ML030930068
| ML030930068 | |
| Person / Time | |
|---|---|
| Site: | Dresden, Quad Cities  |
| Issue date: | 01/31/2003 |
| From: | Richard Anderson, Casillas J, Pappone D General Electric Co |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| FOIA/PA-2005-0108, GENE-0000-0010-4202-01 R0 | |
| Download: ML030930068 (34) | |
Text
GE Nuclear Energy GeneralElectric Company 175 Curtner'Avenue, San Jose, CA 95125 GENE-0000-0010-4202-01 RO CLASS I January 2003 Engineering Evaluation of Impact on Transient and Safety Analyses of Reducing the Low Pressure Isolation Setpoint Analytical Limit to 785 psig Dresden Units 2 & 3 and Quad Cities Units 1 & 2 Prepared by: R.N. Anderson J.L. Casillas D.C. Pappone
GENE-0000-0010-4202-01 RO INFORMATION NOTICE This is a non-proprietary version of the document GENE-0000-0010-4202-01, which has the proprietary information removed. Portions of the document that have been removed are indicated by an open and closed bracket as shown here [ I IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT The only undertakings of GE respecting information in this document are contained in the contract between Exelon and GE for this work, and nothing contained in this document shall be construed as changing the contract. The use of this information by anyone other than Exelon for any purpose other than that for which it is intended, is not authorized; and with respect to any unauthorized use, GE makes no representation or warranty, and assumes no liability as to the completeness, accuracy, or usefulness of the information contained in this document, or that its use may not infringe privately owned rights.
ii
GENE-0000-00104202-01 RO Table of Contents ACRONYMS AND ABBREVIATIONS ......................................... . . v
- 1. Summary .............................................. 1
- 2. Introduction.................................................................................................................1
- 3. Evaluations..................................................................................................................1 3.1 ATWS Evaluation .............................................. 1 3.1.1 Impact on Nominal Condition (No Equipment OOS) ...................................... 4 3.1.1.1 Impact on Peak Vessel Pressure ............................................. 4 3.1.1.2 Impact on Peak Cladding Temperature ............................................. 4 3.1.1.3 Impact on Peak Local Cladding Oxidation ............................................. 5 3.1.1.4 Impact on Peak Suppression Pool Temperature ........................................ 5 3.1.1.5 Impact on Peak Containment Pressure .............................................. 5 3.1.1.6 Impact on Peak Drywell Temperature .............................................. 6 3.1.2 Impact of Equipment OOS ............................................. 6 3.1.2.1 Impact of TBV OOS ............................................. 6 3.1.2.1 Impact of One TCV Stuck Closed .............................................. 6 3.2 Transient Evaluation ............................................. . 9 3.2.1 Impact on PRFO ............................................. 20 3.2.2 Impact on Other Transients ............................................ 21 3.2.3 Impact on EOOS .............................................. 22 3.3 Impact on Accident Analyses . . 22 3.3.1 ECCS-LOCA Performance ............................................ 23 3.3.2 Containment System Response ........................................... 23 3.3.3 Subcompartment Pressurization .......................................... 24 3.3.4 Appendix R Fire Protection ............................................ 24 3.3.5 Station Blackout .............................................. 25 3.3.6 High Energy Line Break .............................................. 25 3.3.7 Radiological Consequences ............................................ 25 3.3.8 Operator Actions ............................................. 26 3.4 GEXL Evaluation ............................................. 26
- 4. Conclusions .............................................. 26 4.1 ATWS Events ............................................. 26 4.2 Transient Events..................................................................................................27 4.3 Current Operations . ............. 27 4.4 Accidents ................... 27 4.5 GEXL Application . . . 27 5.0 References ................... 28 iii
GENE-0000-010-4202-41 RO Table of Tables Table 1: Comparison of Limiting Results to ATWS Acceptance Criteria (Reference 1) .... 2 Table 2: Impact of EOOS on ATWS PRFO ................................................................ 3 Table 3: Summary of PRFO Evaluations ................................................................ 7 Table 4: Impact of Reducing the LPIS AL (825 to 785 psig) on Transient PRFO .............. 21 Table 5: Pressurization Response of Transients Considered .......................... ..................... 22 Table of Figures Figure 1: Neutron Flux versus Time, PRFO with 825 and 785 psig LPIS ......................................... 8 Figure 2: Integrated RV and SSV Steam Flow versus Time, PRFO with 825 and 785 psig LPIS ....9 Figure 3: Neutron Flux versus Time, PROF, 785 psig LPIS, 5 of 9 and 9 of 9 TBV ...................... 10 Figure 4: Integrated Steam Flow versus Time, PRFO, 785 psig LPIS, 5 of 9 and 9 of 9 TBV ....... 11 Figure 5- a: ATWS PRFO, 785 psig LPIS, BOC and 9 out of 9 TBV Active ............................. 12 Figure 5- b: ATWS PRFO, 785 psig LPIS, BOC and 9 out of 9 TBV Active .............................. 13 Figure 5- c: ATWS PRFO, 785 psig LPIS, BOC and 9 out of 9 TBV Active ............ .................. 14 Figure 5- d: ATWS PRFO, 785 psig LPIS, BOC and 9 out of 9 TBV Active .............................. 15 Figure 6- a: ATWS PRFO, 785 psig LPIS, BOC and 5 out of 9 TBV Active .......................... 16 Figure 6- b: ATWS PRFO, 785 psig LPIS, BOC and 5 out of 9 TBV Active .......................... 17 Figure 6- c: ATWS PRFO, 785 psig LPIS, BOC and 5 out of 9 TBV Active .......................... 18 Figure 6- d: ATWS PRFO, 785 psig LPIS, BOC and 5 out of 9 TBV Active .......................... 19 iv
GENE-0000-0010-4202-01 RO ACRONYMS AND ABBREVIATIONS Item, Form iShr Decipt<in 1 ATWS Anticipated Transient Without Scram 2 AL Analytical Limit 3 BOC Beginning of Cycle 4 CPR Critical Power Ratio 5 EOC End of Cycle 6 EOOS Equipment Out-Of-Service 7 EPU Extended Power Uprate 8 FWCF FeedWater Controller Failure 9 FWH FeedWater Heater(s) 10 GEXL GE Critical Quality Correlation 11 HPCI High Pressure Coolant Injection 12 IORV Inadvertent Opening of a Relief Valve 13 LFWH Loss of FeedWater Heater(s) 14 LOCA Loss Of Coolant Accident 15 LOFW Loss Of FeedWater 16 LOOP Loss Of-Offsite Power 17 LPIS Low Pressure Isolation Setpoint 18 LRNBP Load Rejection No ByPass 19 MCPR Minimum Critical Power Ratio 20 MSIV Main Steam Isolation Valve 21 MSIVC Main Steam Isolation Valve Closure 22 MSL Main Steam Line 23 OOS Out-Of-Service 24 PCT Peak Clad Temperature 25 PLU Power Load Unbalance 26 PRFO Pressure Regulator Failure Open 27 RPV Reactor Pressure Vessel 28 RV Relief Valve 29 RWE Rod Withdrawl Error
'30 SFRO Slow Flow Run-Out 31 SLMCPR Safety Limit MCPR 32 SLO Single Loop Operation 33 SRV Safety-Relief Valve 34 SRLR Supplemental Reload Licensing Report 35 SSV Spring Safety Valve 36 TBV Turbine Bypass Valve 37 TCV Turbine Control Valve 38 TTNBP Turbine Trip No ByPass v
GENE-0000-0010-4202-01 RO
- 1. Suniniary An evaluation, of the impact of reducing the LPIS (Low Pressure Isolation Setpoint) analytical limit from 825 to 785 psig, was performed for Dresden Units 2 and 3 and Quad Cities Units 1 and 2. This evaluation considered impact on ATWS, transient, and accident analyses and on application of GEXL at lower pressures. Impact on EOOS (Equipment Out-Of-Service) evaluations was included.
- 2. Introduction As part of the EPU, for Dresden Units 2 and 3 and Quad Cities Units 1 and 2, the turbine throttle pressure was reduced from approximately 950 psig to 912 psig. As a result, when doing turbine stop valve testing, the margin to the LPIS has been reduced. To reduce the possibility of reaching the LPIS during normal plant surveillance, Exelon has requested GE to evaluate the impact of reducing the LPIS analytical limit from 825 to 785 psig. This is to include all transient and safety licensing bases, including EOOS, for Dresden Units 2 and 3 and Quad Cities Units 1 and 2.
- 3. Evaluations 3.1 ATWS Evaluation Reference 1 documents ATWS evaluations of four events: These being the PRFO (Pressure Regulator Failure Open), MSIVC (Main Steam Isolation Valve Closure), LOOP (Loss Of-Offsite Power) and IORV (Inadvertent Opening of a Relief Valve). [
] Table 1 is a copy of the Reference 1 Table 1-1 "Comparison of Limiting Results to ATWS Acceptance Criieria",
with the addition of some of the less than limiting PRFO results. As in Reference 1, unless otherwise stated, all analyses are based upon an equilibrium core of GE14 fuel.
The evaluation will also include the impact, on the ATWS PRFO, of EOOS (Reference 2) and MSIV OOS (Reference 3). The EOOS and their disposition are listed in Table 2. Note that only two EOOS options have the potential to impact the ATWS PRFO, those being the TBV (Turbine Bypass Valves) OOS and the One TCV (Turbine Control Valve) Stuck Closed.
1
GENE-0000-0010-4202-01 RO Table 1: Comparison of Limiting Results to ATWS Acceptance Criteria (Reference 1)
Acceptance ATWS Event and Criteria Allowed Value Limiting Result' Conditions Peak Vessel 1500 1492 (26.8 sec) PRFO/BOC Pressure (psig) 1473 (25.9 sec) PRFO/EOC 1499 (26.7 sec) PRFO/BOC/Legacy Fuel2 Peak Cladding 1418 (57.4 see) MSIVC/EOC Temperature (° F) 2200 1416 (76.4 sec) PRFO/EOC 1335 (87.4 sec) PRFO/BOC Peak Local Cladding 17' Not Calculated 3 N/A Oxidation (%)
Peak Suppression 202 200 (1973 see) PRFO/EOC Pool Temperature 190 (1325 sec) PRFO/BOC
( F)
Peak Containment 62 16.5 (1973 sec) PRFO/EOC Pressure (psig) 13.0 (1325 sec) PRFO/BOC Peak Drywell N/A 311 PRFO/BOC Temperature (° F) . / . .
Notes: 1. Values in parentheses represent the time of peak values in seconds.
- 2. The Legacy fuel consisted of a core loading of GE14 (39%), ATRM9B (57%) and ANF9x9 (4%). The Legacy fuel case was not listed as the limiting case (in Reference
- 1) because it is expected that subsequent reloads will increase the fraction of GE14 in the core and the bounding peak vessel pressure will approach the equilibrium GE14 results.
- 3. Cladding oxidation is not explicitly calculated because the cladding temperature for ATWS is within the TASC application range that is below 1500 (°F). The TASC code does not model water-metal reaction since this reaction is insignificant below 1500
(°F).
2
GENE-0000-0010-4202-01 RO Table 2: Impact of EOOS on ATWS PRFO (Reducing Low Pressure Isolation Setpoint to 785 psig)
EOOS Disposition Justification TBV Potential Impact The state of the TBV OOS may be a combination of slower opening and/or reduced capacity. The reduced TBV capacity will change the depressurization trajectory (pressure versus time characteristic up to the time of dome pressure reaching the LPIS).
FWH No Impact When the feedwater heating is reduced, less power goes into producing steam (some of the power is required to heat up the cooler feedwater). With less steam production, the peak pressure will be reduced, i.e., the ATWS PRFO bounds the ATWS PRFO with FWH OOS.
TCV Slow Closure No Impact This has to do with the rate of TCV closure. The PRFO causes the MSIV to close.
SLO No Impact In SLO, the core power is limited to that defined by the intersection of the maximum core flow (with one active loop) and the maximum rod line. Originating the ATWS PRFO from this reduced power reduces the maximum pressure, i.e.,
bounded by ATWS PRFO initiated from 100% power.
One SRV Not Evaluated This EOOS is not included. ATWS events can't meet the 1500 psig limit when one SRV is OOS. This was addressed in both References 1and 2.
PLU No Impact This EQOS has to do with the timing of the scram in a load rejection. Not applicable to the ATWS event which assumes no scram.
One TCV Potential Impact Reduction of the system steam capacity (TCV plus TBV) may Stuck Closed impact the timing of the depressurization and thus the peak pressure.
One MSL No Impact While it is expected that the use of only three steam lines would change the dynamics of the depressurization slightly, this is l negligible compared to impact of starting the event from 75%
power (with one MSL OOS, the core power is limited to 75%/o),
i.e., starting the PRFO from 75% power would result in substantial peak pressure reduction.
Pressure No Impact The turbine control system has two pressure regulators. A Regulatorcontrolling pressure regulator and a backup regulator. If either fails in the open position, the TCV and TBV are opened to the maximum allowed by the MCFL. Because it only requires the failure of one pressure regulator to have a PRFO, the event is the same for the condition of either one or two pressure regulators in-service.
Note: 1. Pressure Regulator OOS addressed in Reference 26.
3
GENEO000-0010-4202-01 RO 3.1.1 Impact on Nominal Condition (No Equipment OOS)
Table 3 lists the peak vessel pressures for the PRFO events (1) shown in Table 1 (Reference
- 1) and (2) those evaluated for this report. The three nominal condition cases evaluated are the BOC and EOC GE14 fueled cores and the BOC Legacy fueled core. The nominal conditions are indicated by "9 TBV" in the "Case Description" column. This refers to all 9 TBVs being available, i.e., no TBV are OOS.
3.1.1.1 Impact on Peak Vessel Pressure Table 3 shows that reducing the LPIS from 825 to 785 psig has a negligible impact on the peak vessel pressure. For GE14 fueled core, at BOC the peak vessel pressure changes from 1492 to 1491 psig. At EOC, the change is 1473 to 1483 psig. For Legacy fuel, at BOC, the change is 1499 to 1497 psig. All of these peak pressures are within the allowed value of 1500 psig, i.e., the conclusion of Reference 1, that all ATWS events meet the 1500 psig criterion, remains valid for LPIS of 785 psig.
3.1.1.2 Impact on Peak Cladding Temperature While the most limiting PCT (see Table 1) occurs for the MSIVC (1418 'F), the PRFO (EOC) is quite close (1416 'F) and needed to be evaluated.
For the PCT criteria, conformance may be demonstrated by comparison (of the 825 and 785 psig LPIS PRFO transient responses) rather than by explicit calculation. To facilitate a meaningful comparison, a different time zero was defined. That being the time after dome pressure reaches the LPIS. This will ensure that time dependent transient characteristics will be reasonably in phase (the time dependent characteristics of PRFO, with 825 and 785 psig LPIS, will be moving together). The time for the dome pressure to reach the LPIS is listed in Table 3 under "Time of Isolation Initiation".
The 825 and 785 psig LPIS PRFO(EOC) neutron fluxes are shown in Figure 1. The time of PCT was approximately 63 seconds (76.4 [Table 1] - 13.0 [Table 3] = 63.4 sec). In the range of 60 to 65 seconds, it appears that the neutron flux, for the 785 psig LPIS case, is equal to or less than that of the 825 psig LPIS. This would lead to the conclusion that the PCT, for the 785 psig LPIS case, should be very close to that of the 825 psig LPIS. To confirm that this is 4
GENE-0000-0010-4202-01 RO indeed the case, an analysis was performed with the TASC computer code (same as used in Reference 1). The result was a PCT of 1414 'F versus 1416 'F for 825 psig LPIS PRFO(EOC). This confirms that neutron flux is a good predictor for transients with similar time dependent neutron flux shapes.
The conclusion, in Reference 1, that the PCT is below 1500 'F remains valid for the 785 psig LPIS.
3.1.1.3 Impact on Peak Local Cladding Oxidation Given that the PCT remains below 1500 'F, the Reference 1 conclusion that the amount of oxidation is insignificant remains valid. Thus there is no calculation of the cladding oxidation.
3.1.1.4 Impact on Peak Suppression Pool Temperature The analysis results for peak Suppression Pool Temperature, peak Containment Pressure and peak Drywell Temperature, all respond to the long-term integrated steam flows. For the Suppression Pool Temperature and Containment Pressure it is the integrated RV and SSV steam flow. For the Drywell Temperature is the integrated SSV flow. [
I The main input to the determination of the suppression pool temperature is the integrated steam flow through the RV and SSV. For the 825 and 785 psig LPIS PRFO (EGC), the integrated steam flow is plotted on Figure 2. [
] The conclusions of Reference 1, that the Peak Suppression Pool Temperature is less than the allowed value of 202 'F is also true for the condition of the LPIS being set at 785 psig.
3.1.1.5 Impact on Peak Containment Pressure For the same logic that shows that Peak Suppression Pool Temperature not changing, the Peak Containment Pressure will not change, that being because the integrated RV and SSV steam flows are unchanged. The conclusion of Reference 1, that the Peak Containment Pressure is below the allowed value of 62 psig, is also true for the condition of the LPIS being set at 785 psig.
5
GENE-0000-0010-4202-01 RO 3.1.1.6 Impact on Peak Drywell Temperature
[
] Thus, it is concluded that the Reference 1 conclusion for the Peak Drywell Temperature in not changed by reducing the LPIS from 825 to 785 psig. The Peak Drywell Temperature remains at 311 'F.
3.1.2 Impact of Equipment OOS As mentioned earlier (see Section 3.1), the only EOOS options that potentially impacted the PRFO are the TBV OOS and One TCV Stuck Closed. Both of these events are evaluated in the following sections.
3.1.2.1 Impact of TBV OOS TBV OOS means that one-or-more TBV are opening slower than Technical Specification requirements and/or some TBV are completely inoperative. For this study, only the inoperative valves are considered. It is assumed that a slow opening TBV is no different than one that opens within specification. [
] The evaluation only considers TBV capacities that coincide with full valves, i.e., 7 of 9 (total number of TBV is 9), 5 of 9, etc. Table 3 summarizes all PRFO analyses performed for this report.
The results show a weak sensitivity to TBV OOS (1491, 1488 and 1493 psig Peak Vessel Pressure for cases of 9, 7 and 5 TBV available). This scatter in the results is well within the accuracy of the model (ODYN). It is thus concluded that having TBV OOS has negligible impact on the peak vessel pressure results. In addition, a comparison of neutron flux and integrated RV and SSV steam flow is shown in Figures 3 and 4. [
/ ] It is concluded that the TBV OOS has no impact on the Reference 1 conclusions showing the ATWS results conforming to ATWS acceptance criteria (Table 1-1 of Reference 1).
The ATWS PRFO events with 9 of 9 and 5 of 9 TBV active, at BOC and LPIS of 785 psig, are shown in Figures 5-a through 5-d and Figures 6-a through 6-d.
3.1.2.1 Impact of One TCV Stuck Closed If one TCV is stuck closed, the total steam capacity, of the TCV plus TBV, is 111.6%. Note that the case with 3 TBV active has a total TCV+TBV capacity of 115.3% and does not result in isolation during a PRFO (insufficient capacity to depressurize the vessel enough to reach the LPIS of 785 psig). [ ]
6
GENE-0000-0010-4202-01 RO The one TCV stuck closed has a total steam capacity of 111.6% and thus will not depressurize, i.e., reducing the LPIS has no impact on the ATWS PRFO when one TCV is stuck closed.
Table 3: Summary of PRFO Evaluations Low Maximum Pressure Combined Time of Isolation Flow TCV+TBV Isolation Peak Vessel Setpoint Exposure . Limiter Capacity 3 Initiation4 Pressure (psig) Condition Case Description (%j) )( (sec) (psig) 8255 BOC 9 TBV 130 137.6 13.6 1492 8255 EOC 9 TBV 130 137.6 13.0 1473 8255 BOC 9 TBV, Legacy Fuel 130 137.6 13.0 1499 785 BOC 9 TBV 130 137.6 14.4 1491 785 BOC 7 TBV 130 130.2 14.6 1488 785 BOC 5 TBV 130 122.7 18.7 1493 785 BOC 3 TBV 130 115.3 No Pressurization 6 785 BOC 1 TBV 130 107.8 No Pressurization 6 785 EOC 5 TBV 130 122.7 17.8 1482 785 BOC 9 TBV, Legacy Fuel 130 137.6 14.3 1497 785 EOC 9 TBV 130 137.6 13.8 1483 Notes:
- -- . I . . .. .1 1 . d- c-- -
- 1. Unless otherwise stated, all analyses are based on an equillonum core oI %i4MuWei.
A P - .A1
- 2. This value was unchanged from the Reference I evaluation.
While References 24 and 25 show a 33.3% TBV capacity, the larger of the two was used.
- 4. Time the dome pressure reaches the LPIS (785 psig for this evaluation and 825 psig for Reference 1).
- 5. All cases with 825 psig LPIS are from Reference 1.
- 6. Total steam capacity insufficient to depressurize the vessel to the LPIS setpoint, i.e., with no isolation there is no pressurization.
7
GENE-0000-0010-4202-01 RO 150 125 100
-L U.
0 75 C,
z 50 25 0 20 40 60 80 100 120 Time After LPIS (sec)
Figure 1: Neutron Flux versus Time, PRFO with 825 and 785 psig LPIS
/
8
GENE-0000-0010-4202-01 RO 350000-300000X
. 250000 _
E 200000
'~150000-E ^ < l 9 825 psig l PIS (Referen ce 1)l 100000 E 785psigLPIS 9 of 9 TBV Available, EOC 50000 (Both Cases) 0 20 40 60 80 100 120 140 160 180 20C Time After LPIS (sec)
Figure 2: Integrated RV and SSV Steam Flow versus Time, PRFO with 825 and 785 psig LPIS 9
GENE-0000-0010-4202-01 RO 150 125 100 F
EL.
75 z
50 25 0
0 20 40 60 80 100 120 140 Time After LPIS (sec)
Figure 3: Neutron Flux versus Time, PROF, 785 psig LPIS, 5 of 9 and 9 of 9 TBV 10
GENE-0000-0010-4202-01 RO 350000-300000- c __
=^250000-U-
E 2 200000 ___
O 150000 cX / e 3 of 9 T8V,Availablel
.~E of TBV Available[
a PO9
- ~100000 ____
Jk ^ 785 psig LPIS. EOC 50000- / l(Both Cases) _
0-0 20 40 60 80 100 120 140 160 180 200 Time After LPIS (sec)
Figure 4: Integrated Steam Flow versus Time, PRFO, 785 psig LPIS, 5 of 9 and 9 of 9 TBV 11
GENE-0000-0010-4202-01 RO ISO 100 cry UL-LU 50.
LU C0:
LLJ Cl O. L-O. 25. 50. 75. 100.
RNA 010303 OOAC5
- 1. L TIME (SEC)
Figure 5- a: ATWS PRFO, 785 psig LPIS, BOC and 9 out of 9 TBV Active I 12
GENE-0000-0010-4202-01 RO CORE PRESSURE... ...psia 1 DOME PRESSURE... ..psia 4 STEAULINE FRES N ) .psla I1.
rNk STEAMUINE PRES N) 6.psia I frUM NM 01 mxi C .. nay xi101 22 I.
5 a 5
I 4I 0\I, c
en
- 0. 25. 50. 75. 100.
RNA 010303 OXAOS 1=.1 TIME (SEC)
Figure 5- b: ATNVS PRFO, 785 psig LPIS, BOC and 9 out of 9 TBV Active 13
GENE-0000-0010-4202-01 RO 1 ACTUAL LEVEL.... ..... ft Z NR SENSED LEVEL. ..... ft 1 WR SENSED LEVEL. ..... ft
, CORE BORON CONC. ppm/100 l
- 15. l
- 5. I I II a I Ia
-5.
- I I I I I I_ _ _ __ _ _ __I_ __ _ _ _ __I_ _ _ _ _ _ __I_ _ _
-15.
00 25. 50. 7s. 100.
R14& OOCAC cI TIME (SEC) 01o0o" I=.
Figure 5- c: ATWS PRFO, 785 psig LPIS, BOC and 9 out of 9 TBV Active 14
GENE-0000-0010-4202-01 RO
- 10. *INTE. SRV FLO 100.
c .
50.
LIJ
- 0. so
- 5. 75. 100.
010501=.lTIME (SEC)
Figure 5- d: ATWS PRFO, 785 psig LPIS, BOC and 9 out of 9 TBV Active 15
GENE-0000-0010-4202-01 RO 150.
1-u.
cr-Lj-C> 50.
w 0-,
- 0. I
- 0. 25. 50. 75. 100.
RW O12I03 ODLESL O7Q5.0 TIME (SEC)
Figure 6- a: ATWS PRFO, 785 psig LPIS, BOC and 5 out of 9 TBV Active 16
GENE-0000-0010-4202-01 RO I .
1.2 1.
0.
- 0. 25. 50. 75. 100.
RNA 0121G O0ES1 07V5.O TIME (SEC)
Figure 6- b: ATWS PRFO, 785 psig LPIS, BOC and 5 out of 9 TBV Active i 17
GENE-0000-0010-4202-01 RO ACTUAL LEVEL.... ..... ft 1 NR SENSED LEVEL. ..... ft WR SENSED LEVEL. ..... ft 14 CORE BOREN CONC. pprn/100
- 15. I I I 5.
2 1
-5.
I_ _L 2 _ __ _ _ _ __ _ _ _ _ __
__ _ _2__ _ _a__ _ _ _ _2__ __ _
Lo LL
-15.
0~ 25. 50. 75. 100.
RW 012iO OOIEisj 0705.
TIME (SEC)
Figure 6- c: ATWS PRFO, 785 psig LPIS, BOC and 5 out of 9 TBV Active 18
GENE-0000-0010-4202-01 RO 150. -
100. .0 a_- 50. -
LU_
C-,
LU_
L_
- 0. -
- 0. 25. 50. 75. 100.
RW OOJESL 01210O 07o5.0 TIME (SEC)
Figure 6- d: ATWS PRFO, 785 psig LPIS, BOC and 5 out of 9 TBV Active 3.2 Transient Evaluation The transients considered here are the PRFO and the events listed in Reference 4 and evaluated in Reference 5. The PRFO was not included in the EPU evaluations because it is a power decrease event in which fuel thermal margins increase during the event. However, changing the LPIS AL will impact the PRFO and thus it is considered here. All transients evaluated, will consider the affect of EOOS (same OOS items as shown in Table 2).
19
GENE-0000-0010-4202-01 RO 3.2.1 Impact on PRFO The primary design basis for the Main Steam Line LPIS is the vessel cooldown requirement
(<1 00F/hr) under a PRFO event (Dresden and Quad Cities Technical Specification Basis Section 3.3.6.1). A secondary design basis is for the LPIS is to meet the requirement of the SLMCPR (see Sections 3.4 and 4.5 of this report and the Dresden and Quad Cities Technical Specification Basis Section 2.1.1.1).
For this event with the 825 psig LPIS AL, assurance that the RPV cooldown limit (100F/hr) is not exceeded was provided by the closure of the MSIVs. [
] The combination of this small temperature difference (60 F) for the small duration (-5 sec) results in a negligible change in the vessel cooldown rate, i.e., with the 785 psig LPIS AL, the closure of the MSIVs assures that the RPV cooldown limit is not reached.
With regards to thermal limits, the transient PRFO is a very mild event. The peak neutron flux and heat flux occurs at time zero, i.e., the neutron and heat fluxes never increase above their initial values. The peak pressure occurs after the LPIS trip. The peak pressure is bounded by the full power MSIVC direct scram.
The transient PRFO is identical to the ATWS PRFO up to the time at which the LPIS AL is reached. At this point the transient PRFO has a scram initiation from the closure of the MSIV position switches (usually referred to as a direct scram). Subsequent to the isolation and scram, the PRFO transient is basically over, with the exception of the minor pressurization that results from the MSIVC.
To determine the impact of reducing the LPIS AL, on depressurization and water level, compare the ATWS PRFO responses of the original LPIS and that with the reduced LPIS.
This is accurate for time zero up to the time of the isolation. Subsequent to the isolation, the pressure and level responses are similar to that of the transient PRFO with 825 psig LPIS.
Table 4 summarizes the impact on the transient PRFO.
I 20
GENE-0000-0010-4202-41 RO Table 4: Impact of Reducing the LPIS AL (825 to 785 psig) on Transient PRFO Item Characteristic Impact of Reducing LPIS (825 to 785 psig) 1 Vessel Cooldown Rate Negligible impact, MSIVC ensures RPV temperature change limit (100F/hr) not reached.
2 Pre-isolation Pressure and Compare ATWS PRFO for 825 and 785 psig LPIS AL Level response 3 Post-isolation Pressure Similar to transient PRFO with 825 psig LPIS AL, response except that peak pressure will be slightly lower (results because the isolation occurs at lower power than the 825 psig case).
4 Post-isolation Water Level Similar to transient PRFO with 825 psig LPIS AL.
response 3.2.2 Impact on Other Transients All of the transients, listed in References 4 and 5, have been considered. Because none of these events result in a depressurization, there is no impact on any of these events. Table 5 summarizes the pressurization responses for Reference 4 and 5 transients considered.
In addition, all transients in Chapter 15 of the Dresden and Quad Cities UFSARs have been reviewed to see if any transient, other than those listed in Table 5, could be impacted by the reduction in the LPIS, from 825 to 785 psig. The conclusion is that there are no other events impacted by this change.
21
GENE-0000-0010-4202-O1 RO Table 5: Pressurization Response of Transients Considered Item Eventl Pressurization Small Pressure Depressurization
__ _ __. _._._-_ Change 1 LRNBP X _ -
2 TTNBP X - _
3 FWCF X _ _
4 LFWH _ X -
5 Inadvertent HPCI Start - X -
6 RWE - X -
7 Slow Recirculation Increase X -
8 Fast Recirculation Increase X -
9 LRWBP X _ -
10 MSIVC, One Valve and All X _
Valves 11 MSIVC-Flux Scram X _
12 TTNBP-Flux Scram X _
13 LOFW X _
14 Loss of 1 Feedwater Pump' X Note: 1. Transients described and analyzed in References 4 and 5.
3.2.3 Impact on EQOS Section 3.2.2 established that the only transient impacted by the reduction in LPIS, from 825 to 785 psig, is the PRFO. Thus, this Section will only consider the impact of the reduction in LPIS and the inclusion of the EOOS options, on the PRFO.
In Section 3.1 it was concluded that the only EOOS, that would affect the ATWS PRFO was the TBV OOS. Likewise with the transient PRFO, only the TBV OOS has the potential to impact the results. To determine the impact of EOOS, it is necessary to compare the transient PRFO TBV OOS with a LPIS of 825 and 785 psig. [
] Thus, the peak pressure of the 785 psig LPIS case will always be less than that of the case with 825 psig LPIS, i.e., for any given EOOS, the transient PRFO will have a smaller peak pressure with the 785 psig LPIS.
3.3 Impact on Accident Analyses The pipe break accident analyses and other safety events were evaluated for the impact of reducing the MSIV closure low pressure isolation setpoint. The events evaluated were the ECCS-LOCA performance, containment system response, subcompartment pressurization, Appendix R fire protection, station blackout, high energy line breaks, and the radiological consequences resulting from the pipe break accidents. This evaluation is based on a review of the analyses performed for the extended power uprate. None of the accident or safety event analyses reviewed take credit for the low pressure isolation trip. Since many of the 22
GENE-0000-0010-4202-01 RO analyses assume MSIV closure at the beginning of the event and the low pressure isolation setpoint is not used, the evaluation for each event also considered the system response if the MSIVs are not closed at the beginning of the event. The purpose of this part of the event evaluation is to determine if the low pressure isolation trip would be used and if a change in the setpoint would affect the system response. Based on these evaluations, it was concluded that a reduction in the MSIV low pressure isolation trip setpoint will not have an adverse impact on the plant accident analyses, on the plant system response to accidents, or on the radiological consequences of such events.
3.3.1 ECCS-LOCA Performance The ECCS-LOCA performance analyses in Reference 6 consider pipe breaks inside and outside the containment. The MSIVs are assumed to close at the start of the accident for all break locations in the analyses. Therefore, the low pressure isolation trip is not used in the LOCA analyses and the LOCA analysis results are not affected by the reduction in the low pressure isolation setpoint.
If the MSIV closure does not occur at the beginning of the accident, the pressure regulator will close the turbine control and bypass valves in an attempt to maintain the turbine throttle pressure at approximately 925 psig (References 22, 23, 24 and 25) following reactor scram.
Thus, for events other than breaks in the main steamline, the main steamline will effectively become isolated before the low pressure isolation setpoint is reached. For large breaks in the main steamline (inside and outside the containment), the MSIV closure is initiated by a high steamline flow signal at the beginning of the event; well before the low pressure isolation setpoint is reached. For these cases, the ECCS performance is not affected by the reduction in the low pressure isolation setpoint.
If the steamline break is too small to result in a high flow isolation signal, MSIV closure may be initiated by another signal (e.g., high steamline tunnel temperature, low reactor water level) or it may occur due to the low pressure isolation trip. In either case, steamline breaks of any size are not limiting events with respect to ECCS performance and a 40 psi reduction in the low pressure isolation setpoint will not affect compliance with the acceptance criteria of 10 CFR 50.46.
Based on the above discussions, the reduction of the MSIV low pressure isolation setpoint has no impact on the plant response to a LOCA or on compliance with the acceptance criteria of 10 CFR 50.46 and the conclusions of Reference 6 are unchanged.
3.3.2 Containment System Response The containment system response to pipe breaks inside the containment was analyzed in Reference 7. The MSIVs are assumed to close at the start of the accident for all break locations in the containment system response analyses. Therefore, the low pressure isolation trip is not used in the containment system response analyses and the analysis results are not affected by the reduction in the low pressure isolation setpoint.
23
GENE-0000-0010-4202-01 RO If the MSIV closure does not occur at the beginning of the accident, the pressure regulator will close the turbine control and bypass valves in an attempt to maintain the turbine throttle pressure at approximately 925 psig following reactor scram. Thus, for events other than breaks in the main steamline, the main steamline will effectively become isolated before the low pressure isolation setpoint is reached. For large breaks in the main steamline, the MSIV closure is initiated by a high steamline flow signal at the beginning of the event; well before the low pressure isolation setpoint is reached. For these cases, the containment system response is not affected by the reduction in the low pressure isolation setpoint.
If the steamline break is too small to result in a high flow isolation signal, MSIV closure may be initiated by another signal (e.g., low reactor water level) or it may occur due to the low pressure isolation trip. [
] These breaks are large enough to depressurize the reactor regardless of the MSIV closure.
Therefore, a 40 psi reduction in the low pressure isolation setpoint will not affect the peak drywell shell temperature or the drywell temperature EQ envelope.
Based on the above discussions, the reduction of the MSIV low pressure isolation setpoint has no impact on the containment system response.
3.3.3 Subcompartment Pressurization The subcompartment pressurization analysis for a pipe break in the annulus between the reactor vessel and the shield wall is documented in Reference 8. The break mass and energy release used in this evaluation are based on the steady-state reactor operating conditions.
Therefore, the low pressure isolation trip is not used in the subcompartment pressurization analyses. In addition, the peak annulus pressurization loads occur at the beginning of the event (within approximately the first second), well before MSIV closure can occur.
Therefore, the subcompartment pressurization results are not affected by the reduction in the low pressure isolation setpoint.
3.3.4 Appendix R Fire Protection The reactor system response for the Appendix R fire protection analysis is documented in References 9 and 10. The sequences of events for these analyses (Tables 3-4 and 3-2, respectively) show that closure of the MSIVs is initiated on low-low reactor water level.
However, the steam flow is stopped prior to the MSIV closure on low-low reactor water level when the turbine control valves close on low inlet pressure following reactor scram.
The main steamline will effectively become isolated before the low pressure isolation setpoint is reached. Therefore, the reduction of the MSIV low pressure isolation setpoint has no impact on the reactor system response to an Appendix R fire protection event.
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GENE-000OO-10-4202-01 R0 3.3.5 Station Blackout The reactor system response for a station blackout is documented in References 11 and 12.
The initiating event, the loss of offsite power, results in MSIV closure at the beginning of the event. Therefore, the reduction of the MSIV low pressure isolation setpoint has no impact on the reactor system response during a station blackout.
3.3.6 High Energy Line Break The high energy line break (HELB) evaluations are documented in Reference 13. The steamline break is the only break potentially affected by the change in the low pressure isolation setpoint. The steamline break HELB analysis assumes that MSIV closure is initiated by a high steamline flow signal at the beginning of the event. Therefore, the low pressure isolation trip is not used in the HELB analyses and the HELB analysis results are not affected by the reduction in the low pressure isolation setpoint.
The steamline break case determines the short-term peak temperature in the steam tunnel.
There is a limited range of break sizes for which the low pressure isolation trip has the potential to initiate MSIV closure. In order for the low pressure isolation trip to generate the MSIV closure, the break must be big enough to depressurize the vessel below the pressure regulator setpoint of approximately 925 psig, but small enough to avoid the high steamline flow trip. Within this range of break sizes, the reduction in the low pressure isolation setpoint would delay the isolation, resulting in an increase in the mass and energy release.
] Therefore, a 40 psi reduction in the low pressure isolation setpoint will not affect the peak temperature in the steam tunnel. If the break is too small to depressurize the vessel, the peak temperature in the steam tunnel will be determined by the steamline pressure conditions and will not be affected by the low pressure isolation setpoint.
Based on the above discussions, the reduction of the MSIV low pressure isolation setpoint has no impact on the HELB response for a steamline break.
3.3.7 Radiological Consequences The radiological consequences due to a break of the main steamline outside the containment are documented in References 14 and 15. The MSIVs are assumed to close on high steam flow at the start of the accident in the analyses. Therefore, the low pressure isolation trip is not used in the mass release analyses and the radiological consequences are not affected by the reduction in the low pressure isolation setpoint.
If the steamline break is too small to result in a high flow isolation signal, MSIV closure may be initiated by another signal (e.g., high steamline tunnel temperature, low reactor water level) or it may occur due to the low pressure isolation trip. [
25
GENE-0000-0010-4202-01 RO
] Therefore, a 40 psi reduction in the low pressure isolation setpoint will not affect the radiological consequences.
Based on the above discussions, the reduction of the MSIV low pressure isolation setpoint has no impact on the radiological consequences of a main steamline break outside the containment.
3.3.8 Operator Actions The accident analyses do not take any credit for operator action with respect to MSIV closure. However, the emergency procedures (Reference 16, Step RC/Q-1) direct the operator to immediately place the reactor mode switch in SHUTDOWN. [
] Therefore, a reduction in the low pressure isolation setpoint will have no impact on the reactor system response when considering the effects of operator actions.
3.4 GEXL Evaluation One of the design bases for the Main Steam Line LPIS is to meet the requirement for the SLMCPR during normal operation conditions (Dresden and Quad Cities Technical Specification Basis Section 2.1.1.1). The standard GE GEXL correlation range of application is given in Section 2.8.3 of the GE14 fuel Amendment 22 document (Reference 17). [
] Furthermore, the GEXL correlation is applied at the local fluid conditions in the fuel bundle. An LPIS Analytical Limit value of 785 psig will ensure that the GEXL correlation is applied within its approved range because the pressure at the fuel location is several psi higher due to several feet of submergence and the flow pressure drop through the steam separators. l
- 4. Conclusions 4.1 ATWS Events The only ATWS event impacted by the reduction of the LPIS, from 825 to 785 psig, is the PRFO. All other ATWS events never decrease pressure sufficiently to reach the LPIS. For the ATWS PRFO, reduction of the LPIS from 825 to 785 psig and the inclusion of EOOS, results in insignificant change to the peak vessel pressure. The conclusions of Reference 1 are unchanged for all of the ATWS Acceptance Criteria (see Table 1-1 of Reference 1).
26
GENE-0000-0010-4202-01 RO 4.2 Transient Events The transient events considered were the PRFO and the events listed in Table 4 (see References 4 and 5). For the transient PRFO, reduction of the LPIS from 825 to 785 psig and the inclusion of EOOS, results in minor changes to the event. The RPV cooldown limit (100'F/hr) is not reached. For the remaining transients considered, reduction of the LPIS from 825 to 785 psig and the inclusion of EOOS have no impact (none of the events depressurize the vessel).
4.3 Current Operations The current SRLRs (References 18, 19, 20 and 21) for Dresden Units 2 and 3 and Quad Cities Units 1 and 2 have been reviewed and it is concluded that the reduction of the LPIS, from 825 to 785 psig changes none of their conclusions.
4.4 Accidents The pipe break accident analyses and other safety events were evaluated for the impact of reducing the MSIV closure low pressure isolation setpoint. The events evaluated were the ECCS-LOCA performance, containment system response, subcompartment pressurization, Appendix R fire protection, station blackout, high energy line breaks, and the radiological consequences resulting from the pipe break accidents.
The purpose of the MSIV low pressure isolation is to prevent rapid vessel cooldown during certain transient events. The low pressure isolation trip is not used in the accident analyses.
[
] Therefore, a reduction in the MSIV low pressure isolation trip setpoint will not have an adverse impact on the plant accident analyses, on the plant system response to accidents, or on the radiological consequences of such events.
4.5 GEXL Application The Technical Specification requirements for an Analytical Limit (AL) minimum pressure of 785 psig are met by both the GE GEXL CPR correlation and the low flow critical power data. Therefore, a minimum vessel pressure of 785 psig is appropriate for GE fuel in Dresden and Quad Cities.
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GENE-0000-0010-4202-01 RO 5.0 References
- 1. Dresden and Qtad Cities Extended Power UprateEvaluation Task T0902: Anticipated Transient Without Scram, GE-NE-A22-00103-11-01 Rev. 2, February 2002.
- 2. Dresden 2 and 3 Quad Cities 1 and 2 Equipment Out-Of-Service and Legacy Fuel TransientAnalysis, GE-NE-J 11-03912-00-01-Rl, November 2001.
- 3. Evaluation ofLimiting Events with One Main Steam Line Out-Of-Servicefor Dresden, Units 2 and 3 and Quad Cities Units I and 2, NSA 02-350, July 11, 2002.
- 4. Licensing Topical Report, Generic Guidelinesfor GeneralElectricBoiling Water Reactor Extended Power Uprate, NEDC-32424P-A, Class III, February 1999 (ELTR-1).
- 5. Dresden and Quad Cities Extended Power Uprate Task T0900: TransientAnalysis, GE-NE-A22-00103-10-01 Rev. 0, October 2000.
- 6. NEDC-32990P Rev. 1, "SAFER/GESTR-LOCA Loss-of-Coolant Accident Analysis for Dresden Nuclear Station 2 & 3 and Quad Cities Nuclear Station 1 & 2," September 2001.
- 7. GE-NE-A22-00103-08-01 Rev. 1, "Project Task Report, Dresden and Quad Cities Extended Power Uprate, Task T0400: Containment System Response," December 2000.
- 8. GE-NE-A22-00103-31-01 Rev. 0, "Project Task Report, Dresden and Quad Cities Extended Power Uprate, Task T040 1: Subcompartment (Annulus) Pressurization Loads," October 2000.
- 9. GE-NE-A22-00103-56-01-D Rev. 1, "Project Task Report, Dresden and Quad Cities Extended Power Uprate, Task T061 1: Appendix R Fire Protection (Dresden Station),"
January 2001.
- 10. GE-NE-A22-00103-56-01-Q Rev. 1, "Project Task Report, Dresden and Quad Cities Extended Power Uprate, Task T06 11: Appendix R Fire Protection (Quad Cities V Station)," January 2001.
- 11. GE-NE-A22-00103-75-01 Rev. 0, "Project Task Report, Dresden and Quad Cities Extended Power Uprate, Task T0903: Station Blackout (Dresden Station)," October 2000.
- 12. GE-NE-A22-00103-75-02 Rev. 0, "Project Task Report, Dresden and Quad Cities Extended Power Uprate, Task T0903: Station Blackout (Quad Cities Station)," October 2000.
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GENE-0000-0010-4202-41 RO
- 13. GE-NE-A22-00103-66-01 Rev. 0, "Project Task Report, Dresden and Quad Cities Extended Power Uprate, Task T1000: HELB Mass and Energy Release Evaluation,"
October 2000.
- 14. NEDC-32962P Rev. 2, "Safety Analysis Report for Dresden 2 & 3 Extended Power Uprate," August 2001.
- 15. NEDC-32961P Rev. 2, "Safety Analysis Report for Quad Cities 1 & 2 Extended Power Uprate," August 2001.
- 16. BWR Owners' Group Emergency Procedure and Severe Accident Guidelines Rev. 1, July 1997.
- 17. NEDC-32868P Revision 1, GE14 Compliance with Amendment 22 of NEDE-2401 1-P-A (GESTAR II), September 2000.
- 18. "Supplemental Reload Licensing Report" for Dresden Unit 2 Reload 17 Cycle 18, JI 1-03837-SRLR-2957 Rev. 0, September 2001.
- 19. "Supplemental Reload Licensing Report" for Dresden Unit 3 Reload 17 Cycle 18, 0000-0006-9848-SRLR Rev. 1 August 2002.
- 20. "Supplemental Reload Licensing Report" for Quad Cities Unit 1 Reload 17 Cycle 18, 0000-0009-5864-SRLR Rev. 2, October 2002.
- 21. "Supplemental Reload Licensing Report" for Quad Cities Unit 2 Reload 16 Cycle 17, JI 1-03918-SRLR Rev. 1, December 2001.
- 22. "OPL-3 Parameters for Dresden Unit 3 Cycle 18 Transient Analysis", NF2002-9994, to Cheryl Collins (GNF), from William Matos, Jr., April 5, 2002.
- 23. "OPL-3 Parameters for Dresden Unit 2 Cycle 18 Transient Analysis", NFMO100057 SEQ 00, to Tammy Orr (GNF), from Hossein Youssefnia, May 10, 2001.
- 24. "OPL-3 for Quad Cities Unit 1llCycle 18", TODI No. QDC-02-018, to Russ Lindquist (GNF), from David F. Schumacher, April 25, 2002.
- 25. "OPL-3 for Quad Cities Unit 2 Cycle 17 Transient Analysis", NFM 0100103 SEQ 00, to Tammy Orr (GNF), from John M. Freeman, September 25, 2001.
- 26. "Dresden and Quad Cities Pressure Regulator OOS", letter FRL02EX-014, to Jim Nevling, from F. Russel Lindquist, October 22, 2002.
29