ML052450017
| ML052450017 | |
| Person / Time | |
|---|---|
| Site: | Saint Lucie  |
| Issue date: | 09/01/2005 |
| From: | Florida Power & Light Co |
| To: | Office of Nuclear Reactor Regulation |
| Moroney B, NRR/DLPM, 415-3974 | |
| References | |
| Download: ML052450017 (47) | |
Text
FPL/W/NRC Interface Technical Meeting Proposed Reduction in Reactor Coolant System Flow and Reactor Power to Support Increased Steam Generator Tube Plugging September 1, 2005 Agenda Purpose Technical Specifications Changes Models and Correlations Applicability to 42% SGTP Analyses Parameters Comparison of Analyses Input Parameters 300,000 gpm Flow & 42% SGTP vs. 335,000 gpm Flow & 30% SGTP Event Specific Comparisons of Results Non-LOCA Analyses LOCAAnalyses Dose Evaluations W!!-t M
S~de-
Agenda (cont.)
OtherTopics Partial Submittal Other Related PLA Submittals Summary
_41frr Purpose
- Continue briefing the NRC on upcoming plant license amendment in support of increased SGTP (up to 42%)
ReduceTechnical Specifications RCS Flow Reduce Reactor Thermal Power
- Finalize on TS RCS Flow change wording
- Provide an update on ongoing analyses
- Discuss and obtain NRC feedback on partial submittal and any potential issues
_d4
Proposed Tech Spec Changes TS LCO 3.2.5, DNB Parameters: Include additional footnote in Tech SpecTable 3.2-2 for 335,000 gpm RCS flow rate, as follows:
Commencing with the startup for Cycle 16 and until the Combustion Engineering Model 3410 Steam Generators are replaced, if the Reactor Coolant System Flow Rate is less than 335,000 gpm but greater than or equal to 300,000 gpm, then the maximum reactor THERMAL POWER shall not exceed 89% of RATED THERMAL POWER of 2700 MWt:
Slids Proposed Tech Spec Changes (Continued)
TS SR 4.2.5.2: Modify footnote on Tech Spec page 3/42-14 to replace "Ž90%" with "Ž80%"
Definition of RATED THERMAL POWER remains unchanged The Tech Spec change submitted with Sleeving PLA to update the reference to WCAP-1 591 8-P, Rev2 No change to Figure 2.1-1 (Thermal Limit Lines). Figure remains bounding for proposed reduction in RCS flow to 300,000 gpm Sltide 6,
TS Wording Currently Planned for the Sleeving LAR InsertCforTS Page3/44-15:
- 10. Tube Repair refers to sleeving with Westinghouse Leak Limiting Alloy 800 sleeves as described in WCAP-1 591 8-P Revision 2, which are used to maintain a tube in service. Leak Limiting Alloy 800 Sleeves are applicable only to the original steam generators. The pressure boundary portion of the original tube wall in the sleeve/tube assembly (i.e., the sleeve-to-tube joint) shall be inspected prior to installation of each sleeve. In addition, Leak Limiting Alloy 800 Sleeves that have a nickel band hard roll shall be plugged or removed from service after one cycle.
New TS Change identified for 42% SGTP Submittal I InsertCforTS Page3/44-15:
- 10. Tube Repair refersto sleeving with Westinghouse Leak Limiting Alloy 800 sleeves as described in WCAP-15918-P Revision 2*, which are used to maintain a tube in service. Leak Limiting Alloy 800 Sleeves are applicable only to the original steam generators. The pressure boundary portion of the original tube wall in the sleeve/tube assembly (i.e., the sleeve-to-tube joint) shall be inspected prior to installation of each sleeve. In addition, Leak Limiting Alloy 800 Sleeves that have a nickel band hard roll shall be plugged or removed from service after one cycle.
. As amended by Attachment XX of Enclosure XX of FPL Letter L-2 005-XXX Sride 8,
Models and Correlations Each analysis that is being performed has confirmed the applicability of methods and correlations No existing method or correlation has been identified to be outside the bounds of applicability for the revised operating conditions W;-Irz77e;I Applicability of Methods - DNB Correlations
Pressure (psia)
Local Mass Velocity (Mlbm/hr-ft2)
Local Quality (Fraction) 1750to2415 0.8 to 3.16
-0.14 to 0.22
- These conditions are satisfied for all applicable analyses.
SIlde sride!,eg--
Applicability of Methods - DNB Correlation (Pressure)
Inadvertent Opening of a Pzr Safety or Relief Valve 2900 Pressudzer Pressure e1700 lrimit on DNB Correlation (1750 psla) 0.00 2.CO 4.00 e0 800 10.00 100 14.00 16.00 1.00 20.00 Time (seconds)
Sride I Applicability of Methods - DNB Correlation (RCS Flow)
Complete Loss of Fbirced Reactor Coolant Flow 2.5 _
F
~ 0
.g 2.0 -_
, 1.5
.2 0.5 0.0 o
0.0 2.0 4.0 6.0 Time (seconds) 8.0 10.0
,.~. WM-7 '
Slide e12
Applicability of Methods - RETRAN Model RETRAN SG Model Currently used for conditions that bound the conditions associated with the 42% SGTP program
- Note that although total RCS flowtends to drop with increased plugging levels, mass velocity on tube side tends to increase.
- Secondary side global conditions, such as the steam pressure, are within the bounds of what has been modeled in the past for St. Lucie Unit 2 and for other plants (see analysis results plots).
Side 13.o41 Comparison of Analyses Input Parameters Current Analysis New Analysis SGTP 30%
42%
TDF 335,000 gpm 300,000 gpm MMF 341,400 gpm 314,000 gpm Target 100%
89%
Maximum Power Target Tavg 576.5 574.5 Side1jz
Comparison of Analyses Input Parameters TWAWI Mir PI M
W -
F, "IA
-i IN I-hl 14 r
9 10 1 7
,D I4S_
729 177 1424 2__
24 1 B7'77 9
_9 _
9113 9
t2o7___
9W Ti r r_
I_2 921 1fl7 __
l_
12l7 lCM D~p P 132" IM 11%
1 b9 IXI 121 17 ____
ll____
37 17 3.7 17 I1C.L.
?04 V
340u.o9 tn9 7159 449 C
-59q 21 14.19 7714 617 V_1A__
711 4
77
-7
.M. I 12 726
- 9W 1117 4714 714 7195 4 _
o_
74 477 724 39~d
.o 411.29 11.70 1432 57 D79 47 4C43 7
" o7 TVrvk.77B17.
_517 11417 Omi' 4
17a T4444044.%
20 20 42 4
Z_ t_ T 9-1 1
m
.17
.~.
Slde s Comparison of Analyses Input Parameters
- Reactor Protection System Setpoints
- Unchanged
- Low Flow Trip Setpoint accounts for the flow reduction
,Engineered Safeguards Features Setpoints
- Unchanged NCOLR Limits
- Peak Linear Heat Rate reduced from 12.5 to 12.0 kW/ft
- DNB-LCO Figure 3.24 modified to gain analysis margin at lower powers (reduced operating space)
- LPD-LCO Figure 3.2-2 modified to gain analysis margin at lower powers (reduced operating space)
- srd1,
Event Specific Comparisons of Results Overview:
Majority of the analyses are complete Preliminary results from all analyses show results as expected (no surprises)
- All analyses meet the appropriate acceptance criteria (same criteria as 30% SGTP cases)
All models and correlations remain applicable for the new analyzed conditions with no violation of parameter ranges.
Sride DNBR Limits -30% SGTP Program DNBR Uirnits and Margin Summary (30% SGTP Program)
I 2
3
._f I
SUde IS
DNBR Limits - 42% SGTP Program DNBR Limits and Margin Summary (42xX. SGTP Program) ik Dmar~n 1%
2f lNB Marin 5.1% DNB Magin I
2 3
19
,gf% vr!,.-73VT-
J. t Sie I Increase in Feedwater Flow Acceptance Criteria DNBR Case(s) for Reanalysis
- HFP
- HZP Potential Limiting modeling conditions None Slie 0df,
Increase in Feedwater Flow Sequence of Events for Minimum DNBR Case Event 30% SGTP 42% SGTP Time (seconds)
Time (seconds)
Main Feedwvater Control Valves Fail Full Open 0.0 0.0 Variable High Power Trip Setpoint is Reached 38.83 24.54 Rod Motion Begins 39.57 25.28 Minimum DNBR Occurs 40.00 26.00 Turbine Trip 40.83 26.54 Minimum DNBR Value 1.97 2.24 DNBR Limit 1.42 1.34 Sridele-Increase in Feedwater Flow Comparison of DNBRs 3
2.5 2
z 1.5 1
0.5 0
I 2
I 2.-
G-
Increase in Feedwater Flow 1.2 2
.82 0
A
'4,
.4.,
1.2 Time (sec)
I.
Slide Increase in Feedwater Flow 12
.8 e
.4
- _qRE, Slide2,
Increase in Feedwater Flow 2200 U,
21000 V
C-Z5 MOO
_t
&I130 Slide~
Increase in Feedwater Flow A,,~ ~ ~
~
d 2
S 0
S 5W..
0 0
20 30 40 50 60 Time (sec)
Increase in Feedwater Flow Time (sec)
-W Slide 2 Increase in Feedwater Flow 4.5-4 is -
I
Time (sec) 55re280sX"
Pre-Trip Steam System Piping Failure Acceptance Criteria DNBR
- Peak linear heat rate Case(s) for Reanalysis Limiting MDC cases
- Address failure of the fast bus transfer Break coincident with LOOP at time zero Sride2 Pre-Trip Steam System Piping Failure - Results (trip case)
Comparison of DNBRs 3
30% SGTP 42% SGTP 25 -
2 -
alysis 1.53@S3s 1.372 1.533 1DN9R SAL;go I
1 2
Slides
Post-Trip Steam System Piping Failure Acceptance Criteria
- DNBR Case(s)
- With offsite power available Potential Limiting modeling conditions
- None Results-DNB Design Basis satisfied for both 30% and 42% SGTP Programs e521-X 02 Loss of Condenser Vacuum Acceptance Criteria DNBR Primary and Secondary overpressure Case(s) for Reanalysis DNB
- Primaryoverpressure
- Secondaryoverpressure
- Inoperable MSSVs Potential Limiting modeling conditions None rx-zwlr
W*
Sride3
Loss of Condenser Vacuum - Peak RCS Pressure Sequence of Events for Peak RCS Pressure Case Event 30% SGTP 42% SGTP Time (seconds)
Time (seconds)
Turbine Trip 10.0 10.0 Manual Feedwater Termination (Both loops) 10.0 10.0 Reactor Trip on High Pressurizer Pressure 18.9 19.8 Rod Motion Begins 19.6 20.5 Time of Peak RCS Pressure 20.9 20.8 Peak RCS Pressure 2691 psia 2664 psia RCS Pressure Maximum Limit 2750 psia 2750 psia _
5flde3~%9 Loss of Condenser Vacuum - DNB Case Sequence of Events for Minimum DNBR Case Event 30% SGTP 42% SGTP Time (seconds)
Time (seconds)
Turbine Trip 10.0 10.0 Manual Feedwater Termination (Both loops) 10.0 10.0 ReactorTrip on High Pressurizer Pressure 20.3 19.8 Rod Motion Begins 21.0 20.5 Time of Peak RCS Pressure 22.1 20.8 Minimum DNBR Value 2.19 2.24 DNBR Limit 1.42 1.34 4
L
. 4_CR ST
Loss of Condenser Vacuum - DNB Case Comparison of DNBRs 3.
2.5 snow Analysis Result 2-T 1.5 C]
- DNBP!sSt~
I C-44 0.5 - ~
0 1
2 Loss of Condenser Vacuum - Peak RCS Pressure 1.2 I-a, LR.
0R
,.I Sride 360
Loss of Condenser Vacuum - Peak RCS Pressure tIo 10 0 II.
imzs-'s' SEde& 3A Loss of Condenser Vacuum - Peak RCS Pressure 4,
-5
- .1 ee M.
4, 100
,~v ~
Sri d
Loss of Condenser Vacuum - Peak RCS Pressure 600 c 5.o E
w 570 563 a
20 40 E0 B
10O Time (s)
Side39 Loss of Condenser Vacuum - Peak RCS Pressure 1100 lC C-6
- a. 1r00 U,
ua-00 4-,
(.r, 700
-.8b
- 80 10 Sfde4O
Loss of Condenser Vacuum - DNB Case 12 0
10 a.,
Time (s) r-4 Slide 4 Loss of Condenser Vacuum - DNB Case 260a 2500
.O 2420 a 2300 E Ma
(
2100 2020 1900
.."It
,.1 --M lln"T--Tg
-;: -aj
5l2de42
Loss of Condenser Vacuum - DNB Case
=
1100 A,
- c e x-c a-60 80 100 Time (s) i 41-10 srzM Loss of Condenser Vacuum - DNB Case Sfide 11
Loss of Condenser Vacuum - DNB Case 45 4
35 2.5
-q.r-Feedwater System Pipe break Acceptance Criteria Secondary Overpressure Primary Overpressure Case(s) for Reanalysis Primary peak pressure < 110% of design pressure Small break with failure of fast bus transfer (FFBT)
Large break without FFBT Primary peak pressure< 120% of design Large break with FFBT Secondary peak pressure Limiting break size (without FFBT)
Sided
/~'
Feedwater System Pipe break Umiting Breaks for Primary RCS Overpressure Event 30% SGTP 42% SCTP Time (seconds)
Time (seconds)
Initiation of Event -break occurs 0.01 0.01 Manual Feedwater Isolation (both loops) 0.01 0.01 Reactor Trip 30.4 33.7 (Low SO Pressure)
(High Pressurizer Pressure)
Reactor Trip (Breakers open) 30.8 34.1 FailureotFast BusTransfer 30.8 34.1 (Two RCPs Coatdown)
Rod Motion Begins (0.74 seconds 31.5 34.8 following Breaker opening)
Time of Peak RCS Pressure 33.5 33.5 Remaining pumps trip 33.8 37.1 Peak RCS Pressure 2775 psia 2710 psi' RCS Pressure Limit 3000 psia 3&00
¶ V
.iM Slide 4 7,
Asymmetric Steam Generator Transients Acceptance Criteria
-DNBR
- Peak linear heat rate Case(s)
- Maximum and Minimum SGTP (42% and 0%)
Potential Limiting modeling conditions
- None Slide 48.1
Asymmetric Steam Generator Transients Sequence or Events for Maximum Steam GeneratorTube Plugging Event 30% SGTP 42% SGTP Time (seconds)
Time (seconds)
Main Steam Isolation (Loop Two) 10.1 10.1 Manual FeedwaterTennination (Loop Two) 10.1 10.1 Reactor Trip on HSGDP 15.5 15.7 Rod Motion Begins 16.3 16.4 Time of Minimum DNBR 17.4 17.25 Minimum DNBR Value 1.770 2.014 DNBR Limit 1.42 1.37 Asymmetric Steam Generator Transients Comparison of DNBRs for Max Plugging 2.5 2
1.5 I
0.5 zQ 0
.1 2
f-r-51, sfideso p
Asymmetric Steam Generator Transients 1.2 I.-
C..2 0
10 20 30 40 SD 60 70 Time (s) srose Asymmetric Steam Generator Transients 2260 2243 2220 22W0 210\\
4 2160 42 2140 2120 2100 0
Io 20 30 40 5b 70 rime (s)
Sr~de
Asymmetric Steam Generator Transients 530 o56
'IN 0
40 S
0 7
C.-
1-d 550 z 540.
1,..
5030406 Time (s) 5 Hd e Asymmetric Steam Generator Transients 610 570-use 0
10 20 30 40 60 70 Time (s) sr'570O N
.3 s550
lAsymmetric Steam Generator Transients
,,00 Time (s)
,._;r 7 Complete Loss of Flow Acceptance Criteria DNBR Case(s)
- Maximum SGTP Potential Limiting modeling conditions
- Mass flux range associated with the CHF correlation
'3 Slide 5
Complete Loss of Flow 500 EC 700
=
6W0 3:
-60 0
'4 430 0
I.-
Time (Sec) 4 so s2a;ng
Complete Loss of Flow
-3 E-I A.>3 0
2 4
6 1
Time (Sec)
Sr~de Complete Loss of Flow
-a
.4
.2 O1 6
10 Time (Sec) sride60'
Complete Loss of Flow Fz C)4)
-6 4) o0.6
- 4)
.4.-
0 c-Slide Complete Loss of Flow 250Y
-a
- 4) 2300 X-2250 C-
- 5 C-2150 T5.e-M V4,
Complete Loss of Flow r'
6.0
- 4) O 6 640 540 O
2 46 l
Time (Sec)
-j7T, Locked Rotor Acceptance Criteria DNBR Primary overpressure Cladding average temperature Case(s)
Rods-In-DNB with failure of fast bus transfer (FFBT) and subsequent Loss of Offsite Power (LOOP)
Peak pressure/ maximum clad average temperature with FFBT and subsequent LOOP Potential Limiting modeling conditions Mass flux range associated with the CHF correlation Srde 6a
Locked Rotor Results:
Rods-ln-DNB Both 3O% SGTP and 42% SGTP -
<1 % rodsin-DNB Peak pressure:
30% SGTP 4 2646 psia 42% SGTP -
2637 psia Maximum clad average temperature with
- 30%SGTP-*1639F 42%SGTP >1668F ide 6 Locked Rotor-Rods-in-DNB
,I SIi&,
6~~
Locked Rotor - Rods-in-DNB
,0 0
U-C-
'a Time (see)
_ore Locked Rotor - Rods-in-DNB 4.,
8 15
un, 11=41,11111-1 slde68.
Locked Rotor - Rods-in-DNB a-4.1 00~
V-Time (sec)
-W-.
vr-7-g 'ct s._...
51/Be6/
Locked Rotor - Rods-in-DNB 0
4-
-)
.:L Time (sec)
Sr~de 70,4 PI~
Locked Rotor-Rods-in-DNB
- °e 2550 e 2500 u) a) 2450
- I, t
240 Cl Time (sec)
Sride 7.1.4 600 Locked Rotor - Rods-in-DNB 2650 200 245 200 4
2 4
6
)
Time (sec) o r
b Sride 72%
Rod Withdrawal from a Subcritical or low Power Condition Acceptance Criteria DNBR and Fuel centerline temperature Case(s)
- Peak fuel centerline temperature Srde73%
Rod Withdrawal from a Subcritical or low Power Condition Comparison of DNBRs ca a
2.5 2
1.5 0
0.5 0
1 2
3 4
5 6
7 L
..t bikn Sride
CVCS Malfunction
. Acceptance Criteria
- Termination of event before pressurizer fill.
. Results
- The pressurizer volume does not become water solid prior to 20 minutes after event initiation. Event is mitigated by operator action prior to the pressurizer filling and no water is discharged through the PSVs.
Sri & 7,5 Steam Generator Tube Rupture
. Steam GeneratorTube Rupture Analysis of SGTR is completed Results of SGTR analysis are used in the radiological dose calculations sflde76,%
Steam Generator Tube Rupture Analysis Sequence of events for the Steam Cenrator Tube Rupture with Loss or offite power Time (see)
Event Setpoindvaine 30%
42%
30% SGTP 42% SCTP SGTP SCTP_
0.0 0.0 Tube Rupture Occurs 379.2 351.6 Renctoe Trip Signal on Floor ofTMtP. Fsu 2142 2142 380.94 353.24 CEAs Begin to Drop 384.i 361.1 MSSVS open.5 psit 970 970 391.8 364.9 Maximum, SO Prosure. psi 996 971 487.0 520.4 Pressurizer Empties 490.8 5232 SIAS Geneated on Low Presurizer Prtsure. pai 157S 1578 520.3 55t.2 l0'SI Pumps Reach Fuli Speed 1800 1800
- 1. Operator Borates to Cold Shutdossn Coneontration
- 3. Operator Actirates ADVs (Intact SG) to Conmmence Cooldown of RCS Steam Generator Tube Rupture Analysis Sequence of Events for the Steam CeneratorTube Rupture with Lo r orsite Power Time (sfe)
Event Setpoint/Volue 30%
42%
30% SCTP 42% SGTP SCTP SGTP Steam Releases via Turbine to Condenser Prior to Trip 1.255.628 1.030.t05 (50. 1% Affected SGt49 9'/. Intact SGI lb m MS-s Close. Adrectedantact SO Steam Releases. Ibm 78S786 79.756 1
__rom Reactor Trip to I 100 seconds) 77.767 78t940 Affected SO Leakage Berore Trip IMm 20.319 t9084 ffected SO Leakage from ReactorTrip up to IS00 53.021 54,94S 5seccnds.
Ibm RCS to Affected SO Total Tube Leakage up to 1800 73.340 74,032
_ econds Ibm Totd Intact SG Steam Releascs vis ADV(From 1800 572026 572.026 seconds to 2 hl Ibm Total Intact SG Steam Releases via ADV (from 1800 1.479.S54 1.479,854 seconds to 8 hnl. Ibm MSSVs Cycle Until Operator Activates ADVs iB 100 Sec Srde 780-
Steam Generator Tube Rupture Analysis Dose Results l
EAB LZ Control Room Worst 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 30day 30-da mrem TEDE ren TEDE rein 1OE J0X Tube Pre-Acident Iodine PIlMring Sp ke 0.21 0 27 3.3 42% Te Pre-Accdent Paying lodine Spike 0.30 0.29 3JI ChUnte 7.1%
- 74.
14.4%
Acceptance Criteria 25 S
J30Y TWbe Concurrn Iodine PfVntoMn Spike C.0 0 aS0 9O' 42% We Concurrent Iodine Pta nr Spike 0.08 0.0 0394
_ CtCanve 0.0%
0.0%
2.2%
AcceptnceCrtitora 2.5 2.5 S
A value orO.96 rem fom the oiinrl AST analysis (13% SGP) was reported In the 20% submital since it was more conservatve than tez 30. tube plugging value.
Slide LOCA Analyses Evaluation Models used are the same as for 30% SGTP
- Sensitivity studies were performed to identify the limiting conditions and single failure.
- A spectrum of breaksizes (large and small) is analyzed to identify the limiting break size.
- Post LOCA long term cooling analysis is also re-performed
- The results show a reduction in PLHGR rate from 12.5 kw/ft to 12.0 kw/ft is needed.
Slide'
- Comparison of Input Parameters NAppendix K LOCA Analyses is performed at 102%
of Licensed Thermal Power of 2404 MWt (89% of 2700) per proposed TS footnote I
30%SGTPAnalyses l42%SGTPAnalyses Power Uncertainty 2% of 2700 MWt 2% of 2404 MWt
.r Large Break LOCA Preliminary Results Parameter 30% SGTP 42% SGTP Evaluation Model 1999EM Same Core power, %
102 % of Rated Power 102 % of 89%
Core Power, MMt 2754 2452 RCS Flow rate, gpm 335,000 300,000 Tcold (Min), F 532 Same PLHGR, wv/ft 12.5 12.0 Limiting Single Failure No Failure Same Safety Injection Performance Max. Performance Same Break Spectrum Results Limiting Break Size 0.6 DEG/PD 0.4 DEGIPD PCT, F 2130 2112 PLO, %
16.10 16.54 Core Wide Oxidation, %
< 1.0
< 1.0 Slide 82.
Small Break LOCA Preliminary Results Parameter 30% SGTP 42% SGTP Evaluation Model S2M Same Core power, %
102 % of Rated Power 102 % of 89%
Core Power, Mwt 2754 2452 RCS Flow rate, gm 335,000 300,000 Tcold (Max). F 552 Same PLHGR, kw/ft 13.0 Same Axial Shape Index (AS 1)
-0.15 at Rated Power
-0.15 at 89% Power Moderator Temperature 0.3 x 10' Same Coefficient at Initial Density, IF Limiting Single Failure Diesel Generator Same HPSI Performance Min.
Same SIT Credit No No for the Limiting Break Break Spectrum Results Limiting Break Size, fS 0.05 0.045 PCT, F 1943
<1600 PLO,%
9.80
<10.0 Core Wide Oxidation, _<1.0 LO f!'t.
Stde 8,"
Long Term Cooling Preliminary Results Parameter 30% SGTP 42% SGTP Evaluation Model CENPD-254 Same Core power, %
102 % of Rated Power 102% of 89%
Core Power, Mwt 2754 2452 RCS Cooldown Rate, F/hr
.75 Same Shutdown Cooling Entry 300 Same Temperature, F CST Inventory (Minimum),
262,400 Same Rallons Number of ADVs credited One per SG Same ADV Capacity, Ibm/hr at 55 psia 51,300 Same Decay Heat Multiplier 1.2 for< I000 Sec Same 1.1 for>1000 Sec Cooldown Results Acceptable LTC for the entire Yes Yes spectrum of break sizes SUde,4<0'
Post-LOCA Boric Acid Precipitation Analysis Post LOCA Boric Acid Precipitation analysis results are not affected by an increase in SGTP.
The effect of voids in the core on boricacid precipitation is addressed forSt. Lucie Unit 2.
Waterford EPU approach is used forthe analysis Slide 0
Post LOCA Boric Acid Precipitation Evaluation Approach Parameter Waterford Analysis St. Lucie Analysis Approach Approach Void fraction in the core Based on the CEFLASH-Same and upper plenum 4AS phase separation model Mixing Volume in core Core outlet plenum up to Core outlet plenum up to outlet plenum the top of hot leg elevation the bottom of hot leg in the upper plenum elevation in the upper plenum Credit of the Lower Plenum 509/%
Same for mixing -
Solubility Limit of the A solubility limit of 36 wtO/h A solubility limit of 27.6%
Boric Acid was used based on the is used (No credit of TSP is ternary solution of water, taken in calculating the boric acid and trisodium solubility limit) phosphate (TSP)
Mixing of Charging Credited Same (BAMT inventory) and Safety Injection (RWST inventory) in cold legs Slide 86
Post LOCA Boric Acid Precipitation Evi Preliminary Results fluation
- Analysis for 30% SGTP shows:
BoriccAdd concentration in the core may reach solubility limit at 7.8 hrs post-LOCA if hot and cold side injection is not initiated An initiation of hot and cold leg injection starting between 2 and 6 hrs post-LOCA prevents boric add concentration in the core from reaching the solubility limit The Preliminary results of the updated analysis show:
Boric Ad precipitation time increases when the effects of voids and compensating factors are considered.
The hot and cold leg initiation time between 2 and 6 hrs. remains valid WWW Sidee87 Oe E
Other Events Containment Peak Pressure Analysis Evaluation against the current analysis of record based on no increase in core AT is ongoing Break in Instrument Line or Others lines from the Reactor Coolant Pressure Boundary that Penetrate the Containment Noexplicitanalysis Event is insensitive to RCS vessel flow rate and tube plugging S-I;e
- I Dose Evaluation
- Preliminary Results of Dose Evaluation for all events (other than SGTR)
Doses for all events remain bounded by current UFSAR doses similar to 30%SGTP case.
,,,,p Slide 89 i
Dose Evaluation
.- Tnp 31.. Lk. 8kbpmd.s AMIyg.rRwI l%
ZZ pZoIG C-W by 53 Tb. L.Ip I%
DN3 L
Om AR Unit
_~tk ftui~ee LSm0rTWLa 500 PM-OIOPM VIaMr 1,
L d-AOR I.K-Ttp Swn Linc B-hius Au" OR Ed 41 pk.d so rub.
- 13J% CLM F
Pd fkil.u w Pmi.q Aal,4 I
e
,Wwfie Cu -
by 50 Td. Lo.
43%CLM L Qh AOR bi
]
(-bid. co..4 ~i~0.f
____G rT__
L___
1.0OPM OJ GPM "W
t-m Gn.,
27 1 oco In L., Ih. AOR fwd~nlff Lim Brk R
T A.
h" Ay6 of Plul 5tl TW-27AI Om Slie eo409
Other Topics PartialSubmittal by October 3, 2005
- Completed sectionswill include approximately 7of the Non-LOCA events.
Loss of flow, Loss of condenser vacuum, Asymmetric steam generator transient, Feedwater malfunction, CVCS malfunction, CEA withdrawal from a subcritical condition, Steam generator tube rupture
- Submittal will include framework for uncompleted sections 5Sde',l Other Topics Other Related LAR SG Tube Sleeving submittal Pressure drop across sleeves will increase by 65 psid.
The justification forthe increased pressure differential will be submitted as a part of the 42% SGTP LAR.
Slide91q'
Summary
- TS changes are minimal, mainly related to RCS flow.
- The analyses completed show compliance with all the acceptance criteria.
- Remaining analyses are dose to being completed and preliminary conclusions show no violation of criteria.
- All models and correlations remain applicable for the new analyzed conditions with no violation of parameter ranges.
- Final submittal will be made November 1, 2005
- Pre-submittal can be made by October 3, 2005 SU9,