P-A (Proprietary), and WCAP-ll373-A (Non-Proprietary),
1; October 1986.-
- 14. Lee,' N., et al, " Westinghouse Small Break LOCA ECCS Evaluation Model using the NOTRUMP Code," WCAP-10054-P-A (Proprietary), and WCAP-10081-A'(Non-Proprietary), August 1985.
I-j E
l- . . -
N L
0 0
T
-_ C -
CORE PRESSURE, CORE T A
y FLOW.WIXTURE LEVEL p MC FUEL RCO POWER HISTORY l
o< TIME < CORE COVERED l
l^
U m I-l-
COMANCHE PEAK S.E.S.
FINAL SAFETY ANALYSIS REPORT i UNIT 2 IT Code Interface Description for Small Break Model FIGURE 15.6-6
24ss. ---
2299.
2988.
tees. .
31888.
E w 1488 I. L 'N g tam.
N 9 18s6. N BAS. ,
'88
- 3. see, ines. isas asse, asas sees. ssee me.
TIME (SEC)
COMANCHE PEAK S.E.S.
FINAL SAFETY ANALYSIS REPORT UNIT 2 RCS Depressurization Transient (3 Inch Brekk)
FIGURE 15.6-34
_ _ _ _ _ _ _ _ _ _ - - - - - - _ _ - _ _ _- - - _ - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - ~
52.,
i
\ i
\
53.
f ; -
I I
- 29. I---\
! )
n 26.i- n i w
a
- 24 W
TOP OF CORE '
22."------- ----- ------
- a. r /
( /-
L I
B. 588. 1980. 1598. 2998. 2539. 5898. 5500. ~4000.
TIE (SEC)
COMANCHE PEAK S.E.S.
FINAL SAFETY ANALYSIS REPORT UNIT 2 Core Mixture Height (3 Inch Break)
FIGURE 15.6-35 I
(
),
)
l
)
1500. ,
?400, 1300.
7
\ '
I C
o
[1200.
5 1100. . --
m I
s l1000. % '
< 900 - x 3
v 800.
g700.
600.
500.
800. 1200. 1600. 2000. 2400. 2800.
TDE (SEC)
CONANCHE PEAK 5.E.S.
FINAL SAFETY ANALYSIS REPORT UNIT 2 Clad Temperature Transient (3 Inch Break)
FIGURE 15.6-36
toi ,
' =
k2 l Iod 7 E: =
E: _-
g' to 1 g I s- =
o -
$3 -
y li4s-3 3
to'3 i f f !f flf l f f f fIfl I f f fIIII I I IIIIII I I III!!!
to 1 2 s to8 - a s tel a a ses s ses, , 3,4 Tus Arran Taw maci COMANCHE PEAK S.E.S.
FINAL SAFETY ANALYSIS REPORT UNIT 2 Core Power After Reactor Trip FIGURE 15.6-37
420.
200. ~~ --
IBO.
l >
I g 160. , i ., i I40. Y3
- 120. g I
R lee. \
$ N g s i b I,, ,. , ,,, .
g 60. ~
u 40.
20.
B. 508. 1000. 1500. 2006. 2500.
1 5000. 5500. 4000.
TIE (SEC)
COMANCHE PEAK S.E.S.
! FINAL SAFETY ANALYSIS REPORT UNIT 2 Steam Flow l (3 Inch Break)
FIGURE 15.6-38 I
l l
I l
10' -
i i
O I e
l N
h10 3 S -::C lii ! ,!
U u i E
w W
-v 5
e 102
, , - -- ~
10 1 800. 1200, 1600. 2000. 2400. 2800.
TIM (SEC)
COMANCHE PEAK S.E.S.
FINAL SAFETY ANALYSIS REPORT UNIT 2
< Rod Film Heat Transfer coefficient (3 Inch Break)
FIGURE 15.6-39
l l
12ee. c ;
1:ca.
A c'
I
/
i 1200. ;- -
I
^
g aee. f E w }
b f g 800. ,
- t=
ca 7ee. -
3 i cea- \
u f 500. x; I
4ee ?. see. taee. 150a 2ees. 2sae. 5e00. 5see. 4e00, ifE (SEC)
COMANCHE PEAK S.E.S.
FINAL SAFETY ANALYSIS REPORT UNIT 2 Hot Spot Fluid Temperature (3 Inch Break)
FIGURE 15.6-40
__ _ _ _ _ _ - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - ------- -------------- - - ---- ------ ~-------- - - - ~
- 3 4N . - ~
2200.
2980.
- 1989.
m 5
2:
- 1888.
m i 1488.
g E
1290.
i cw ,
1808. w %-
- 1. 1900. 2000. 5000. 4808. 5800. 6808.
T M (SEC)
CONANCHE PEAK S.E.S.
FINAL-SAFETY ANALYSIS REPORT UNIT 2 RCS Depressureization Transient (2 Inch Break)
FIGURE 15.6-41
t 2W._
22N .
2988. --
1989.
1 BM.
b U 14H.
h 12N.
E 1808. \1 3 1
\
x .
.. \
4N. '
298
- 1. 205. 488. 333. 300. ISOS 1200.14N.t$N.1808,2988.
TM (SEC)
CONANCHE PEAK S.E.S.
FINAL SAFETY ANALYSIS REPORT UNIT 2 RCS Depressurization Transient (4 Inch Break)
FIGURE 15.6-42
g.
J 51.l _
50 Si 29.,
s.
- 29. N Il
- v. t
~ 26. l t
I L.g 24 25.
2 2 . - - - -~ ~- - - - - - - - - - ' - - - - - - ' ' - TOP
- - -OF
- 'CORE i
% l 20
- 9. 1988. 2000. 5998. 4899. 5908. 6000.
TDE (SEC)
CONANCHE PEAK S.E.S.
FINAL SAFETY ANALYSIS REPORT UNIT 2 Core Mixture Height (2 Inch Break)
FIGURE 15.6-43
- 52. ,
5*-
-4
) , ,
i i
l I
k- l
- 28. l l l
- 26. S b l E
~.2 \ -
)
h
- 22. =u-**- ' * - * " - "
( TOP OF CORE g"'--- %
N .,
,0. _ .-
/
.N I8. N l 16 D. 090. 490. 680. SH. 1996.1290.1490.1600.1000. 2000.
TD8E (MC) s CONANCME PEAK S.E.S.
FINAL SAFETY ANALYSIS REPORT UNIT 2 Core Mixture Height (4 inch Break)
FIGURE 15.6-44
_ _ _ - - _ _ _ _ _ _ _ _ _ - - - _ - - _ - _ . _ - - _ - - - - - _ - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - ^ - - _ - - -- ~
l 1050.y I
1000. --
^
950' E
w 900, r NV !
w 850.
800. - 1 750.
t 700. - ~ ~
3 u
g350.
k600. I i
$50, hD00. 2500. 3000. 3500. 4000. 4500. 5000. 5500. 6000.
TIME (SEC)
COMANCHE PEAX S.E.S.
FINAi. SAFETY ANALYSIS REPORT llNIT 2 ._ .
Clad Temperature Trtns*
(2 Inch Break) kIGURE 15.6-45 4
- -- ^
1300. ,
- i 1200. -
! I I
i 1100.
g / l' '
1 I
g$000.
200. ' '
o 800.
g700.
600.
E h
NJ ,
- % S i
400 f00. 400. 600, 800. 1000. 1200. 1400. 1600. 1800. 2000.
TDE (MC) l COMANCHE ?EAK S.E.S.
FINAL SAFETY ANALYSIS REPORT UNIT 2 Clad Temperature Transient (4 Inch Break)
FIGURE 15.6-46 l
2500
\
2000-n 1500.
$e -
1000-Q 500-1 0
5O' 1601$0 2b0 2$0 380 350 4E 4$0 580 550 FLOW RATE (LB/SEC)
CONANCHE PEAK S.E.S.
FINAL SAFETY ANALYSIS REPORT UNIT 2 Large Break Safety Injection Flow Rate FIGURE 15.6-47a
2500
(
2000 -
15C@
1000-a 500-L 0 . . . . , , . . . .
0 50 100 150 200 250 300 350 400 450 500 550 FLOW RATE (LB/SEC) i CONANCHE PEAK S.E.S.
FINAL SAFETY ANALYSIS REPOR7 UNIT 2 Small Break Safety Injection Flow Rate FIGURE 15.6-47b 5
l i
13 1
12-11 -
10 -
L g.
f x 8-7-
@ 6-W
& 5-I 4-2-
1 0 '
1 $ $ 4 $ $ 7 $ $ . j'o 11 12 CORE HEIGHT (FT)
COMANCHE PEAK S.E.S.
i
' FINAL SAFETY-ANALYSIS REPORT UNIT 2 Small Break Power Distribution FIGURE 15. "0 I
WEST!!1GHOUSE LETTER. WPT 14479 TV ELECTRIC COMPANY COMAtlCHE PEAK STEAM ELECTRIC STATION UNIT NUMBER 1 NOTRUMP SHALL BREAK LOCA ANALYSIS - ENGINEERit4G ASSESSMENT IN SUPPORT Of C0f1TINUED OPERAT!0t1 APRIL 15, 1992 El1 CLOSURE 3 TO TXX-92323 (TOTAL PAGES - 12)
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= q g, g fg 9 WPT-14479 l ET-NSL-0PL-II-92-185 l Westinghouse Energy Systems Ba 3t$ !
Electric Corporation ^
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- S 300355 !
i April 15, 1992 Mr. W. J. Cahill, Jr., Executive Vice President 5.0. No. TBX.4708 Nuclear Engineering & Operations TV Electric Company P. O. Box 1002 Ref: 1. WPT-14387
-Glen Rose, Texas 76043 Attention: W. Choe (No Response Required) l TU ELECTRIC COMPANY COMANCHE PEAK STEAM ELECTRIC STATION UNIT NUMBER 1 NOTRUMP SMALL BREAK LOCA ANALYSIS - ENGINEERING ASSESSMENT IN SUPPORT OF CONTINUED OPERATION
Dear Mr. Cahill:
As' discussed with Mr. Whee Choe of TV Electric, a single.small break LOCA analysis was performed for Comanche Peak Unit I using the NOTRUMP model. This analysis is based on the NOTRUMP. analysis performed for Unit 2 and transmitted via Reference 1 above.
. Attached please-find an Engineering Assessment based on this Unit I analysis.
TU Electric may use this assessment in support-of a Justification for Continued 0)eration (JCO) of Comanche Peak. The need for a JC0 could arise in
-the event0 tie existing W-Flash analysis Peak Clad Temperature (PCT) exceeds the 2200 F. acceptance criteria due to penalities-associated with new safety issues and/or plant changes resulting in 10CF50.59 Safety Evaluations. '
If there are any questions on the above ur attached please contact Craig ;
Thompson on 412/374-4409 or Roy Owoc on 412/374-4037.
-This letter closes Westinghouse open item No. 10404-7.
F Very truly yours, WESTINGHOUSE ELECTRIC CORPORATION R. H. Owoc.
J. L. Vot , Manag9r k
Comanche Peak Projects -
[ l
- _ _ . ~ . . _ _ ~ . _ _ _ _ _ _ _ . _ . . _ _ . . _ _ _ _
COMANCHE PEAK STEAM ELECTRIC STATION UNIT No.1 NOTRUMP SMALL BREAK LOCA ANALYS!$
ENGINEERING A55ES$ MENT IN SUPPORT OF CONTINUED OPERATION SACKGROUND Westinghouse (Ref:1) transmitted the results of a 10CFR50.5g safety evaluation to remove the LOCA analysis credit for the turbine driven i
auxiliary analyst feedwater pump from the current licensing basis small break LOCA 2133.65g. F. This evaluation-increased the small break LOCA PCT to Otscussions with TU Electric regarding current Westinghouse open Potential Items (P!s). in particular the item on Small Break.LOCA
. Burst and Blockage Consideration, reported to TU Electric in Reference 2, coulg,whenfullyresolvedresultinthecurrentsmallbreakexceedingthe 2200 F criteria. Westinghouss/TU Electric agreed to reanalyze CPSEs-1 with the newer NOTRUMP evaluation model as a means to support continued operation. Westinghouse would provide TU Electric with an engineerin ;
assessment which TU Electric can use to support continued operation. g Since application of the NOTRUMP small break methodology-to C?$ES-1 has not received NRC approval, application of the NOTRUMP methodology is considered outside the licensing basis:for CPSE$-1.
Small treak LOCA-Eneineerine Assessment The snail-break LOCA analysis-of recoH for' Comanche Peak Unit I was performed using the WFLASH model (Ref:3). The limiting break size was a four: inch diameter cold leg break which predicted.a peak clad temperature of 1787.5'F. Safety evaluations;that have been performed against-this-analysis are listed in Table 2. The cumulative result of these safety evaluations is a final PCT of 2133.65'F. This result when combined
-withcurrentlyopenPotentialIssueswhichaffectsmallbreakLOCA analysis-could result in a PCT above the 2200 F 10CFR60.46 Criteria. In order to have a-basis for continued operation, in the event that-i -
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. unacceptable results would be obtained for the current W-FLASH analysis, the Comanche' Peak Steam Electric St: tion Unit No.2 NOTRUNP small break LLOCA analysis'(Ref:4) was used as-a basis to perform a CPSES-1 NOTRUMP small break'LOCA analysis.
The NOTRUMP CReft 5&6 small break LOCA code has received NRC approval for use in a Itcensing'ame)ndment in support of .
small break LOCA analyses performed under the requirements of Appendix K to 1C;CFR part 50.- The most limiting break tc'entified in the Reference 4
' analysis, .the 5-inch cold leg break,: was repeated by changing ' appropriate input to model the CPSES-1-core having 17X17 Standard Fue' (Fuel Rod 0.D.
p of 0.374 inches)-and changes necessary to model the CPSES-1 model 04 steam generator, since CPSES-2 has a model 05 design.
[ Theregultsofthe.CPSES-1NOTRUMPsmallbreakanalysiswereaPCTof 1418.4 F, and a local maximus' zirconium water oxidation of 0.55%.- These ,
results are such that the-additional 10CF#50.46 criteria- for core wide ~
oxidation, into question. coolable geometry and long-term core cooling are not called-The'CP5ES-1. current licensin 4-inch cold leg break to be:g basis analysis, using W-FLASH, had shown the-limiting. :However, the 3-inch cold' leg' break:
was analyzed: for:CPSES-1 using NOTRUMP'since the Reference 4 analysis has shown this break to.be more l'atting for CPSE5-2. s Traditionally, analyses
+F-wt,--**4 y,eae---w *'ey>- -n"-w--r rv r- a-"'=a-et-*- -ee"r$-T>N'-1P 'g' -'uripii:,g--ye--g-Ag r -y- g- y-ir7-ey +e ,y em-v+ew4.-wyw"' -
gr ee ' W
o COMANCHE PEAX STEAM ELECTRIC STATION UNIT NO.1 NOTRUMP SMALL BREAX LOCA ANALYS!$
ENGINEERING ASSESSMENT IN SUPPORT OF CONTINUED OPERATION using the NOTRUMP code hav6 shown smaller breaks to be more limiting when compared to W-Flash results for the same plant. Therefore a shift to a break smaller than the current WFLASH 4-inch break was not unexpected when CPSES-2 was anslyzed.
Further, based on the Reference 4 analysis, which has shown a greater difference in calculated PCT between the 2-inch and 4-inch breaks when compared to the 3-inch break, than calculated for the difference between the CPSES-1 & 2 cores, the CPSES-1 single break analysis (3-inch cold leg break) is justified and a spectrum of breaks is, in Westinghouse's judgement, r'Jt required in support of this engineering assessment of coatinued operation.
Since use of the NOTRUMP small break LOCA evaluation model has not bee approved for use on CPSES-1, via the licensing amendment process, the above single break analysis for CPSES-1 is considered to be outside the licensing hasis for CPSES-1.
Laroe Break LOCA. LOCA Hydraulic.Forcino functions. Post-LOCA SubcriticalitrRecuirement.
recirculation ad._SwitchcVer to prevent potential of the ECCS to hot leo boron oreciottation.
The remaining LOCA licensing requirements listed above are unaffected by changes in small break LOCA analysis, or choice of small break LOCA evaluation model. Therefore, an evaluation of these licensing requirements is not provided with this engineering assessment.
Conclusion i A new small break LOCA analysis, using the NOTRUMP small break evaluation model, has been performed for CPSES-1. This new analysis, using all changes previously evaluated under the provision of 10CFR50.59.. calculated a low of 2200PCT C
F. showing a large amount of margin to the 10CFR5G.46 requirement This result indicates that NOTRUMP calculates improved core cooling when compared to the older W-FLASH model.
Should the W-FLASH 0 F criteria, the analysis be evaluated to have a PCT over the 2200 l
i NOTRUMP result anellorates any concern with regard to safe operation. In
! the event that 0 the W-FLASH analysis for CPSES-1 is evaluated to have a PCT above 2200 F, the NOTRUMP analysis can be used as a basis for continued operation of CPSES-1.
6.0 REFERENCES
- 1) letter WPT-14390, d. L. Vota (W) to W. J. Cahill, Jr. (TUE),
' Comanche Peak Steam Electric Station Unit Number 1, Safety
. Evaluation to Remove LOCA Analysis Credit for the Turbine Driven
. Auxiliary Feedwater Pump', March 9, 1992.
- 2) Letter WPT-12933, J. L. Vota (W) to W. J. Cahill, Jr. (TUE),
' Comanche Peak Steam Electric Station, ECCS Evaluation Model Changes', June 20, 1991
- 3) WCAP-8200 Rev.2 (PROPRIETARY) AND WCAP-8261 Rev.1 (NON-PROPRIETARY), 'WFLASH - A Fortran-IV Computar Program for Simulation of Transients in a Multi-Loop PWR", June 1974
- 4) Letter WPT-14387, J. L. Vota (W) to W. J. Cahill, Jr. (TUE),
' Comanche Peak Steam Electric Station Unit Number 2, Small Break LOCA ECCS Reaiialysis", February 26, 1992
- 5) WCAP-100079-P-A-(PROPRIETARY) and WCAP-10080-A (NON-PROPRIETARY)
'NOTRUMP: A 900AL. TRANSIENT SMALL BREAK AND GENERAL NETWORK CODE', AUGUST 1985.
- 6) WCAP-10054-P-A (PROPRIETARY), ' WESTINGHOUSE SMALL BREAK LOCA ECCS
- EVALUATION MODEL USI' d G THE NOTRUMP CODE", AUGUST 1985.
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TABLE 1 Safety Evaluations for the Comanche Peak thit 1 Large Break LOCA Analysis PCT Penaltv Reference Evaluation Descriotion
- 1. 0.0'F CWS-TBX-895 Reduced SI flow would reduce spilling, with no impact on core or downcomer levels during reflood.
- 2. 0.0'F SED..SA-296 Bottom of core recovery delayed less than 0.02 sec. Later, downcomer filled slightly earlier due to higher flow. Supersedes evaluation number 1.
- 3. 6.2*F SED-SA-340 Modified steam generator bypass flow. Increase in initial core inlet temperature.
- 4. 0.0'F SED-SA-774 Revised SI flow tech. spec.
Increased SI is a benefit since Comanche Peak 1 is not a max-SI plant.
- 5. 10.0*F SED-SA-884 Reduced accumulator water volume by 6 cubic feet.
- 6. 0.0'F SED-SA-1048 Reduced auxiliary feedwater flow.
- 7. 0.0'F SECL-88-706 Increased the signal processing delay time from I sec. to 2 sec.
E
- 8. 0.0*F SECL-89-210 Installed heated junction i
1 thermocouples and shrouds, l
.a TABLE 1 cont.
Safety Evaluations for the Comanche Peak Unit 1 Large Break LOCA Analysis PCT Penalty Leference Evaluation Descriptien
- 9. 18.6'F SECL-89-594 Rev 1 Increase in S/G tube plugging.
2.1% area correction and 1% SGTP.
- 10. ......... SECL-89-494 Steam generator feedwater fi split. Same as evaluation 3,
- 11. 1.0'F SECL-89-432 Reduced RHR flow due to delay in isolating the miniflow lir,es.
- 12. 0.0*F SECL-89-672 Increased the main steam safety valve blowdown.
-_13. 0.0*F SECL-80-1011 Increased the upper nitrogen pressura limit for the accumulators.
- 14. 0.0*F SECL-89-964 Increased the AFW purge volume used to calculate the time to switchover to the lower enthalpy.
- 15. 0.0*F- WPT-11168 Comanche Peak Steam Electric Station Setpoint Study-Information; Pressurizer = Low Pressure SI at 1700 psig and-containment HI-1 at-5'0 psig.
- 16. 0.0*F' SECL-90-135 ' Automatic AFW Contraller Safety Evaluation.
i
TABLE 1 cont.
Safety Evaluations for the Cominche Peak Unit 1 Large Break LOCA Analysis PCT Penalty _ Reference Evaluation Description
- 17. 0.0*F SECL-90-195 Revised Charging Flow Evaluation.
- 18. 0.0*F SECL-90-215 Reevaluation of the effect on small break LOCA for reductions in Ch:rginq SI and HHSI. This evaluat'on rescinds SECLs90-135, 195 and SED-SA-296.
- 19. 0.0'F SECL-90-293 Increased AFW purge volumes due to check valve back leakage.
- 20. 12.0'F Thimble tube modeling penalty, NRC GENERIC LETTER 86-016.
- 21. 0.0*F SECL-90-329 Revised Auxiliary Feedwater purge volumes.
- 22. 0.0*F SECL-90-352 Increased Main Feedwater Isolation time.
- 23. 0.0*F SECL-90-545 Increased Auxiliary Feedwater flow from 625 gpm to 1225 gom, entire purgevolumeassumedtobeat 440 F.
- 24. 0.0'F SECL-91-088D Increased start time for the stearu driven turbine auxiliary feedwater pump. The PCT change is based on-an assumed total auxiliary feedwater flow rate of 1290 gpm compared to the SECL-90-545 assumption of 1225.5 gpm.
- 25. 7.2'F WPT-13635 Permanent changes to the ECCS evaluation model.
- 26. 0.0*F SECL-91-3670 ECCS Flow changes to prevent runout of the Charging /SI and HHSI during post-LOCA recirculation.
TABLE 1 cont.
Safety Evalua+. ions for the Comanche Peak Unit 1 Large Break LOCA Analysis I
- 28. 0.0*F WPT-xxxxx Engineering assessment for small break LOCA performed with NOTRUMP.
No affect on large break LOCA.
PCT Penalty;. Reference Eyaluation Descriplion 55.0'F Total PCT penalty for 10CFR50.59 changes and permanent ECCS model changes.
2010.7'F Limiting Case PCT 2065.7'F Total Limiting Case PCT
TABLE 2 l
! Safety Evaluations for the Comanche Peak Unit 1 Small Break LOCA Analysis ECl_ffndtL Mferenc9 fldVatior, De;qIjffion
- 2. 88.0'F SED-SA-296 4.4% shortfall is SI flow delivered over time period of interest.
Supersedes evaluation number 1.
- 3. 0.0*F SED-SA-774 Revised SI flow tech. spec.
Increased SI is a benefit.
- 4. 11.0'F SED-SA-1048 Reduced auxiliary feedwater flow from 1410 to 1290 gpm.
- 5. 9.0*F SECL-88-706 Increased the signal processing delay time from I sec. to 2 sec.
- 6. 0.0'F SECL-89-kl0 Installed heated junction thermocouples and shrouds.
- 7. 0.0'F SECL-89-594 Rev. 1 Increase in S/G tube plugging.
2.1% area correction and 1% SGTP.
- 8. 0.0*F SECL-89.-494 Steam generator feedrater flow rplit.
- 9. 0.0'F SECL-89-432 Reduced RHR flow due to delay in isolating the miniflow lines.
I
TABLE 2 cont, 1
Safety Evaluations for the Comancho Peak Unit 1 Small Break LOCA Analysis PCT Ptnalty Reference [ valuation Dascripljon l 1
l
- 10. 0.0*F SECL-89-672 Increased the main steam safety valve blowdown.
- 11. 0.0'F SECL-89-1011 Increased the upper nitrogen pressure limit for the accumulators.
- 12. 53.0'r SECL-89-964 Increased the AFW purge volume used to calculate tha time to switchover to the lower enthalpy.
- 13. 2.0'F WPT-11168 Comanche Peak Steam Electric Station Setpoint Study Information. Pressurizer Low Pressure SI at 1700 psig.
- 14. 75.5'F SECL-90-135 Automatic AFW Controller Safety Evaluation.
- 15. 84.0*F SECL-90-135 Revised Charging Flow Evaluation.
- 16. 121.0'F SECL-90-215 Reevaluation of the effect on small
-88.0'F break LOCA for reductions in
-75.0'F Charging SI and HHSI. This
-84.0*F evaluation supersedes SECLs90-135, 195 and SED-SA-296.
- 26.0'F
g a' =
TABLE 2 cont, l
Safety Evaluations for the Comanche Peak Unit 1 Small Break LOCA Analysis PCT Penalty Refer 9nge fraluation Description __
- 17. 19.2*F SECL-90-293 Increased AFW purge volumes due to check valve back leakage.
- 18. 0.5'F SECL-90-329 Revised AFV purge volumes.
-11.0'F Supersedes evaluation performed in SED-SA-1048 (07/01/85). The 110F penalty has been removed since the analysis value of 625 gpm is tonservative when compared to the CPSES Unit No.1 Aux feed flow of 1290 gpm.
- 19. 0.0*F SECL-90-352 Increase in the M;in feedwater Isolation time.
- 20. -25.0*F SECL-90-545 Increase ir. t% Auxiliary Feedwater flow rate for 625 gpm to 1225.5 9pr:, entire purge volume assumed to be at 4400F,
- 21. 2.0'F SECL-90-545 Adjustment to the small break analysis results for the correction-to the Zirc/ Water error.
- 22. 0.0'F SECL-91-0880 Increased start tirm for th) steam driven turbine auxiliary feedwater pump. The PCT chan2e is based on an as.umed total auxiliary feMwater flow rate of 1290 gpm compared to the SECL-90-545 l assumption of 1225.5 gpm.
L___
TABLE 2 cont.
Safety Evaluations for the Comanche Peak Unit ) Small Break LOCA Analysis ECT Penalty _, Eti m ate Evaluation Descriotion
- 23. 0.00*F WPT-13635 Permanent changes to the ECCS evaluation model.
24, 64.85'F SECL-91-3670 ECCS Flow changes to prevent runout of the Charging /SI and HHS! during post-LOCA recirculation.
346.15'F Total PCT penalty for 10CFR50.59 changes and permanent ECCS model channes.
1787.5'F Limiting Case PCT 2133.65'F Total Limiting Case PCT l
l
- 26. -715.25'F WPT-XXXXX Engineering assessment for CPSES-1 l NOTRUMP small break LOCA analysis.
! This is a temporary use of PCT margin until the Engineering assessment can be replaced.
1418.40*F Total Limiting Case PCT using the NOTRUMP methodology.
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