ML20134N202
| ML20134N202 | |
| Person / Time | |
|---|---|
| Site: | Calvert Cliffs  |
| Issue date: | 11/20/1996 |
| From: | Cruse C BALTIMORE GAS & ELECTRIC CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9611260217 | |
| Download: ML20134N202 (83) | |
Text
_
CurrLES II. CRUSE Baltimore Gas and Electric Company Vice President Calvert Cliffs Nuclear Power Plant Nuclear Energy 1650 Calvert Cliffs Parkway Lusby, Maryland 20657 410 495-4455 November 20,1996 U. S. Nuclear Regulatory Commission Washington,DC 20555 ATTENTION:
Document Control Desk
SUBJECT:
Calvert Cliffs Nuclear Power Plant Unit Nos.1 & 2; Docket Nos. 50-317 & 50-318 Request for Additional Information - Change to the Moderator Temperature Coefficient
REFERENCE:
(a)
Letter from Mr. C. H. Cruse to NRC Document Control Desk, dated March 28, 1996, License Amendment Request:
Change to the Moderator Temperature Coefficient In response to verbal NRC Staff questions. Baltimore Gas and Electric Company is providing the additional information requested. The information is provided in Attachment (1). The additional information does not alter the Determination of No Significant Hazards submitted in the referenced letter.
Qh r
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s 9611260217 961120 PDR ADOCK 05000317 P
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t' Documsnt Control Desk November 20,1996 Page 2 1
Should you have additional questions regarding this matter, we will be pleased to discuss them with you.
Very truly yours,
't sbW M
STATE OF MARYLAND
- TO WIT:
1 COUNTY OF CALVERT I hereby certify that on the AO-Ot) day of M#l#AMlflAI.19[4 before me, the subscriber, a Notary Public of the State of Maryland in and for (A /id/r/ > //11Mfm
. personally appeared Charles 11. Cruse, being duly sworn, and stateitliat he is Vice Pres'ideg of the Baltimore Gas and Electric Company, a corporation of the State of Maryland; that he provides the foregoing response for the purposes therein set forth; that the statements made are true and correct to the best of his knowledge, information, and belief; and that he was authorized to provide the response on behalf of said Corporation.
WITNESS my liand and Notarial Seal:
1/Mid Notary Public My Commission Expires:
b date CllC/ PSF / dim Attachment (1)
Baltimore Gas and Electric Company's Response to NRC's Request for Additional Information Concerning License Amendment Request: Moderator Temperature Coefficient Change Appendix A - Transient Data cc:
A. W. Dromerick, NRC (Without Appendix A)
D. A. Brune, Esquire Resident Inspector, NRC J. E. Silberg, Esquire R. I. McLean, DNR Director, Project Directorate I-1, NRC J. II. Walter, PSC
- 11. J. Miller, NRC
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i ATTACHMENT (1) 1
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BALTIMORE GAS AND ELECTRIC COMPANY'S RESPONSE TO NRC'S REQUEST FOR ADDITIONAL INFORMATION CONCERNING LICENSE AMENDMENT REQUEST: MODERATOR TEMPERATURE
)
COEFFICIENT CHANGE i
I Calvert Cliffs Nuclear Power Plant November 20,1996
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I ATTACHMENT (1)
BALTIMORE GAS AND ELECTRIC COMPANY'S RESPONSE TO NRC'S REQUEST FOR ADDITIONAL INFORMATION CONCERNING LICENSE AMENDMENT REQUEST: MODERATOR TEMPERATURE COEFFICIENT CIIANGE Ouestion 1:
Identify the transients and accidents that are not anected by increasing the number ofplugged steam generator (SG) U-tubes. Providejustipcation as to why these events are not affected.
Response
The limiting safety analyses for the following transients and accidents are not affected by increasing the number of plugged SG U-tubes:
Excess Load CEA Eject Loss of Feedwater Flow (SG Dryout)
Main Steam Line Break Excess Feedwater Heat Removal Steam Generator Tube Rupture Reactor Coolant System (RCS) Depressurization Seized Reactor Coolant Pump (RCP) Rotor Loss ofNon-Emergency AC Power Excess Charging Control Element Assembly (CEA) Drop The limiting safety analyses for these events are not affected since plugging SG U-tubes either would have no effect on the analysis results, or improve the results.
The r. umber of plugged SG U-tubes will be controlled by our reload design process such that the minimum RCS flow assumed in the limiting safety analyses for these events is maintained at 370,000 gpm. The steady state full power RCS temperatures (Tuor and TAVE) are therefore not affected.
As a result, the primary effect of increasing the number of plugged SG U-tubes, with respect to these events, is to reduce the SG heat transfer area.
Excess Load. Excess Feedwater Heat Removal. and Main Steam Line Break All of these events are initiated by an increase in RCS heat removal by the SGs. The effect of plugging SG U-tubes, while maintaining constant RCS flow, is beneficial for these events. The net effect of plugged tubes is to reduce the available heat transfer area of a SG. This would result in a slower rate of RCS heat removal than assumed in these safety analyses. Therefore, the limiting safety analyses for these events are not affected.
Loss of Feedwater Flow (SG Dryout)
The time to SG dryout is dependent on the available SG water inventory at the SG low level trip setpoint and core decay heat. Plugging SG U-tubes does not affect the available SG inventory or decay heat assumed in the limiting safety analysis for this event, and this analysis is therefore not affected.
RCS Depressurization. Loss of Non-Emergency AC Power. CEA Dron. and Seized RCP Rotor These events involve approach to departure from nucleate boiling (DNB) and/or linear heat rate (LHR) limits as a result of changing RCS pressure, flow, or core power distribution. Since plugging SG U-tubes within the limits established and controlled by our reload design process will not affect the limiting RCS parameters (flow, temperature, and pressure) or the limiting power distribution, and other core 1
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ATTACHMENT m BALTIMORE GAS AND ELECTRIC COMPANY'S RESPONSE TO NRC'S REQUEST FOR ADDITIONAL INFORMATION CONCERNING LICENSE AMENDMENT REQUEST: MODERATOR TEMPERATURE COEFFICIENT CilANGE characteristics assumed in the limiting safety analyses, these analyses are not affected. The seized RCP rotor analysis assumea a step change in RCS flow rate when the RCP rotor seizes, and so the flow transient assumed for this event is not affected by the increased RCS flow resistance due to plugging SG U-tubes.
CEA Elect Accident This event is a rapid reactivity insertion coupled with a rapid RCS depressurization. Again, plugging SG U-tubes will not affect the RCS or core parameters assumed in the limiting safety analysis. This analysis is therefore not affected.
Steam Generator Tube Ruoture This event involves offsite dose due to the release of activity from the affected SG. The amount of activity released is a function of the ruptured tube leak rate and subsequent steam release rates offsite.
The number of plugged SG U-tubes allowed by internal controls will not cause SG steam pressure to decrease below that assumed in the limiting safety analysis. Therefore, the ruptured tube leak rate and steam release rates are not affected, and neither is this analysis.
j Excess Charging This event involves increasing RCS pressure as a result ofinadvertent charging system operation. The time available for operators to respond to the event is determined by the available pressurizer bubble volume and the charging system capacity. Plugging SG U-tubes will not affect these parameters, and the limiting safety analysis for this event is not affected.
Containment Resnonse to Loss-of-Coolant Accident and Main ' Steam Line Break These events result in rapid pressurization of the containment due to the mass and energy blowdown from the RCS and SGs. Plugging SG U-tubes slows the energy transfer between the RCS and SGs, which has a beneficial effect on the results of the limiting analyses for these events. In addition, plugging SG U-tubes reduces RCS inventory slightly which has an additional beneficial effect on the loss-of-coolant accident (LOCA) analysis. Again, since RCS and SG parameters for initial flow, temperature, and pressure assumed in the limiting analyses for these events are not affected by plugging i
SG U-tubes within the limits established and controlled by the reload design process, these limiting l
analyses are not affected.
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AlTACHMENT (1)
BALTIMORE GAS AND ELECTRIC COMPANY'S RESPONSE TO NRC'S REQUEST FOR ADDITIONAL INFORMATION CONCERNING LICENSE AMENDMENT REQUEST: MODERATOR TEMPERATURE COEFFICIENT CilANGE Q1stanon.21 Identify the transients and accidents that are not signincantiv affected by increasing the number of pluggedSG U-tubes.
Responac:
The following transients are not significantly affected by plugging SG U-tubes. Evaluations have been performed for these events to confirm that the results of the limiting safety analyses for these events still meet appropriate NRC acceptance criteria.
Boron Dilution Loss of Flow Asymmetric SG Boron Dilution The analyses for the Modes 1 through 4 Boron Dilution events assume an RCS volume corresponding to zero plugged SG U-tubes. Plugging SG U-tubes reduces the available RCS volume. Reducing the RCS volume will exacerbate the effect of the diluting water from the charging system assumed for the event.
An evaluation was performed to verify the adverse effect of reducing the RCS volume is not significant.
The Mode 1 Boron Dilution event involves an approach to the DNB and LHR specified acceptable fuel design limits (SAFDLs). The current Mode 1 Boron Dilution safety analysis demonstrates that the maximum rate of reactivity addition is several orders of magnitude slower than a CEA withdrawal event.
Therefore, the reduced RCS volume due to plugging SG U-tubes would not have a significant effect on the maximum reactivity addition rate in the Mode I scenario. The CEA withdrawal event remains limiting. Analysis of the CEA withdrawal event demonstrates that the DNB and LHR SAFDLs are not exceeded.
l The Modes 2 through 4 Boron Dilution events involve an erosion of the available shutdown margin. The NRC acceptance criteria for Calvert Cliffs for these events is that shutdown margin will not be lost in less than 15 minutes. The current analyses for the Modes 2 through 4 Boron Dilution events demonstrate at least 45 minutes is available prior to loss of shutdown margin. The maximum number of plugged SG U-tubes allowed under internal controls represents less than 5% of the total available RCS volume.
l Therefore, the change in RCS volume associated with plugging SG U-tubes will not significantly reduce the time to loss of shutdown margin. At least 15 minutes will remain available. Therefore, plugging SG U-tubes will not significantly affect the limiting safety analysis for the Boron dilution event.
I Loss of Flow l
The limiting Loss of Flow event for Calvert Cliffs is a concurrent loss of power to all four RCPs. Total i
RCS flow rapidly decreases until the RCS low flow trip causes a reactor trip. The rate of RCS flow decrease is determined by the momentum of the RCP rotating element and the flow resistance of the RCS. The flow "coastdown" rate assumed in the limiting analysis was established based on test data taken in 1981 for an essentially zero plugged tube condition. The NRC acceptance criteria for this event is that the DNB and LHR fuel SAFDLs are not exceeded.
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ATTACHMENT (1)
BALTIMORE GAS AND ELECTRIC COMPANY'S RESPONSE TO NRC's REQUEST FOR ADDITIONAL INFORMATION CONCERNING LICENSE AMENDMENT REQUEST: MODERATOR TEMPERATURE COEFFICIENT CilANGE Plugging SG U-tubes will increase the RCS flow resistance seen by the RCPs. As previously discussed, Calvert Cliffs will ensure that steady state RCS flow remains above 370,000 gpm, the minimum RCS flow rate assumed in the limiting Loss of Flow safety analysis. The increased RCS flow resistance will increase the rate of RCS f'ow coastdown slightly during the Loss of Flow event. An evaluation was performed to determine the impact of a slightly faster RCS flow coastdown on the results of the limiting analysis for this event.
The effect of plugging SG U-tubes on the four-pump-flow coastdown was quantified using ABB/CE's COAST code. It was determined that the more rapid flow coastdown results in 0.4% less flow during the time of minimum departure from nuclear boiling ratio (DNBR) for this event. Review of the available margin established by the Reactor Protective System (RPS) setpoints, in conjunction with the Technical Specification LCOs, confirmed that the fuel SAFDLs are not exceeded given the change in RCS flow coastdown. Therefore, plugging SG U-tubes while maintaining constant RCS flow does not significantly affect the results of the limiting safety analysis for Loss of Flow.
Asymmetric SG The limiting Asymmetric SG event at Calvert Cliffs is a Loss of Load to one SG caused by closure of one Main Steam Isolation Valve. During this event the non-uniform core inlet temperature distribution I
in conjunction with the moderator temperature reactivity feedback causes an increase in local core power peaking and an approach to the fuel SAFDLs. The reactor trip during this event is initiated as a result of increasing SG differential pressure.
An evaluation was performed to address the effect of a larger SG tube plugging asymmetry on the limiting safety analysis for this event. As more SG tubes are plugged, the potential exists to have a larger differential in the number of plugged tubes between the SGs. A maximum difference between the number of plugged tubes in each SG was established and will be controlled by our reload design process.
This maximum difference was evaluated for this event. Additionally, a calculation was performed to confirm that the RCS flow splits caused by the tube plugging asymmetry would not affect the core inlet flow distribution. The effect of the tube plugging asymmetry on SG differential pressure was also considered. The concern was that a larger SG tube plugging asymmetry may cause a pressure difference between the SGs prior to event initiation that would affect the results of the limiting asymmetric event l
safety analysis.
The thermal margin degradation occurs during this event as a result of a temperature diff:rence across the core caused by the asymmetric SG loading. It was determined that the inlet temperature difference across the core at the time of the reactor trip on high SG differential pressure would not be significantly affected by the initial pressure difference between SGs. Therefore, the evaluation co icluded that the established maximum SG tube plugging asymmetry does not significantly affect the results of the limiting safety analysis for the Asymmetric SG event.
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ATTACHMENT (1) c HALTIMORE GAS AND ELECTRIC COMPANY'S RESPONSE TO NRC'S REQUEST FOR ADDITIONAL INFORMATION CONCERNING LICENSE AMENDMENT REQUEST: MODERATOR TEMPERATURE COEFFICIENT CIIANGE Question 3:
Identify the transients and accidents that are sienificantiv affected by increasing the number ofplugged SG U-tubes. Provide detailed results ofyour reanalysesfor these events with the proposed Technical Specipcation change for the value of the MTC (moderator coefficient temperature). Include the following:
1.
Confirm that the methodologies usedin the reanalyses are NRC approved.
1 2.
Confirm that all assumptions used in the reanalyses are consistent with the analyses of record.
i 3.
Provide transient data j
4.
Confirm that the acceptance criteria are metfor each event analy:ed.
Identify the transients and accidents that are significantly affected by increasing the number ofplugged SG U-tubes.
Response
The limiting safety analyses for the following transients and accidents are significantly affected by increasing the number of plugged SG U-tubes:
CEA Withdrawal (Hot Full Power, Hot Zero Power, Over-Pressure)
Loss of Load Loss of Feedwater(Over-Pressure)
Feed Line Break LOCA (Small & Large Breaks)
The first four events above are events during which the RCS heats up as a result of increasing reactor power or a degradation of the secondary side heat sink. During these four events, plugging SG U-tubes causes further degradation of heat transfer to the secondary system, which exacerbates the RCS heatup.
Therefore, the limiting safety analyses for these events are significantly affected by plugging SG U-tubes. Reanalyses of these events have been performed to quantify the effect of plugging SG U-tubes on these events. Credit for a less positive (more restrictive) MTC than required by Technical Specifications was taken in order to mitigate the effect of the exacerbated RCS heatup.
For Small Break LOCA the number of plugged SG U-tubes is significant for two reasons: RCS inventory is reduced, and heat transfer to the secondary system is reduced. The reduced RCS inventory results in a more rapid core uncovery during the course of the accident. The reduced heat transfer to the secondary system results in elevated RCS pressure, which reduces flow from the high pressure safety
[-
pumps and prolongs the core uncovery period. Therefore, the limiting Small Break LOCA analysis is significantly affected, and the event has been reanalyzed to quantify the effect of plugging additional SG U-tubes.
For Large Break LOCA, the number of plugged SG U-tubes is significant primarily since it affects the available flow area through the SGs. This flow area is important for cold leg breaks (the limiting break location at Calvert Cliffs) since it affects the pressure drop caused by steam flow from the core, through 5
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l ATTACHMENT (1) l BALTIMORE GAS AND ELECTRIC COMPANY'S l
RESPONSE TO NRC'S REQUEST FOR ADDITIONAL INFORMATION CONCERNING LICENSE AMENDMENT REQUEST: MODERATOR TEMPERATURE COEFFICIENT CIIANGE the SG, and out the break. A smaller flow area results in a larger pressure drop, which depresses the two-phase water level in the core. This effect will elevate the peak clad temperature for the event. Therefore, l
the limiting safety analysis for Large Break LOCA is significantly affected, and the event has been reanalyzed to quantify the effect of plugging additional SG U-tubes.
1.
Confirm that the methodologies usedfor the reanalyses are consistent with the analyses of record.
Respanse:
The methodologies used in the reanalyses are consister,t with those previously used in the most recent analyses reviewed by NRC.
For the non-LOCA transients, CESEC was used as the general plant transient code to calculate the time dependent core power, flow, coolant temperatures, and pressure. CETOP, based on a simplified TORC model, is the code used to calculate the DNBR during the course of the transient using input from CESEC.
l For the Small Break LOCA analysis, CEFLASH-4AS is used to calculate the RCS blowdown.
PARCH is used to calculate the hot rod heatup.
l i
For the Large Break LOCA analysis, CEFLASH-4A is used to calculate the RCS blowdown.
COMPERC-Il is used to calculate the RCS refill and reflood, PARCH is used to calculate the steam cooling heat transfer coefficients, and STRIKIN-il is used to calculate the hot rod heatup.
2.
Confirm that all assumptions used in the reanalyses are consistent with the analyses ofrecord.
Response
The assumptions used in the reanalyses are consistent with the assumptions previously used in the most recent analyses reviewed by the NRC. The significant assumptions used for each reanalysis are provided in the " Initial Conditions and input Parameters" tables of Appendix A.
The only "significant change from previous assumptions is the more restrictive MTC 4
(+0.15 x 10 Ap/ F in place of + 0.3 x 10 Ap/ F) supported by the subject change to the Technical Specifications. In addition, a reduced RCS flow rate (358,900 gpm in place of 370,000 gpm) is conservatively used in anticipation of the need to reduce the required minimum RCS flow rate in the future. Other assumptions are changed in an insignificant manner to match the physical effects of tube plugging, to update core and physics parameters, to incorporate RPS and Engineered Safety Features Actuation Signal analytical setpoints based on updated uncertainty calculations, and to update the main steam safety valve model and pressurizer safety l
valve model based on vendor test data.
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ATTACHMENT (1) l BALTIMORE GAS AND ELECTRIC COMPANY'S RESPONSE TO NRC'S REQUEST FOR ADDITIONAL INFORMATION CONCERNING
. LICENSE AMENDMENT REQUEST: MODERATOR TEMPERATURE COEFFICIENT CIIANGE 3.
Provide the transient data associated with the reanalyses.
Response
l Transient data are provided in Appendix A in the form of plots of time-dependent values for the significant parameters for each event.
4.
Confirm that the NRC acceptance criteria are metfor each event reanaly:ed.
Response
I CEA Withdrawal The reanalysis of the CEA Withdrawal event demonstrates that the action of the RPS prevents exceeding the fuel SAFDLs and the RCS pressure safety limit of 2750 psia.
Loss of Load. Loss of Feedwater (over oressure) and Feed Line Break The reanalyses of these events demonstrates that the action of the RPS, the pressurizer safety valves, and the main steam safety valves is sufficient to ensure that the RCS pressure safety limit of 2750 psia, and the secondary system pressure limit of 1100 psia, are not exceeded.
Small Break LOCA and Large Break LOCA The reanalyses of these events demonstrates that the acceptance criteria of 10 CFR 50.46(b) are met.
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i APPENDIX (A) i i
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i TRANSIENT DATA i.
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Calvert Cliffs Nuclear Power Plant November 20,1996
t TABLE 14.2-1 INITIAL CONDITIONS AND INPUT PARAMETERS FOR THE CEA WITHDRAWAL EVENT PARAMETER UNITS UNIT 1 UNIT 2 i
Initial Core Power Level HZP MWt 0
0 HFP 2700'3/2754(b) 2700*l/2754 )
I f
fb Core Inlet Coolant Temperature HZP F
532 532 548 *)/550(b)
HFP 548 'I/550(b)
I I
RCS Pressure psia 2200(*)
2200*)
C 2165(b) 2165(b)
MTC 10-* Ap/ F
+0.7 to -3.0
+0.7 to -3.0 l
Doppler Coefficient Multiplier
.85
.85 CEA Worth at Trip - (HFP) 10-2 Ap
-5.0
-5.0' l
CEA Worth at Trip - (HZP) 10-2 Ap
-3.5
-3.5 l
Reactivity Insertion Rate X10-* Ap/sec 0 to 1.6 0 to 1.6 Holding Coil Delay Time sec 0.5 0.5 CEA Time to 90% Insertion (Including Holding Coil Delay) sec 3.1 3.1 CALVERT CLIFFS UFSAR 14.2-9 Rev. xx
..c TABLE 14.2-1 (Continued).
INITIAL CONDITIONS AND INPUT PARAMETERS FOR THE CEA WITHDRAWAL EVENT PARAMETER UNITS UNIT 1 UNIT 2 I
Rod Group Withdrawal Speed in/ min 30.0 30.0 CEA Differential Worth X10 Ap/ inch 0.0 to 3.2 0.0 to 3.2
(*)
For DNBR calculations, effects of uncertainties on these parameters were combined statistically.
(b)
For the peak RCS pressure case, the effects of uncertainties on these parameters were combined deterministically.
CALVERT CLIFFS UFSAR 14.2-10 Rev. xx
4 TABLE 14.2-2 SEQUENCE OF EVENTS FOR ZERO POWER CEA WITHDRAWAL EVENT TIME ANALYSIS 1
(sec)
EVENT SETPOINT OR VALUE 0.0 CEA Withdrawal Causes Uncontrolled 1
Reactivity Insertion 26.9 VHPT Signal Generated 40% of 2700 MWt-l l
27.3 Reactor Trip Breakers Open 27.7 Core Power Reaches Maximum 131% of 2700 MWt l
27.7 Maximum Peak LHR and Maximum 43.6kW/ft Fuel Centerline Temperature 3,260'F l
27.8 CEAs Begin to Drop into Core l
28.5 Core Heat Flux Reaches Maximum 58.7% of 2700 MWt l
28.5 Minimum DNBR Occurs 1.22 l
29.9 RCS Pressure Reaches Maximum 2384 psia l
4 CALVERT CLIFFS UFSAR 14.2-11 Rev. xx i
4 TABLE 14.2-3 SEQUENCE OF EVENTS FOR FULL POWER CEA WITHDRAWAL EVENT TIME ANALYSIS (sec)
EVENT SETPOINT OR VALUE 0.0 CEA Withdrawal Causes Uncontrolled Reactivity Insertion 3.95 HPT Signal Generated 110.2% of 2700 MWt l
4.35 Reactor Trip Breakers Open l
4.55 Core Power Reaches Maximum 118.3% of 2700 MWt l
4.65 Minimum DNBR Occurs
> 1.21 l
4.8E CEAs Begin to Drop Into Core l
5.15 Core Heat Flux Reaches Maximum 110.2% of 2700 MWt l
6.1 RCS Pressure Reaches Maximum 2263 psia l
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CALVERT CLIFFS UFSAR 14.2-12 Rev. xx
TABLE 14.2-4 SEQUENCE OF EVENTS FOR FULL POWER CEA WITHDRAWAL EVENT WITH RESPECT TO PEAK PRESSURE J
TIME ANALYSIS (sec)'
EVENT SETPOINT OR VALUE 0.0 CEA Withdrawal Causes Uncontrolled Reactivity Insertion 56.3 High Pressurizer Pressure Trip 2420 psia l
Signal Generated 56.9 VHPT Signal Generated 112.2% of 2700 MWt l
l 57.2 Trip Breakers Open l
57.7 CEAs Begin to Drop Into Core 57.8 Core Power Reaches Maximum 112.4% of 2700 MWt l
57.9 Core Heat Flux Reaches Maximum 111.5% of 2700 MWt l
59.4 Pressurizer Pressure Reaches 2496 psia (*) -
l Maximum 60.1 Steam Generator Safety Valves 1010 psia Begin to Open l')
Maximum RCS pressure includes elevation head.
l CALVERT CLIFFS UFSAR 14.2-13 Rev. xx
~.
FIGURE 14.2-1 CEA Withdrawal Event Core Power Versus Time 150 ZERO POWER
[
125 2
100 8sW LLO c.
75 Lu3O c.
LU T
O 50 25 0
0 20 40 60 80 100 TIME, SECONDS
i i
l'.
l l
FIGURE 14.2-2 l
CEA Withdrawal Event
\\
l Core Heat Flux Versus Time l
l i
l i
100 l
ZERO POWER e
80 l
l 2
o 60 f
t a
'N (N
LL O
N 40 D
J LL r
H b
i z
20 O
O i
2 0
-20 0
20 40 60 80 100 TIME, SECONDS l
FIGURE 14.2-3 CEA Withdrawal Event RCS Temperatures Versus Time 580 ZERO POWER e
u.
570 vi
^
w CD l--<
T out cw 560 0
c.2w F--
2m Hm 550 O
l T avg 5O 8
540 t
m in O
F-O<
1 W"
530 l
1 520 0
20 40 60 80 100 J
l TIME, SECONDS l
l
4 l
l~
FIGURE 14.2-4 1.
CEA Withdrawal Event RCS Pressure Versus Time i
l l
ZERO POWER 2600 l
1 2500 m-O.
uI
~
CD 2400 w
cc Q.
2
~
w-
. f--m>
2300 w
' F-Z j
~
l O
i OO 2200 s
cc O
l
>-O uic 2100 l
2MO l
0 20 40 60 80 100 l
TIME, SECONDS
4
'I i
FIGURE 14.2-5 CEA Withdrawal Event i
Core Power Versus Time i
i l
i 1
h i
i 120 HOT FULL POWER 100 h
2 80 o
O b.
N LL O
60 cc W3 oc.
W o
40
~
O 20 0
0 20 40 60 80 100 TIME, SECONDS
t i
l FIGURE 14.2-6 i
CEA Withdrawal Event l
Core Heat Flux Versus Time l
l 120 HOT FULL POWER l
?
100 1
80 oo r
t N
CNI l
t,L O
>f 60 D
J LL LU l
I y.
40 l
O 1
O l
20 0
0 20 40 60 80 100 i
TIME, SECONDS
~.
FIGURE 14.2-7 CEA Withdrawal Event RCS Temperaturer, Versus Time I
j 620 l
HOT FULL POWER u.
600 T out vi w
i C
Cw 580 c_2wg 2w F-m 560
- T avg
>w F-z in 5o 8
540 2
C O
o I
~
6 i
520 l
I
~
500 0
20 40 60 80 100 TIME, SECONDS
7
-.... -.... ~
FIGURE 14.2-8 CEA Withdrawal Event RCS Pressure Versus Time 2300 HOT FULL POWER 2200 m
O.
g 2100 m
Dm m
wC Q.
3 2000 t
w
&m m
HZ 1900 OOO CO F-0 1800
-s e
1700 l
I 1600 0
20 40 60 80 100 l
TIME, SECONDS l
i u
p :.
FIGURE 14.2-10 l
CEA Withdrawal Event 1
i Core Power Versus Time i
l l
120 l
.12 E-4 Delta Rho /lNCH 100 s
80 oO l
N l
m LL O
l 60
~
ccw5o
~
a.
m o
40 O
l 20 0
20 40 60 80 100 TIME, SECON,DS
4.
FIGURE 14.2-11 CEA Withdrawal Event Core Heat Flux Versus Time 1
1 120
.12 E-4 Delta Rho /lNCH 100 4
2.
o.
80 Ra LL O
[
60 es
<C W
'E E
40
~
O O
20 0
0 20 40 60 80 100 TIME, SECONDS
i
\\
FIGURE 14.2-12 l
CEA Withdrawal Event RCS Pressure Versus Time r
4 2600-1
.12 E-4 Delta Rho / INCH 2500 i
i m
c.
i
'g
.2400 c
3m 4
W.
w C
i 0.
y 2300 w
1--w pw 1
F-Z-
2200 j
O O
\\
O
)
e i
OO 2100 bC 1
2000-i i
1 1900
.i 0
20 40 60 80 100 TIME, SECONDS
y..
FIGURE 14.2-13 CEA Withdrawal Event RCS Temperatures Versus Time l
l l
l f
l 640 i
j
.12 E-4 Delta Rho / INCH l
u.
620 vi
~
m h
1 c
T out cw-600 c.
2 I
w-F-
2 W
F-.
T avg CD 580 en F--
I Z
5O O
o 560 s
2.
l c
O T in F--
O l
C 540 520 j
O 20 40 60-80 100 TIME, SECONDS l
FIGURE 14.2-14 CEA Withdrawal Event Steam Generator Pressure Versus Time 1100
.12 E-4 Delta Rho / INCH 1050 2G m
n.
1000 aww w
C Q.
T O
950 ke i
w zw i
0
-l 2
900 I-m 850-800 O
20 40 60 80 100 TIME, SECONDS
v TABLE 14.5-1 INITIAL CONDITIONS AND INPUT PARAMETERS FOR THE LOSS OF LOAD EVENT TO CALCULATE MAXIMUM RCS PRESSURE PRESSURE UNITS UNIT 1 UNIT 2 Initial Core Power Level MWt 2754(b) 2754(b)
Initial Core Inlet Coolant
- F 550 550 Temperature 6
Core Mass Flow Rate X 10 lbm/hr 129.8 129.8 Initial RCS Pressure psia
-2165(*)
2165(*)
3 Initial Pressurizer ft 975 975 Liquid Level at Full Power Initial SG Pressure psia 831 831 MTC X 10" Ap/ F
+0.15
+0.15 Doppler Coefficient 0.85 0.85 Multiplier Number of Plugged SG Tubes 1500 1500 Axial Shape Index
+0.6
+0.6 CEA Worth at Trip
% Ap
-5.0
-5.0 Time to 90% Insertion sec 3.1 3.1 of SCRAM Rods i
RRS Operating Mode Manual Manual SDBS Operating Mode Inoperative Inoperative Minimum MSSV Opening psia 1010 1010 Pressurizer Pressure Operating Mode Manual Manual Control System Pressurizer Level Operating Mode Manual Manual Control System j
i
(*)
Corresponds to Technical Specification minimum indicated pressure of 2200 psia. The value includes an uncertainty of 35 psia.
(b)
Value does not include 17 MWt of pump heat added to core power in CESEC. l l
CALVERT CLIFFS UFSAR 14.5-6 Rev. xx
4 TABLE 14.5-2 SEQUENCE OF EVENTS FOR LOSS OF LOAD EVENT TO MAXIMIZE CALCULATED RCS PEAK PRESSURE TIME (sec)
EVENT SETPOINT OR VALUE 0.0 Loss of Secondary Load 5.85 Steam Generator Safety Valves Begin 1010 psia l
to Open 7.44 High Pressurizer Pressure Trip 2420 psia l
Signal Generated l
7.94 CEAs Begin to Drop Into the Core 8.39 Pressurizer Safety Valves Begin to Open 2550 psia l
I 10.06 Maximum RCS Pressure 2658 psia *)
l N
11.24 Maximum SG Pressure 1092 psia l
14.'5 PSVs are Fully Closed 2448 psia l
(*)
RCS pressure includes elevation head.
N Steam Generator pressure incivdes downcomer liquid head.
CALVERT CLIFFS UFSAR 14.5-7 Rev. xx
?
TABLE 14.5-4 INITIAL CONDITIONS AND INPUT PARAMETERS FOR THE LOSS OF LOAD EVENT TO CALCULATE MAXIMUM SECONDARY PRESSURE PRESSURE UNITS UNIT 1 UNIT 2 Initial Core Power Level MWt 2754(b) 2754 )
0 Initial Core Inlet Coolant "F
550 550 1
Temperature 6
Core Mass Flow ate X 10 lbm/hr 133.9 133.9 Initial RCS Pressure psia 2165*)
2165(*)
I 3
Initial Pressurizer ft 800 600 Liquid Level at Full Power Initial SG Pressure psia 865 865 MTC X 10" Ap/ F
+0.15
+0.15 Doppler Coefficient 0.85 0.85 Multiplier CEA Worth at Trip
% Ap
-5.0
-5.0 Number of Plugged SG Tubes 0
0 Axial Shape Index
+0.6
+0.6 Time to 90% Insertion sec 3.1 3.1 of SCRAM Rods RRS Operating Mode Manual Manual i
SDBS Operating Mode Inoperative Inoperative Minimum MSSV Opening psia 1010 1010 Pressurizer Pressure Operrt ing Mode Auto Auto Control System Pressurizer Level Operating Mode Auto Auto Control System W
Corresponds to Technical Specification minimum indicated pressure of 2200 psia. The value includes an uncertainty of 35 psia.
D)
Value does not include 17 MWt of pump heat added to core power in CESEC.
A CALVERT CLIFFS UFSAR 14.5-9 Rev. xx
.s FIGURE 14.5-1 LOSS OF. LOAD EVENT CORE POWER VERSUS TIME.
. I i
+
l
.c.
.s.
120
.t 100 -
4 s
f 4
60 -
6.
5,
. o o
. e-n E
i O
Y 60 -
' C I
es 3
O t
L'i=
0 0-4 20 t
1 l
1
)'
0 0
20
-Q 53 80 1CO 4
4 TIME. S ECONCS i
e s
y e r e
+-q "Tr-t-v'w+
- - - +
--.g e-=-- y-9 -r T -- *
-ay y-h,v
r.,
$t i
i FIGURE 14.5-2 4
LOSS OF LOAD EVENT-CORE HEAT FLUX VERSUS TIME J
t i
1 4
120 i
u
~
3,.,.
I I
J l
a 1
j 100 -
=
h 2
2 I:
80 --
d
~
ed
\\
0 i
Y N
~3 s
A 60 -
p j-d i
x W
's O
k a
4 y><
40 -
W e
o.
]
U d
I 20 -
i J
0 0
20 40 60 80 100 i
1 TIME, SECONOS l
.---m.
_.. _. ~ _ _ _ _... _....
. =...
l.
l i
t l
l FIGURE 14.5-3 i
LOSS OF LOAD EVENT RCS PRESSURE VERSUS TIME l
2800 l
~
l I
2600 e
i f
CL.
uIc: 2400 o
M m
mEc.
E w
1' H
$ 2200 -
o l
H2 5
l 8
i O
l CfO 2000 w
o l
i W
1800 -
l j.
1600 0-20 40 60 80 1,00 TIME, SECONDS s
I l
l l
l.
z r
t 1
FIGURE 14.5-4 t
LOSS OF LOAD EVENT i
1 I
RCS TEMPERATURES VERSUS TIME i
v.
,y 650 I
t
~625 -
vi
\\
W' s'
i e
+
f_y_
&~ InJ t he l
l E
N l
W.
H 2
i w
$ 575 -
V) i H
=
-2
.j'
.i 8u
{"
g z
'O e
p r
o W
l.
e i
525 -
\\.\\
3 i
e SCO 0'
20 40 60 80 t00 TIME, SECONOS i
.l j
~
-~
l t
i FIGURE 14.5-5 r
i t
LOSS OF LOAD EVENT l
l STEAM GENERATOR PRESSURE VERSUS TIME 4
1 i:
1200 i
t i
i t
I~
1100 -
t.
1 i
l 2 1000 -
ur-e D
an M
We
- A. :
5. sco -
5 L'J
^6 0-
- 2 d
8Co --
-m l
7Co -
I 6co o
20 4
60 80 100 l.
TIME. SECONDS L
- Does not include downcomer liquid head.
4
,.m-
4 TABLE 14.6-1 INITIAL CONDITIONS AND INPUT PARAMETERS FOR TNE LOFW EVENT TO MAXIMIZE CALCULATED RCS PEAK PRESSURE WITN LOAC l
PARAMETER UNITS UNIT 1 UNIT 2 Initial Core Power Level MWt 2754 2754 Initial Core Coolant Inlet F
550 550 Temperature Initial RCS Vessel Flow Rate gpm 358,900 358,900 l
Initial RCS Pressure psia 2165 2165 l
Initial SG Pressure psia 865 865 l
3 Initial Pressurizer Liquid ft 975 975 Volume MTC X 10 Ap/ F
+0.15
+0.15 l Doppler Coefficient Multiplier 0.85 0.85 High Pressurizer Pressure psia 2420 2420 l
Analysis Trip Setpoint RRS Operating Mode Manual ')
Manual *)
I I
SDBS Operating Mode Manual (')
Manual (')
l J
CALVERT CLIFFS UFSAR 14.6-9 Rev. xx
t TABLE 14.6-1 (Continued)
INITIAL CONDITIONS AND INPUT PARAMETERS FOR THE LOFW EVENT TO MAXIMIZE CALCULATED RCS PEAK PRESSURE WITH LOAC l
PARAMETER UNITS UNIT 1 UNIT 2 PPCS Operating Mode Manual *)
Manual ')
I I
1 i
PLCS Operating Mode Manual ')
Manual ')
I f
~
(*)
These modes of control system operation maximize the peak RCS pressure.
4 t
?
?
i i
l i
\\
CALVERT CLIFFS UFSAR 14.6-10 Rev. xx
l.
l I
TABLE 14.6-3 SEQUENCE OF EVENTS FOR LOFW EVENT TO MAXIM 1ZE CALCULATED RCS PRESSURE WITH LOAC TIME
.[ttel EVENT SETPOINT OR VALUE 0.0 Loss of MFW 21.2 High Pressurizer Pressure Trip 2420 psia l
Setpoint Reeched 22.1 Trip Breakers Open l
22.4 Turbine Stop Valves Close 22.6 CEAs Begin to Drop into Core, Loss of AC Power 24.1 SG Safety Valves Begin to Open 1010 psia l
24.5 Primary Safety Valves Begin to Open 2550 psia l
26.3 Maximum RCS Pressure 2629 psia (*)
l 29.8 Maximum SG Pressure 1099 psia (b) l 31.2 Primary Safety Valves Close 2448 psia l
)
Pressure includes elevation head.
(b)
Steam Generator pressure includes downcomer liquid head.
l l
CALVERT CLIFFS UFSAR 14.6-13 Rev. xx i
i i
~
FIGURE 14.6-1 LOFWFLOWEVENT/MAXIMIZERCSPEAKPRESSURE WITH LOAC FOLLOWING TRIP CORE POWER VERSUS TIME t
120 100
\\
b2 80 8
Si u.O 60 ccw 3:
O
~
w o
40 2
0 1
1 20 0
0 20 40 60 80 100 120 i
i TIME, SECONDS
FIGURE 14.6-2 LOFWFLOWEVENT/MAXIMIZERCSPEAKPRESSURE WITH LOAC FOLLOWING TRIP l
CORE HEAT FLUX VERSUS TIME i
u 120 F
100
}
b2 o
80 Rw u.
O y
60 d
i
!E tu I
r 40 OO 20 0
0 20 40 60 80 100 120 N
l TIME, SECONDS
.m.
.m..-
,__.,_.m
~
~
s FIGURE 14.6-3 LOFWFLOWEVENT/MAXIMIZERCSPEAKPRESSURE WITH LOAC FOLLOWING TRIP RCS TEMPERATURES VERSUS TIME 700 u.
650 vi
~
C E
y T ou g
600 Ta H
2 550 m
s Z
5o 8
500 2
CO O
~
C 450 l
400 0
20 40 60 80 100 120 TIME, SECONDS
l1 FIGURE 14.6-4 LOFW FLOW EVENT / MAXIMIZE RCS PEAK PRESSURE WITH LOAC FOLLOWING TRIP RCS PRESSURE VERSUS TIME 2700 l
2550 w
Cl.
LII
~
CD 2400 w
C Q.
2
~
w Hw 2250 W
H l
Z 5
l
\\
OO0 2100 4
C O
HO bC 1950 3
l 1800 0
20 40 60 80 100 120 TIME, SECONDS i
i i
FIGURE 14.6-5 LOFWFLOWEVENT/MAXIMIZERCSPEAKPRESSURE WITH LOAC FOLLOWING TRIP STEAM GENERATOR ~ PRESSURES VERSUS TIME 1200 se 1100 w
CL 1000 2m
'Is 4g C
J G-o 900 LE CW ZW G
2 800 4
4W in 700 600 O
20 40 60 80 100 120 TIME, SECONDS 1
i j
]
TABLE 14.17-1 CALVERT CLIFFS UNIT 1 COMPARISON OF SIGNIFICANT SYSTEM PARAMETERS VALUES PARAMETERS CYCLE 2 CYCLE 13 l
Reactor Power Level 2754 2754 (102% of Rated), MWth Average LHR 6.5474 6.36 l
(102%ofRated),kw/ft MTC at Initial
+0.3
+0.3 Density, x 10" Ap/ F System Flow Rate (Total),
139.08 134.68 l
6 x 10 lbs/hr 6
Core Flow Rate, x 10 lbs/hr 134.21 129.69 l
Initial System Pressure, psia 2250 2250 Core Inlet Temperature, F 550 550 Core Outlet Temperature, F 600.6 602 l
Active Core Height, ft 11.39 11.39 Fuel Rod Outside Diameter, in 0.44 0.44 Number of Cold Legs 4
4 Number of Hot Legs 2
2 Cold Leg Diameter, in 30 30 Hot Leg Diameter, in 42 42 CALVERT CLIFFS UFSAR 14.17-26 Rev. xx
l TABLE 14.17-1 (Continued)
CALVERT CLIFFS UNIT 1 COMPARISON OF SIGNIFICANT SYSTEM PARAMETERS VALUES PARAMETERS CYCLE 2 CYCLE 13 l
SIT Pressure, psia 215 195 SI Response Time, sec-30 40 Number of Tubes Plugged per SG Large Breaks 0
s 1500 Small Breaks 0
s 2130 3
SITGas/WaterVolume,ft 855/1145 910/1090
.PLHGR,kw/ft 14.2')
14.5 I
16.5 Gap Conductance at PLHGR, 683.4(*)
1851 l
Btu /hr-ft - F(b) 2000.0 2
Fuel Centerline Temperature 3894.8(*)
3492 at PLHGR, *F(")
3788.3 Fuel Average Temperature 2609.3(*)
2160 at PLHGR, "F(b) 2303.6 Hot Rod Gas Pressure, psia (b) 1198.9(*)
1208 I
I CALVERT CLIFFS UFSAR 14.17-27 Rev. xx
j TABLE 14.17-1 (Continued)
CALVERT CLIFFS UNIT 1 COMPARISON OF SIGNIFICANT SYSTEM PARAMETERS val.UES PARAMETERS CYCLE 2 CYCLE 13 l
i Hot Average Rod Burnup 3402')
1000 f
(Min Hgap), MWD /T(b) 680 l
(*)
For low density fuel, when gap conductance is minimum.
(b)
Fuel rod values given are those which yield the limiting ECCS performance results.
CALVERT CLIFFS UFSAR 14.17-28 Rev. xx
i l
J TABLE 14.17-2 CALVERT CLIFFS UNIT 2 i
COMPARIS0N OF SIGNIFICANT SYSTEM PARAMETERS l
VALUES PARAMETERS CYCLE 2 CYCLE 11(*)
Reactor Power Level 2754 2754 (102% of Nominal), MWth Average LHR 6.5205 6.36 l
(102% of Nominal), kw/ft MTC at Initial
+0.3
+0.3 i
Density, x 10" Ap/*F System Flow Rate (Total),
139.08 134.68 l
4 6
x 10 lbs/hr 6
Core Flow Rate, x 10 lbs/hr 134.21 129.69 l
Initial System Pressure, psia 2250 2250 Core Inlet Temperature, F 550 550 Core Outlet Temperature, F 600.6 602 l
Active Core Height, ft 11.39 11.39 Fuel Rod Outside Diameter, in 0.44 0.44 Number of Cold Legs 4
4 Number of Hot Legs 2
2 Cold Leg Diameter, in 30 30 Hot Leg Diameter, in 42 42 CALVERT CLIFFS UFSAR 14.17-29 Rev. xx
1 TABLE 14.17-2 (Continued)
CALVERT CLIFFS UNIT 2 COMPARIS0N OF SIGNIFICANT SYSTEM PARAMETERS VALUES PARAMETERS CYCLE 2 CYCLE 11(*)
SIT Pressure, psia 215 195 SI Response Time, sec 30 40 Number of Tubes Plugged per Generator.
Large Break 0
s 1500 Small Break 0
s 2130 3
SITGas/WaterVolume,ft 855/1145 910/1090 PLHGR,kw/ft 15.5 14.5 Gap Conductance at PLHGR, 2000.0 1851 l
Btu /hr-ft*- F(b)
Fuel Centerline Temperature 3528.4 3492 at PLHGR, *F(b)
Fuel Average Temperature 2126.3 2160 at PLHGR, 'F(b)
Hot Rod Gas Pressure, psia (b) 1636.9 1208 Hot Average Rod Burnup 27506 1000 (Min h p), MWD /T(b) g
(*)
The Unit 1 Cycle 13 data conservatively apply for Unit 2 Cycle 11.
l (b)
Fuel rod values given are those which yield the limiting ECCS performance results.
CALVERT CLIFFS UFSAR 14.17-30 Rev. xx
TABLE 14.17-2 (Continued)
CALVERT CLIFFS UNIT 2 COMPARISON OF SIGNIFICANT SYSTEM PARAMETERS VALUES PARAMETERS CYCLE 2 CYCLE 11(*)
SIT Pressure, psia 215 195 SI Response Time, sec 30 40 Number of Tubes Plugged per Generator Large Break 0
s 1500 Small Break 0
s 2130 3
SIT Gas / Water Volume, ft 855/1145 910/1090 l
PLHGR,kw/ft 15.5 14.5 l
Gap Conductance at PLHGR, 2000.0 1851 l
Btu /hr-f t' 'F(b)
Fuel Centerline Temperature 3528.4 3492 F)
Ib at PLHGR, Fuel Average Temperature 2126.3 2160 at PLHGR, 'F(b)
Hot Rod Gas Pressure, psia (b) 1636.9 1208 Hot Average Rod Burnup 27506 1000 (Min h.p), MWD /T(b) g l
(*)
The Unit 1 Cycle 13 data conservatively apply for Unit 2 Cycle 11.
l (b)
Fuel rod values given are those which yield the limiting ECCS performance results.
CALVERT CLIFFS UFSAR 14.17-30 Rev. xx
i TABLE 14.17-6 i
TIME OF INTEREST AND FUEL ROD PERFORMANCE
SUMMARY
FOR 0.1 FT* BREAK FOR UNIT 1 CYCLE 13 l
l Time for HPSI pump on 68 sec Time for Low Pressure Safety Injection (LPSI) pump
(*)
and SI tanks on i
(b)
Time for SI H O level to reach bottom of fuel 2
l Hot spot PCT occurs 1473 sec l
Maximum clad surface temperature 2031 F l
Elevation of hot spot (from bottom of core) 10.82 ft l
Core-wide zirconium oxidation (')
< 0.95%
l l
l Peak local clad zirconium oxidation 6.40%
l l
l')
l Calculation terminated before LPSI pump or SI tank actuation.
(b) l Core never totally uncovers.
I')
l The average oxidation over the length of the hot rod is used as a i
conservative representation of the core-wide zirconium oxidation.
l l
i 2
1
)
CALVERT CLIFFS UFSAR 14.17-36 Rev. xx i
v
TABLE 14.17-14 CALVERT CLIFFS UNIT 1 AND 2 COMPARISON TO NRC ACCEPTANCE CRITERIA MAXIMUM ALLOWABLE CRITERION (2)
CRITERION (3)
LIMITING
- PLHGR, CRITERION (1)
MAX. CLAD MAX. HYDROGEN CYCLE BREAK kw/ft PCT. 'F OXIDATION. %
GENERATION. %
i UNIT 1 2
0.8x0ES/PD 14.2 2145 7.2 less than 0.541 13 0.6xDEG/PD 14.5 2143 7.27 less than 0.51 l
I UNIT 2 *)
11 0.6xDEG/PD 14.5 2143 7.27 less than 0.51 l
Criterion (1)
Peak Clad Temperature.
"The calculated maximum fuel element cladding temperature shall not exceed 2200 F."
Criterion (2)
Maximum Claddina 0xidation.
"The calculated total oxidation of the cladding shall nowhere exceed 17% of the total cladding thickness before oxidation."
Criterion (3)
Maximum Hydroaen Generation.
"The calculated total amount of hydrogen generated from the chemical reaction of the cladding with water or steam shall not exceed 1% of the hypothetical amount that would be generated if all of the metal in the cladding cylinders surrounding the fuel, excluding the cladding surrounding the plenum volume, were to react."
(*)
The Unit 1 Cycle 13 results conservatively apply for Unit 2 Cycle 11.
l CALVERT CLIFFS UFSAR 14.17-51 Rev. xx
TABLE 14.17-15
SUMMARY
OF ECCS PERFORMANCE RESULTS FOR THE LIMITING BREAK SIZE (0.6xDEG/PD)
(for Unit 1 Cycle 13 and Unit 2 Cycle 11) l 1
LIMITING CASE (MAXIMUMINITIALFUEL STORED ENERGY)
UNIT 1 UNIT 2' PARAMETERS CYCLE 13 CYCLE 11 l
RodAverageBurnupMWD/MTV 1000 1000 PC f, F
2143 2143 l
Time of PCT, seconds 266 266 l
Time of Clad Rupture, seconds 26.0 26.0 l
Peak Clad 0xidation, %
7.27 7.27 l
Core Wide Oxidation, %
<.51
<.51 Maximum Allowable Peak 14.5 14.5 LinearHeatGenerationRate,kw/ft The Unit 1 Cycle 13 results conservatively apply for Unit 2 Cycle 11.
l CALVERT CLIFFS UFSAR 14.17-52 Rev. xx
-. 3 l
s I
I FIGURE 14.17-260 l
0.1 SQ FT BREAK IN PUMP DISCHARGE LEG, i
NORMALIZED TOTAL CORE POWER VS TIME 1
1 1.8
)
)
1.5 l
1 1.2 M
x uj F
i 3
0 c.
0.9 H
O H
0.6 0.3 W
~
~
0.0 i
0 10 20 30 40 50 TIME IN SEC m
y..
..._r.
..r
+
-4
s FIGURE 14.17-261
~ 0.1 SQ FT. BREAK 'IN _ PUMP DISCHARGE LEG, INNER VESSEL PRESSURE VS TIME 2400 2
2000 t
1600 55 c
W i
x a
1200 en in W
C 0
4 800 [
g i
e
'400 _
x m
4 0
500 1000 1500 2000 2500 TIME IN SEC
-.y i
?
FIGURE 14.17-262 i
0.1 SQ FT BREAK IN PUMP DISCHARGE LEG, LEAK FLOW RATE VS TIME i
P i
6000 t
P
~
i 5000 i
E
~
i 4000 O
[
w E2 l
s r
co J
w
~
I s
3000 l
CC L
3:
o g
t 2000 1
}
1000
- V
)
i o :.........
l 0
500 1000 1500 2000 2500 TIME IN SEC l
l l
I
FIGURE 14.17-263 0.1 SQ FT BREAK IN PUMP DISCHARGE LEG, INNER VESSEL INLET FLOW RATF,VS TIME 35000 28000 I
21000 OwW I
1 co 4
i
- 14000 Q~
3oJ
~
~
7000 0
-7000 0
500 1000 1500 2000 2500 l
TIME IN SEC
-.- s._---
i i
FIGURE 14.17-264-0.1'SQ FT BREAK IN PUMP DISCHARGE LEG, INNER VESSEL TWO PHASE MIXTURE HEIGHT.
i 1
i P
J 42 i
1 4
r 35 4
l t
i a
28 h
u, h
W J
~
1 Top of Core l
W 21 W
t
.,,,i
'f N
f i;
~
14 ~
+
'h
'em, Bonom of Core 4
4 7.
4 N
4 m
4 pe g
0 500 1000 1500 2000 2500 TIME IN SEC 1
4
3 8
FIGURE 14.17-265 0.1 SQ FT BREAK IN PUMP DISCHARGE LEG, HEAT TRANSFER COEFFICIENT AT LOCATION OF PCT 5
10 g
4 10
.a
.G W
3 0
10 O
C
~
H
~
H I
W O<
2 10 D
(n
..e I
10
.d
.6 e
.9 0
9 9
0 4
0 9
4 9
- 4 9 e o e e 0
500 1000 1500 2000 2500 l
t TIME IN SEC I
l l
4 -
a I
FIGURE 14.17-266 0.1 SQ FT BREAK IN PUMP DISCHARGE LEG, COOLANT TEMPERATURE AT LOCATION OF PCT 1200 7
~
1000 1
I 800 a.
e 2
w H
H t
2 600 5
O O
O I
k_
400 _
e 200.
i O
O 500 1000 1500 2000 2500 TIME IN SEC 1
i i
I
FIGURE 14.17-267 4
0.1 SQ FT BREAK IN PUMP DISCHARGE LEG, 1
PEAK CLADDING TEMPERATURE VS TIME i
i 2200 se 1900 1
4 1600
[
~
1 a.
l 2
w
}-
w O
1<
1300 CC D
O O
5 o
1000.
s 4
4 4
700 W
400 O
500 1000 1500 2000 2500 TIME IN SEC
s e,-
FIGURE 14.17-268 0.6 DOUBLE-ENDED GUILLOTINE BREAK IN PUMP DISCHARGE LEG, CORE POWER VS TIME 1.2 4
us e
1.0 4
4
- w 4
4 MD e
0.8 d
4 cc tu 3
o a.
0.6 O
1-uG e
4 0.4.
in euen 4
e 4
e up 0.2 _
4 0
1 2
3 4
5 TIME IN SEC
1 FIGURE 14.17-269 1
f 1
0.6 DOUBLE-ENDED GUILLOTINE BREAK IN PUMP DISCHARGE LEG, PRESSURE IN CENTER H0T ASSEMBLY N0DE VS TIME i
h
~
}
2400 1
i i
2000 l
9 4
1600 J
4 5
i a
w g
D 1200 b
w i
C a
i i
800.
x
~
6m 4
se e
l 400.
s le se le 0
0 5
10 15 20 25 TIME IN SEC
.7
.t 7
FIGURE 14.17-270 i
0.6 DOUBLE-ENDED GUILLOTINE BREAK IN PUMP DISCHARGE LEG, REACTOR COOLANT PUMP SIDE LEAK FLOW RATE VS TIME 1
k i
1 120000 t
4 W
e 100000 i
4 80000 O
w
~
ca
.J
~
W Q 60000 _
.g o
_a u-40000 _
e l
20000.
se e
4 I N........
g
~
0 5
10 15 20 25 TIME IN SEC r
i
s FIGURE 14.17-271 0.6 DOUBLE-ENDED GUILLOTINE BREAK IN PUMP DISCHARGE LEG, REACTOR VESSEL' SIDE LEAK FLOW RATE VS TIME 120000 100000 80000 O
w Sn ^
5 w
Q 60000 cc So
{N 40000 N
y
.=
20000 N
~
0 0
5 10 15 20 25 TIME IN SEC l
' ?,
FIGURE 14.17-272 j
0.6 DOUBLE-ENDED GUILLOTINE BREAK IN PUMP DISCHARGE LEG, HOT ASSEMBLY FLOW RATE (BELOW HDT SPOT) VS TIME 1
30 i
20 l
)
10 O
W (D
0 K
V e
-10
-20
-30 0
5 10 15 20 25 TIME IN SEC
s j
i FIGURE 14.17-273 0.6 DOUBLE-ENDED GUILLOTINE BREAK IN-PUMP. DISCHARGE LEG, HOT ASSEMBLY FLOW RATE (ABOVE HOT SPOT) VS TIME-30 20 10 O
W en J
o E
J 8
e
-10
-20 _
-30 0
5 10 15 20 25 TIME IN SEC
FIGURE 14.17-274 1
0.6 DOUBLE-ENDED GUILLOTINE BREAK IN PUMP DISCHARGE LEG, HOT ASSEMBLY QUALITY (BELOW HOTTEST REGION) VS TIME 1.0 W
- )
0.8 I
i l
m 4
e 0.6 1
D o
)
c_
N
- 0.4 _
"N l
i l
i 0.2 l
l h
0.0 O
5 10 15 20 25 TIME IN SEC l
FIGURE 14.-17-275 O.6 DOUBLE-ENDED GUILLOTINE BREAK IN PUMP '1ISCHARGE LEG, 1
HOT ASSEMBLY QUALITY (AT HOTTEST REGION) VS TIME 1
I 1
i p-1.0 r
1
)
0.8 so l
0.6 r
D O
c.
W 0.4.
~
l
~
l 0.2
}-
0.0 O
5 10 15 20 25 TIME IN SEC
f d
FIGURE 14.17-276 0.6 DOUBLE-ENDED GUILLOTINE BREAK IN PUMP DISCHARGE LEG, HOT ASSEMBLY QUALITY (AB0VE H0TTEST REGION) VS TIME
/
1.0 4
0.8 4
0.e
/
~
a D
O c.
0.4 l
V
=
0.2 0.0 0
5 10 15 20 25 TIME IN SEC
-.z
,t
- g r
i FIGURE 14.17-277 k
0.6 DOUBLE-ENDED GUILLOTINE BREAK IN PUMP DISCHARGE LEG, i
CONTAINMENT: PRESSURE VS TIME I
i 0
4 1
60 50 i
?
1
~
40 l
g w
- I c.
tu c
D l
V) 30 l
1 a-L c.
p
~
zo
-o 20 _
)
10 _
4
~
o O'
80 160 240 320 400 i
TIME AFTER BREAK, SEC
FIGURE 14.17-278 0.6 DOUBLE-ENDED GUILLOTINE BREAK IN PUMP DISCHARGE LEG, MASS ADDED TO CORE DURING REFLOOD 120000 100000
- Time, ReDood Rate, see in/see 0.0 - 11.4 1.69 11.4 - 60.9 1.15 60.9-500.
0.67
/
80000 en C3 a
in cn.
s 60000 a
7 OO c'
uJ CC 40000' [
\\
/
20000 _
0 80 160 240 320 400 TIME AFTER CONTACT, SEC n
1 FIGURE 14.17-279 0.6 DOUBLE-ENDED GUILLOTINE BREAK IN PUMP DISCHARGE LEG, PEAK CLADDING TEMPERATURE VS TIME
~
j 2400 2
~
i 2100 N
M 1800 e
2 w
F.
a
-f Z
e 1500 w
CC O
E N
1200 f
l@
900.
W
~
600 0
100 200 300 400 500 TIME IN SEC 4
FIGURE 14.17-280 0.6 DOUBLE-ENDED GUILLOTINE BREAK IN PUMP DISCHARGE LEG, GAP CONDUCTANCE AT LOCATION OF PCT a
1800 1
nn
~
1500 1200 c.<a 900 _
I 600.
W"
- (
300
_q
- r 0
O 100 200 300 400 500 TIME IN SEC
-.......~.
FIGURE 14.17-281 0.6 DOUBLE-ENDED GUILLOTINE BREAK IN PUMP DISCHARGE LEG, MAXIMUM LOCAL CLADDING OXIDATION VS TIME 18 m
4 e
15 e
W 4
W p.
e e
12 T
I W
3:
6 c
9 R
H o
E 7
/
6 y
2 N
4 e
I
=
l 3
me MG 4
l 0
100 200 300 400 500 i
TIME IN SEC
7__.__...___.~__
- 4.
a i-FIGURE 14.17-282 i
l
.I l
O.6DEG/PD,TEMPERATUREOFFUELCENTERLINE,FUELAVERAGE, 3
CLADDING AND COOLANT AT LOCATION OF PCT VS TIME l
l 3000 l
i l
l 2500-i Fuel Centerline
" * ^*N '
~
\\
f 2000 y
Z
- N my Cladding W
k k 1500 v
u.
w>
)
1000
)
i l
l 500 _
i Coolant
\\.
J 0
j 0
100 200 300 400 500 TIME IN SEC
\\
i FIGURE 14.17-283 O.6 DOUBLE-ENDED GUILLOTINE BREAK IN PUMP DISCHARGE LEG, HEAT TRANSFER COEFFICIENT AT LOCATION OF PCT VS TIME i
180 150 120 mO O
U) 2<
90 cr F
H<w I
60 30.
0 100 200 300 400 500 TIME IN SEC
j FIGURE 14.17-284
)
0.6 00!!BLE-ENDED GUILLOTINE BREAK IN PUMP DISCHARGE LEG, I
HOT R0D INTERNAL GAS PRESSURE VS TIME i
1800 1500
~
)
INITIAL GRESSURE i= 1208.3 PSIA i
< 1200 s
M G.
W 3
900 _"
m m
W E
CLADOING RUPTURE AT 26.0 SEC
~
600 _
300 ~:
~
O ' ' ' ' ' ' ' ' ' 20 40 60 80 100 O
TIME. SECONDS
4 TABLE 14.26-1 INITIAL CONDITIONS AND INPUT PARAMETERS ASSUMED IN THE FEEDWATER LINE BkEAK EVENT PARAMETER UNITS UNIT 1 UNIT 2 Initial Core Power Level MWt 2754.0 2754.0 Initial Core Coolant Inlet F
550.0 550.0 Temperature Initial RCS Vessel Flow Rate gpm 358,900 358,900 l
Initial RCS Pressure psia 2165.0 2165.0 Initial SG Pressure psia 828.8 828.8 l
Initial Pressurizer Liquid Volume ft 975.0 975.0 Effective MTC x10 Ap/ F
+0.15
+0.15 l
Doppler Coefficient Multiplier 0.85 0.85 High Pressurizer Pressure psia 2470 2470 Analysis Trip Setpoint AFW Actuation
% WR Tap Span 29.1 29.1 l
SG Differential Pressure psid 10.0 10.0 Analysis Setpoint CEA Worth at Trip
% Ap
-5.3
-5.3 RRS Operating Mode Manual *)
Manual (*)
I l
CALVERT CLIFFS UFSAR 14.26-10 Rev. xx
e 4
TABLE 14.26-1 (Continued)
INITIAL CONDITIONS AND INPUT PARAMETERS ASSUMED IN THE FEEDWATER LINE BREAK EVENT PARAMETER UNITS UNIT 1 UNIT 2 SDBS Operating Mode Manual (')
Manual (')
PPCS Operating Mode Manual (*)
Manual (*)
PLCS Operating Mode Manual (*)
Manual (*)
(*)
These modes of control system operation maximize the peak RCS pressure.
l CALVERT CLIFFS UFSAR 14.26-11 Rev. xx
TABLE 14.26-3 SEQUENCE OF EVENTS FOR FEED LINE BREAK EVENT WITH LOAC FOLLOWING REACTOR TRIP ANALYSIS TIME (sec)
EVENT SETPOINT OR VALUE 0.0 Break in Main Feedwater Line 0.325 ft*
1 22.1 Heat Transfer Area Rampdown 19691 lbm l
in LHSG Begins 24.9 High Pressurizer Pressure Trip 2470 psia l
Analysis Setpoint is Reached 25.7 Level in the Ruptured SG 5000 lbm l
decreases below the assumed nozzle level; steam will be blown out of the break 25.9 First Primary Safety Valve 2550 psia l
Begins to Open 26.3 Trip Breakers Open l
26.8 CEAs Begin to Enter Core; l
LOAC on Turbine Trip; RCS Pumps Begin to Coastdown N
28.9 Peak RCS Pressure 2747 psia l
35.0 Undamaged SG Safety Valves 1010 psia l
Begin to Open 37.7 Damaged SG Safety Valves 1010 psia Begin to Open CALVERT CLIFFS UFSAR 14.26-13 Rev. xx
e TABLE 14.26-3 (Continued)
SEQUENCE OF EVENTS FOR FEED LINE BREAK EVENT WITH LOAC FOLLOWING REACTOR TRIP ANALYSIS TIME (sec)
EVENT SETPOINT OR VALUE 38.7 Maximum SG Pressure, 1040.6 psia ')
l f
Undamaged 39.0 Maximum SG Pressure, 1011.5 psia l
Damaged 41.7 Damaged SG Safety Valves 1010 psia are Closed 42.1 Primary Safety Valves are 2448 psia Closed 48.5 Undamaged SG Safety Valves 1010 psia l
are Closed 63.0 AFW Analysis Setpoint is 29.1% of WR Tap Span l
reached in Undamaged SG 163.0 Main Steam Isolation Signal 600 psia l
163.9 MSIVs Begin to Close l
169.9 MSIVs are Fully Closed l
247.6 AFW Flow Established to 180 gpm l
Undamaged SG 375.5 Undamaged SG Safety Valves 1010 psia l
l Begin to Open 453.5 First Primary Safety Valve 2550 psia l
j Begins to Open CALVERT CLIFFS UFSAR 14.26-14 Rev. xx
TABLE 14.26-3 (Continued)
SEQUENCE OF EVENTS FOR FEED LINE BREAK EVENT WTvN LOAC FOLLOWING REACTOR TRIP ANALYSIS TIME (sec)
EVENT SETP0 INT OR VALUE 626.3 Operator Increases AFW Flow l
l 630.4 Primary Safety Valves are Fully 2448 psia l
Closed i
l')
SG pressure includes downcomer liquid head.
(O Peak RCS pressure includes elevation head.
l I
l t
i CALVERT CLIFFS UFSAR 14.26-15 Rev. xx