ML030660448
| ML030660448 | |
| Person / Time | |
|---|---|
| Site: | Susquehanna  |
| Issue date: | 02/27/2003 |
| From: | Pennsylvania Power & Light Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| EP-PS-324, Rev 9 | |
| Download: ML030660448 (11) | |
Text
Feb. 27, 2003 Page 1 of 1 MANUAL HARD COPY DISTRIBUTION DOCUMENT TRANSMITTAL 2003-9388 USER INFORMATION:
FLA *LAUREL EMPL#:23244 CA#: 0363 Address: CSA2 Phone 254-3658 TRANSMITTAL INFORMATION:
Tt0: Fn-I*AMtLAo~n D u '02/27/2003 LOCATION: DOCUMENTCONTROL DESK PROM: NUCLEAR RECORDS DOCUMENT-CONTROL CENTER
ý..NUCSA-2)
HE FOLLOWING CHANGES HAVE OCCURRED TO THE HARDCOPY OR ELECTRONIC MANUAL ASSIGNED TO YOU:
324 - 324 - FUELS LEAD ENGINEER - EMERGENCY PLAN POSITION SPECIFIC PROCEDURE REMOVE MANUAL TABLE OF CONTENTS DATE: 08/26/2002 ADD MANUAL TABLE OF CONTENTS DATE: 02/26/2003 CATEGORY: PROCEDURES TYPE: EP ID: EP-PS-324 REPLACE: REV:9 UPDATES FOR HARD COPY MANUALS WILL BE DISTRIBUTED WITHIN 5 DAYS IN ACCORDANCE WITH DEPARTMENT PROCEDURES. PLEASE MAKE ALL CHANGES AND ACKNOWLEDGE COMPLETE IN YOUR NIMS INBOX UPON RECEIPT OF HARD COPY. FOR ELECTRONIC MANUAL USERS, ELECTRONICALLY REVIEW THE APPROPRIATE DOCUMENTS AND ACKNOWLEDGE COMPLETE IN YOUR NIMS INBOX.
Tab 4 EP-PS-324-4 CORE DAMAGE ESTIMATE I (Primary System Breach Inside Containment)
NOTE: It is important to quickly provide a status of the present situation and a prognosis on whether the situation is expected to degrade, improve, or remain the same, (i.e., within 5 to 10 minutes of a change in plant status).
1.0 INDICATORS USED 1.1 Containment Radiation Use Attachment 1, A, B, or C, as applicable, to determine the amount and type of
-fuel damage using containment radiation monitors. These figures were taken from the US NRC Response Technical Manual, RTM-96. Obtain the containment radiation levels from SPDS or the Control Room indicators.
NOTE (1): Correction for the pre-release background radiation levels may be required as listed below.
Gap or In-Vessel Melt - The background radiation monitor value is normally low (< 4 R/hr) relative to 1% gap or in-vessel melt release. Consequently, the monitor reading does not require correction for background level in determining the type and amount of fuel damage. If the background radiation monitor reading is > 4 R/hr, the monitor reading should be corrected for the background level in determining the type and amount of fuel damage.
Spiked or Normal Coolant - The radiation monitor value requires correction for the background level. Correct the monitor reading to account for the normal background level in determining the type and amount of fuel damage.
NOTE (2): Containment radiation will go up if there is fuel damage. The increase will depend on the type of fuel damage, and whether or not there was a LOCA, Drywell and/or Wetwell sprays were used, and the amount of blowdown from the Reactor Vessel to the Suppression Pool.
In the case of a LOCA, the fuel damage estimate depends strongly on whether or not containment sprays are being used.
Special care should be taken to confirm the operation of containment sprays.
EP-AD-000-457, Revision 6, Page 1 of 10
Tab 4 EP-PS-324-4 1.2 Containment Hydrogen Use Attachment 2, taken from the US NRC Response Technical Manual RTM 96, to determine the amount and type of fuel damage using Hydrogen Concentration. Obtain the containment Hydrogen levels from SPDS or the Control Room indicators.
NOTE: Containment Hydrogen will increase if there is a LOCA inside the containment and significant fuel damage.
1.3 Coolant Fission Product Concentration vs. Core Damage
.Coolant sampling will indicate the amount of fuel damage, but in most cases, will take too long for use in dose projections. If PASS sample data becomes available, the Nuclear Fuels Engineer is responsible for assuring a fuel damage calculation based on the measured fission product inventories is performed. The results of this analysis should be compared to previous calculations using other methods.
1.4 Plant Transient Precipitating Fuel Damage Ifthe core experienced a loss of coolant accident and is not covered within 15 minutes, refer to Attachment 3 taken from the US NRC Response Technical Manual RTM-96. The amount of time the core was uncovered can be determined using SPDS. Using the attached figures will provide an estimate of potential fuel damage. Coolant samples must be taken to accurately assess fuel damage.
The type of transient experienced by the reactor leading to fuel damage can be an indicator of the amount and type of fission products released.
"* If the core experienced an overpower/pressure transient, a gap release may have occurred.
"* If the core experienced a mechanical failure, which could produce flow blockage, there may be localized fuel melt.
"* If the core experienced a mechanical perturbation, such as a seismic event or a large steam line break causing a large delta pressure across the core, a gap release could result.
"* If the Reactor failed to shut down (ATWS) with a subsequent loss of cooling, there may be fuel melt.
EP-AD-000-457, Revision 6, Page 2 of 10
Tab 4 EP-PS-324-4 Containment Radiation Monitor Response Direct Release Path to Dry well (Sprays Off) 1.E+07 0 1.E+06 .100%
=-50%%°0 0100%5 1.E+05 -10% 50%. .10o%
5 O% " 100%
-* 5
- ~1.E+042 1 _ -% l% 5%
E C, %1.E+0 "
0) 0 1 "0 1.E+01 -100%
- 5 u2 50%50%
0 ("1.E+002°/ CD ~50%1 *1oO 10 100 o 1.E-01 -O 10%%
1%
5% .=-50%
1.E-03 -_1%- 10%
-55%
1.E-04 1%.
1 .E-05 I 1h 24h Ih 24h I1h 24 1Ih 24h In-Vessel Melt Gap Spiked Coolant Normal Coolant Core Damage and Time After Reactor Shutdown Note 1: This figure should be used only when there is a primary system breach inside containment and a direct release path to the Drywell.
Note 2: See Attachment 3 to determine if fuel melt occurred (core uncovered or fuel blockage).
ATTACHMENT 1A EP-AD-000-457, Revision 6, Page 3 of 10
Tab 4 EP-PS-324-4 Containment Radiation Monitor Response Direct Release Path to Dry well (Sprays On) 1.E+07 1.E+06 1.E+05 -50%
=-10%"" 100%
100%
1.E+04_- 5% :W-50% __
-1% -10% 50%'
S-- 100%
10%
1.E+03*" 5% ~~5%
5 ' -50%
- O E 1% _ -110%
"E 4,
1.E+02 0
0 1.E+01 a)
"Z I.E+00 0_.100%
0 o 1.E-01 0 ::_10%
T--100%,
1.E-02 5% 50% lOOO/=
=50%
-0% .. %1 .100% 1 1.E-03 -5% -
==-10%
H-5 50 1% - =---10% i 1.E 04 1 1.E-05 1h 24h lh 24h lh 24h lh 24h In-Vessel Melt Gap Spiked Coolant Normal Coolant Core Damage and Time After Reactor Shutdown Note 1: This figure should be used only when there is a primary system breach inside containment and a direct release path'to the Drywell.
Note 2: See Attachment 3 to determine if fuel melt occurred (core uncovered or fuel blockage).
ATTACHMENT 1 B EP-AD-000-457, Revision 6, Page 4 of 10
Tab 4 EP-PS-324-4 Containment Radiation Monitor Response Direct Release to Wetwell and Not to Drywell 1.E+06 1.E+05 1.E+04 1.E+03 1.E+02 Ca E
"1E 1.E+01 0
0 1.E+00 C,,
"a 1.E-01 Cr.
ý0 0
S1.E-02 0
1.E-03 1.E-04 1.E-05 1.E-06 lh 24h lh 24h lh 24h lh 24h In-Vessel Melt Gap Spiked Coolant Normal Coolant Core Damage and Time After Reactor Shutdown Note 1: This figure should be used only when there is a primary system breach inside containment and a direct release path to the Wetwell without a primary release to the Drywell.
Note 2: See Attachment 3 to determine if fuel melt occurred (core uncovered or fuel blockage).
ATTACHMENT 1 C EP-AD-000-457, Revision 6, Page 5 of 10
Tab 4 EP-PS-324-4 CONTAINMENT HYDROGEN VS CORE DAMAGE
% eLsaM-Was Rleaction & Core Dwnage State 40 30 4<m Ponaftl Melt Through 20 4., Poss~icl Uzwooable core 10 *S= .. Mek 0
0.1 I 10 100 H2 % In Containment
-*BWRMk I&I11 S==. NupmEicR-2726. p. 4-3: damzge stats, NUREG4524. Val. 5.;
fhld pe:rec-ge, NUnr.EG-13'70: NUREGCY=4041; NUREGICR-5W6, Table 4.9, p. 71, amf dry. voblr ATTACHMENT 2 EP-AD-O0O-457, Revision 6, Page 6 of 10
Tab 4 EP-PS-324-4 WATER INJECTION REQUIRED TO COOL CORE BY BOILING CAUTION:
I These rates are those required to remove decay heat from a 3000 MW(t) plant by boiling. If
- there is a break requiring make up or injected water, more water than indicated will be required to both keep the core covered and cooled.
CAUTION:
Ifthe core has-been uncovered, the fuel temperature will have increased significantly.
Additional flow will be required to accommodate the heat transfer necessary to return to equilibrium fuel temperature.
NOTE:
These curves are based on a 3000 MW(t) plant operated at a constant power for an infinite period and then shutdown instantaneously. The decay heat power is based on ANS-5.1/N18.6.
Assuming the injected water is at 800 F, these curves are within 5% for pressures between 14 psia to 2500 psia. These curves are within 20% for injected water temperatures up to 212 0 F.
ATTACHMENT 3 (Page 1 of 4)
EP-AD-000-457, Revision 6, Page 7 of 10
Tab 4 EP-PS-324-4 WATER INJECTION REQUIRED TO COOL CORE BY BOILING While the top of the active core is uncovered, assume that the fuel will beat up at 1-2"F/sec. The increased core temperature will result in fuel pin damage as shown below.
Jaure A-NOTE:
These estimates are reasonable (factor of M4wg oft*J VSkI 0
- 2) if the core is uncovered within a few hours of shutdown (including failure to scram). If there is sufficient injection, - 42W0F core heatup say be stopped or slowed due to steam cooling.
Steam cooling may not prevent core damage under accident Fmn.imW dt"1quiWad*ir*jW conditions. disaokme in mmfd ;nnM~
VWy rapi _- = m-
~smof H2 id 9x&t *J( i
- law,'
PO~sd~i burs - maiming of amo pradu infuel pin gap
- 12C0F
- U*F NommW g lairrpaxe I
curoe: NUuzc-Og00, ]RZGC/-4524, NIURWG-0956 ATTACHMENT 3 (Page 2 of 4)
CAUTION: If the core is severely damaged, it may not be in a coolable state even if covered again with water.
NOTE: Ifthere is sufficient injection, core heatup may be stopped or slowed due to steam cooling. Steam cooling may not prevent core damage under accident conditions.
EP-AD-000-457, Revision 6, Page 8 of 10
Tab 4 EP-PS-324-4 WATER INJECTION REQUIRED TO COOL CORE BY BOILING INJECTION (onm) = 96 WATER~ 1=S
-By BOILING WE TO DECAr BEAT FOR A 3000 101(t)
PlAN (1/2-24 HOURS AFTER SHUTDOWN) 250 s.o 2 3 -4 5 10 95 24-b4o.w-M At*e- Shuirdn n MY InJczTiox (gpm) PREQunmE To RzPlhcE WATER LOST BY BOILING DXUE TO DECAY HEAT FOR A 3000 10T(t)
PLANT (1 to 30 DAYS AFTER SHUTDOWN)
C0m Do?IaI r nIat.
I00 so "so 60 40 20 20
! ! I * ! !t a
2 3 4 5 9 7 9 9710 20 30 00" Af~w 6iftutaw~
ATTACHMENT 3 (Page 3 of 4)
EP-AD-000-457, Revision 6, Page 9 of 10
Tab 4 EP-PS-324-4 WATER INJECTION REQUIRED TO COOL CORE BY BOILING Core damage vs. time that reactor core uncovered m
Time PWR or20% of BWR active core is o e e uncovered (h) (OF) (oC) Possble care damage 0 >600 >315
- None 0.5 to 0.75 1800-2400 980-1300
- Local fuel melting
- Burning of cladding with st*e prodm on (exothetmic Zr-H20 re- *on with rapid H2 ge=-rAion)
- Rapid fadl cadding haum (gap redef from the core) 0.5 to 1.5 2400-4200 1300-2300 e RaW reIease of volatile fission prod= (in-vessel seves coae damage release fomn cord) e Possible relocation (slump) of molten w Possible uncoolable core I to 3+ >4200 >2300 a Melt-thmrgh of vessel with possible comaine failure and release of additional less-volatile fission Sow-,=: NURE.CR-4245. NUREG/CR-4624. NUREGICR-462.9. NUREGKC-5374, NUREG-0900.
NMEG-0956. NUREG-I 150. and NtMEG-1465.
ATTACHMENT 3 (Page 4 of 4)
EP-AD-000-457, Revision 6, Page 10 of 10