Large and Small Break LOCA Analysis/Result
Course 22.39, Lecture 15 11/1/06
Professor Neil Todreas
11/1/06 22.39 L ecture 15: Large Break LOC A Anal ysis/Result 1
Professor N e il Todre a s
Event Categories
Frequency/Reactor
Description
Category
(from Lillington, UK)
11/1/06
2
|
1 |
Conditions that occur regularly in normal operation |
~10 |
|
2 |
Faults that are expected during the life of the plant: Anticipated moderately frequent events requiring safety response |
~1 |
|
3 |
Faults not expected during the life of a particular plant: Anticipated infrequent events requiring safety response |
~10 -2 |
|
4 |
Improbable events not expected to occur in the nuclear industry but provided for by the design |
~10 -4 |
|
5 |
Extremely improbable events not provided for in the design of the plant |
~10 -6 |
Figure by MIT OCW . After Lillington, 1995.
22.39 Lecture 15: Large Break LOCA Analysis/Result 2
Professor Neil Todreas
Example Events
(from Lillington, U K )
Categories
Events
|
Bringing the Reactor to Full Power |
1 |
|
Loss of External Grid Loss of Feedwater Loss of Reactor Coolant Pump |
2 |
|
Small LOCA V alves Open |
3 |
|
Lar ge LOCA Main Steam Line Break |
4 |
|
LOCAs without ECCS T ransients with T otal Loss of ON- and Of f-Site Power |
5 |
Figure by MIT OCW . After Lillington, 1995.
11/1/06 22.39 Lecture 15: Large Break LOCA Analysis/Result 3
Professor Neil Todreas
High pressure make-up water supply at ambient temperature
Energy Outflows as Steam and W ater
Bleed and Feed
Primary coolant circuit
Mass lost as steam = GOOD
Low mass flow rate High enthalpy change
Heat input
from core Mass lost as water = BAD
High mass flow rate Low enthalpy change
Figure b y MIT OC W .
Decay Power and Integral Decay Power As A Function of Time
By C h arl e s F o rsberg. C o urtesy o f Oak Rid ge N ati o n a l Laboratory.
• P ressurize d Water R e actor
• S NF Burnu p : 33 GW(d )/MTIHM
0.1 0.1
Decay Power [MW(th)]/[MW(th) Reactor Output]
0.05 0.05
0.02 0.02
0.01 0.01
0.005 0.005
0.002 0.002
0.001 0.001
0.01 0.02 0.05 0.1 0.2 0 .5 1.0 2 .0 5.0 1 0 2 0 5 0 1 00
Time (d)
Integral Decay Power [MWd]/[M W(th) Reactor Output]
11/1/06 22.39 L ecture 15: Large Break LOC A Anal ysis/Result Professor N e il Todre a s
5
05-017
Diagra m m atic representation of PWR prim ary and secondary circuits and the emergency cooling systems
Diagram removed d u e to cop yright res t rictions.
Figure 4.4 in Collier, J. G., an d G. F. Hewitt. Introduction to Nuclear P o wer . Washington, DC: Hemispher e Publishing, 1987.
11/1/06 22.39 L ecture 15: Large Break LOC A Anal ysis/Result 6
Professor N e il Todre a s
Accumulator
Inside containment
Outside containment
Boron injection tank
Refueling water storage tank
High pressure pump
Steam generator
Intermediate pressure pump
Cold leg
Hot leg
Low pressure pump
Pump
Normal core cooling system
EMERGENCY CORE COOLING SYSTEM (SCHEM A TIC)
Figure b y MIT OCW .
Connection pipe diameter/cross sect ion/percentage spectrum of a PWR. (Solid lines) Prim ary loop system ; (dashed lines) pressurizer
Diagram removed d u e to cop yright res t rictions.
Figure 4.21 in Collie r, J. G., and G. F. Hewitt. Introduction to Nuclear Power . Washington, DC: Hemispher e Publishing, 1987.
11/1/06 22.39 L ecture 15: Large Break LOC A Anal ysis/Result 8
Professor N e il Todre a s
Appendix C: Basic Assumptions of the LOCA
1. The reactor has been operating for an infinite period of time at an assumed slight overpower condition. No power or other transient preceeds the accident.
2. Peak core power density or lin ear power generation is at the maximum allowable value.
3. A double-ended rupture of one primary coolant loop is assumed (largest existing pipe)
PWR: rupture of cold leg imposes most severe conditions BWR: rupture of a recirculation loop
Remaining intact loops continue their operation as dictated by available electric supply or stored rotational energy.
4. Off-site power is lost upon initiation of the accident and is restored after several days. (continued)
11/1/06 22.39 L ecture 15: Large Break LOC A Anal ysis/Result 9
Professor N e il Todre a s
5. Reactor scram systems need not contribute to the nuclear shut down because voiding of the core provides suffici ent negative reactivity for shutdown.
6. The reactor is isolated after the initiation of the accident,
i.e. the regular heat sink is removed.
PWR: Upon initiation of the accident the steam generators are isolated on th e secondary side by closing the steam supply valves and the feedwater valves.
BWR: Upon receipt of a reactor-vessel low-water signal the main steam isolation valves close within 10 seconds. Feedwater flow ramps to zero within four seconds. (continued)
11/1/06 22.39 L ecture 15: Large Break LOC A Anal ysis/Result 10
Professor N e il Todre a s
7. EECS are actuated automatically b y appropriate signals. No corrective operator action is assum e d for the first 10 minutes following initiation of the accident.
8. A single failure criterion is applied to the reactor system whereby an additional fault is postulated which may render inoperative any one of the following:
 Mechanical active com ponents (e.g. pum p )
 Active or passive electrical components
11/1/06 22.39 L ecture 15: Large Break LOC A Anal ysis/Result 11
Professor N e il Todre a s
Events in the reactor pressure vessel during a large-break LOCA. a) Nor m al operation; b) blowdown phase; c) refill phase; d) reflood phase
Diagram removed d u e to cop yright res t rictions.
Figure 4.18 in Collie r, J. G., and G. F. Hewitt. Introduction to Nuclear Power . Washington, DC: Hemispher e Publishing, 1987.
11/1/06 22.39 L ecture 15: Large Break LOC A Anal ysis/Result 12
Professor N e il Todre a s
PWR operating conditions
Diagram removed d u e to cop yright res t rictions.
Figure 4.5 in Collier, J. G., an d G. F. Hewitt. Introduction to Nuclear P o wer . Washington, DC: Hemispher e Publishing, 1987.
11/1/06 22.39 L ecture 15: Large Break LOC A Anal ysis/Result 13
Professor N e il Todre a s
Adiabatic heat-up for PWR fuel (17 x 17)
Diagram removed d u e to cop yright res t rictions.
Figure 4.1 in Collier, J. G., an d G. F. Hewitt. Introduction to Nuclear P o wer . Washington, DC: Hemispher e Publishing, 1987.
11/1/06 22.39 L ecture 15: Large Break LOC A Anal ysis/Result 15
Professor N e il Todre a s
NRC Appendix K Criteria
1) Pea k Claddin g Te mperature . The calculated m axi mum fuel ele m e n t cladding te mperature shall not exceed 2200 ï‚° F.
2) Maxi m u m C l addin g Oxidation . The calculated total oxidation of the cladding shall nowhere exceed 0.17 ti mes the total cladding thickness before oxidation.
3) Maxi m u m Hydroge n Generation . The calculated total a m ount of hy drogen generated fro m the che m ical reaction of the cladding with water or stea m shall not exce ed 0.01 ti m e s the hypothetical a m o unt that would b e generated if all the metal in the cl adding cylinders su rrounding the fuel, excluding the cladding surrounding th e plenu m volu m e, were to react.
4) Coolabl e Geom e t ry . Calculated changes in co re geo m etr y shall be such that the core rem a ins am enable to cooling.
5) Long-Ter m C o oling . After any calculated successful initial operation of the ECCS, the calculated core tem p er ature shall be m a intained at an acceptable low value and decay heat re moved for the extended period of ti me re quired by the long-lived ra dioactivity re mai n ing in the core .
11/1/06 22.39 L ecture 15: Large Break LOC A Anal ysis/Result 16
Professor N e il Todre a s
Variation of peak clad temperature with time for a large-break LOCA
Diagram removed d u e to cop yright res t rictions.
Figure 4.19 in Collie r, J. G., and G. F. Hewitt. Introduction to Nuclear Power . Washington, DC: Hemispher e Publishing, 1987.
11/1/06 22.39 L ecture 15: Large Break LOC A Anal ysis/Result 17
Professor N e il Todre a s
Schem a tic calculated fuel clad temperatures for a PWR LOCA
Diagram removed d u e to cop yright res t rictions.
Figure 5- 9 in Nero, A. V. A Guidebook to Nuclea r Rea c tor s .
Berkeley, CA: University of C a lif ornia P r ess, 1979. ISBN : 0520034821
11/1/06 22.39 L ecture 15: Large Break LOC A Anal ysis/Result 18
Professor N e il Todre a s
Significant Phenomenon
T emperature ( o C)
Tem perature at which significant phenom ena occur during core heat-up
|
350 |
Approximate cladding temperature during power operation. |
|
800-150 |
Rod internal gas pressure in the post-accident environment causes cladding to perforate or swell Some fission gases release Solid reactions begin stainless steels and Zircaloy Clad swelling may block some flow channels |
|
1450-1500 |
Zircaloy steam reaction may produce ener gy in excess of decay heat Gas absorption embrittles Zircaloy Hydrogen formed Steel alloy melts |
|
1550-1650 |
Zircaloy-steam reaction may be autocatalytic unless Zircaloy is quenched by immersion. |
|
1900 |
Zircaloy melts |
|
2150 |
Increasingly significant fission product release from UO 2 |
|
2700 |
UO 2 and ZrO 2 melt |
Figure by MIT OCW . After Hewitt and Collier.
11/1/06 22.39 Lecture 15: Large Break LOCA Analysis/Result 24
Professor Neil Todreas
Hole ar ea r equir ed to r eject decay heat as steam (m 2 )
0.01
0.005
0
1 10 100
T ime (seconds)
1000
10,000
Hole Size to Remove Decay Heat as Steam
11/1/06 22.39 L ecture 15: Large Break LOC A Anal ysis/Result Professor N e il Todre a s
F i g u r e b y M I T O C W .
28
Primary pressure vs ti m e for sm all-break LOCAs in a PWR. (  ) Primary tem perature 175  C; (  ) reflood tank empty. (Two HPI pumps; reflood tanks 4 x 286 m 3 ; no LPI pumps)
Diagram removed d u e to cop yright res t rictions.
Figure 4.22 in Collie r, J. G., and G. F. Hewitt. Introduction to Nuclear Power . Washington, DC: Hemispher e Publishing, 1987.
11/1/06 22.39 L ecture 15: Large Break LOC A Anal ysis/Result 30
Professor N e il Todre a s
Small-break LOCA; mixture level and clad temperatures
Diagram removed d u e to cop yright res t rictions.
Figure 4.27 in Collie r, J. G., and G. F. Hewitt. Introduction to Nuclear Power . Washington, DC: Hemispher e Publishing, 1987.
11/1/06 22.39 L ecture 15: Large Break LOC A Anal ysis/Result 31
Professor N e il Todre a s
