ML031990308
| ML031990308 | |
| Person / Time | |
|---|---|
| Site: | Susquehanna  |
| Issue date: | 07/03/2003 |
| From: | Susquehanna |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| 28401 EP-PS-136, Rev 0 | |
| Download: ML031990308 (14) | |
Text
Jul. 03, 2003 Page 1 of 1 MANUAL HARD COPY DISTRIBUTION DOCUMENT TRANSMITTAL 2003-31679 USER INFORMATION:
Name:GERLACH*ROSE M EMPL#:28401 CA#:0363 Address: NUCSA2 Phone#: 254-3194 TRANSMITTAL INFORMATION:
TO:
GERLACH*ROSE M 07/03/2003 LOCATION:
DOCUMENT CONTROL DESK FROM:
NUCLEAR RECORDS DOCUMENT CONTROL CENTER (NUCSA-2)
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4 obed
Tab 2 EP-PS-1 36-2 CORE DAMAGE ESIMATE 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 7, Page 1 of 10
Tab 2 EP-PS-1 36-2 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 If the 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 7, Page 2 of 10
Tab 2 EP-PS-1 36-2 Containment Radiation Monitor Response Direct Release Path to Drv well (Sprays Off) 1.E07 1.E+06 -
100%
50%/
100%
1.E.05 -
10%
50%
100%
5%
50%
100%
1 0%
1.E+04 -
1%
10%
-50%=
1%
10%
c 1.E+03 1%
5%_
0
_1%
EC E 1.E+02 8
°00 I.E+01 -
100r/0 CD 0 ~ ~~~~~~~~~~~~0 2 1.E+00 10%
tl:
__ 100%
o 1.E-01 %
10 10%
- 2.
50%
1.E-02 10%
-0a i
i 5%
50h 1.1E-03 1%
10%/
-5%
1.E-04 1%
1.E-5 h12h 24h 1 h in 24h i
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 IA EP-AD-Oo-457, Revision 7, Page 3 of 10
Tab 2 EP-PS-1 36-2 Containment Radiation Monitor Response Direct Release Path to Drv well (Sprays On) 1.E+07 1.E+06 100%
1.E+05 50%
_ 10%
100%_ 1O:
10%~~~10 1.E+04 5%
-0%
1% =
10% _
=:10%
1.E-+03 50 L
as. +
-~~~~~5%
_W/
E 1%
10%
C
° 1.E+01 ID CD o-1%
.E-02 100%
i z~~~~~~~~~~~~~1_
10 1oo%
U
_ 1.E-03
-so%.
1.EZ2
-5%
1sox I woo%._5 5%-. _ _5%_~~~_1 150%-
100%o 1.E005 1h 24h 5h 24h 1h 24h 0
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 0
EP-AD-000-457, Revision 7, Page 4 of 10
Tab 2 EP-PS-1 36-2 Containment Radiation Monitor Response Direct Release to Wetwell and Not to Drvwell 1.E1+06 1.E+05 1.E+04 1.E+03
-=1.E+02 0
E
° 1.E+02 C
0C)
,,1.E+01
° 1.E-O 0,
0 C
01.vE-O1 0
o 1.E-01 I-0 I.E-03 I.E-04 1.E-05 1.E-06 1h 24h lh 24h 1h 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 7, Page 5 of 10
Tab 2 EP-PS-1 36-2 CONTAINMENT HYDROGEN VS CORE DAMAGE
% M Wae Pmcdon & Ce Dmage Stat s0 40 10 POs Ma nThoah 20 10 0
.Poit Ueoolbae core 4.Cad wbje 0.1 1
10 100 H2 % In Containment
- -WRMkI&N S== NUREGICR4726. p. 4-3; damage sbes, NUEG45.
Vol. S.:
TbU PELCM NURa-1370 NUREGCR4041; NUREGICI-S67, Table 4.9. p. 71 ATTACHMENT 2 EP-AD-000-457, Revision 7, Page 6 of 10
Tab 2 EP-PS-1 36-2 WATER INJECTION REQUIRED TO COOL CORE BY BOILING CAUTION:
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:
If the 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 2120F.
ATTACHMENT 3 (Page 1 of 4)
EP-AD-000-457, Revision 7, Page 7 of 10
I Tab 2 EP-PS-1 36-2 WATER INJECTION REQUIRED TO COOL CORE BY BOIUNG hiale the top of the active core s uncovered, assume that the fuel viti heat up at
-2 1/sec.
The ncreased core teratrei result in fuel pin damage as shown below.
NOT:
The" estbmtes are reasonable (factor of Mi d* 000nw 91
- 2) if the core s
ncovered vithin a
4wF fev hours of sntdmrn nie ia aI np~c (including failure to scram).
there is sufficient injection,
-20VF core heatup ay be stopped or slowed due CM to stam cooling.
Stam cooling may not 3aMF prevent core AMMage under accident F~n uJ-cond+/-tiona
- a We e Wiypd
.mmdciob Vsrj mml m -
borm~
F (Page sq 2apiof 4)f
-W nows at.
HE dra PCAUTIN I c baast -les d
ors inganwih a SCUMc: ZMM0900, ICMMG-0954 09S ATTACHMENT 3 (Page 2 of 4)
CAUTION: If the core s severely damaged, It may not be In a coolable state even f covered again with water.
NOTE: If there 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 7, Page 8 of 10
Tab 2 EP-PS-1 36-2 IAV=
I& IO%'nfM DOefl ItO0fl Tre t~f~el I -fM RV RnhII-I~r Tw-IE u-"ga' EItL-%p I E%.01 I 1"m..&
6~
E INJCTZON (M)
REZQUIBtED TO MU IC"=V ~ lOS
,BY BOILING MM TO DECILY M
F A 3000 201(t)
PLINT (1/2-24 HOURS LPIM SUTDOWN)
.c act e C
jecteo0 200 250 200 to so 10
. 5 1
2 ~~~~~~
4 A
bIi.Ater Stdmalwn I~
xwzEcTxON (gpm) REQunRD T REwlAcE WhAER LOS BY 8OILING VUE TO ECAY EAT FOR A 3000 101(t pLAN
(
to 30 DAYS LITER SHUTDOWN)
£0 I=
so 40 20 C*1 a
a 4aa 5
70a3a10 to
&W
- SiUtdou, 2t so ATTACHMENT 3 (Page 3 of 4)
EP-AD-000-457, Revision 7, Page 9 of 10
Tab 2 EP-PS-1 36-2 WATER INJECTION REQUIRED TO COOL CORE BY BOILING Core damage vs. time t reactor core is umcovered Tim PR or 20% of BWR ave coze s
()
(E)
(C)
Possible co damage 0
>600
>315
- Nae 0.5 o 0.75 1800-2400 9801300
- Lca ftel medmi
- Brmi4 of dadding ft sem prodnztio (ezofhermc Zx-Hp zeacion with upid I2 Senratio)
- Rapid fuel daddg arc (gap Idease from &C core*)
0.5 to 1S 24U-4200 1300-20
- Rid ease of volatile fission p
Cn-*E seere core daagt =
se from cor)
- Possible relocation (sltmp) of =ot=
core
- Possible unoolable core I to 3+
>4200
>2300 Melwdmhr of vessel with possible coninem il o
f and =16ae of addkional lessvolatile fission
~
~~pcs
.Sow=s NUR.EGfCR-424S. NUREGICR-4624. NUREGICR-46Z9. KUREGFCR MU7. NUJREG0900.
NUPREG-0956. KUtREG-I1 S. and NUREG-146S.
ATTACHMENT 3 (Page 4 of 4)
EP-AD-000-457, Revision 7, Page 10 of 10
Tab 3 EP-PS-1 36-3 CORE DAMAGE ESTIMATE II (Small or no primary system breach inside Containment)
This instruction provides a method of estimating the percentage of fuel that has failed using the Containment Post-Accident Radiation Monitor (CPARM) readings on panel 1C601 (2C601) during an accident. Since the Containment Post-Accident Radiation Monitor readings are readily available, this calculation provides a quick assessment of core damage. This estimate only applies if there is a small or no primary system breach within containment.
1.0 LIMITATIONS OF THE METHOD 1.1 This procedure will only determine qualitatively the amount of fuel damage. The method uses Containment Post-Accident Radiation Monitor Readings to calculate the percentage of failed fuel during an accident where the fission products are released from the fuel rod cladding. The methodology is based on assumptions with large uncertainties that can significantly affect the results.
1.2 To use this method, the accident scenario up to the time of the Containment Post-Accident Radiation Monitor Reading must be well understood-to estimate the fuel temperatures required by this procedure.
1.3 In addition, a Containment Post-Accident Radiation Monitor Reading and the time the reading was obtained must be available.
2.0 RESPONSIBILTES 2.1 The Nuclear Fuels Engineer, Lead Technical SuDort Engineer, or designee collects information and makes estimates and determinations described in this procedure.
3.0 INSTRUCTIONS 3.1 Determine if Cladding Failure, Fuel Overheat, or Fuel Melt has occurred:
3.1.1 Cladding Failure is expected if peak cladding temperature remains less that 22000F, but the Containment Post-Accident Radiation Monitor readings have increased.
3.1.2 Fuel Overheat is expected if peak cladding temperature exceeds 22000F, but the maximum volume-averaged fuel pellet temperature remains less that 45000F.
3.1.3 Fuel Melt is expected If any volume-averaged fuel pellet temperature exceeds 45001F.
EP-AD-000-270, Revision 6, Page 1 of 3
Tab 3 EP-PS-1 36-3 3.2 Since the fuel melt temperatures are dependent on the event progression, specific guidelines cannot be given to cover all scenarios. Some judgment will have to be made or specific temperature calculations will have to be performed during the event. However, the following provides guidelines for a few known scenarios.
3.2.1 If a main steamline high radiation trip causes the scram and the core remains covered, usually cladding failure can be assumed and is possibly due to debris fretting, short term DNB, or PCI. However, if channel flow blockage is suspected, overheat or melting may occur.
3.2.2 For loss-of-inventory-after-the-reactor-is-shutdown scenarios, use to Tab 4 to estimate if Fuel Melt has occurred.
3.3 Determine the Time After Reactor Shutdown that a Containment Post-Accident Radiation Monitor Reading was obtained.
3.4 Determine if the event has resulted in a primary system breach inside primary containment (increase in drywell pressure/temperature and inventory makeup to the vessel is required to maintain level in the vessel). If the total primary system water released to the drywell is equivalent to less than 9,000 gallons or no primary system breach has occurred inside primary containment, use Figure 1.
Otherwise, use Core Damage Estimate I (Tab 4).
Note: The 9,000 gallon value is-about 10% of the fluid volume of the reactor vessel and primary piping (main steam, reactor recirculation, and feedwater).
3.5 Determine Fraction of Fuel Failed (FFF) as follows:
CPARM Reading FFF-E0rx 100 Expected I100% Fuel Failure CPARM Reading
- EP-AD-000-270, Revision 6, Page 2 of 3
Tab 3 EP-PS-136-3 FIGURE A CONTAINMENT HIGH RANGE RADITON MONITR READINGS THAT ARE EXPECED WIrH 100% OF THE FUEL FAILED FOR AN EVENT WITH NO PRIMARY SYSTEM BREACHf INSIDE CONWNMENT 19 z
C C
S.
Ds 0
°-
i Cl S
L
.......s.........
........ ~.
t---a --- ---- --.- *----
4...................................
LOC 4
\\
\\
F MEL T
. 4-
~
~
~~~..................
a 44...
.4.
0
- .1 S
./...................
I
-P R1a OVERHEAT<.bv t
I 1
A........
.4.
4.a..
I':.
..... 0....
'CLADDING FAIWU~
rr
- f v
4 q
q
- v 6
5
- at io 25 Post-Shutdown Time (HRS)
EP-AD-000-270, Revision 6, Page 3 of 3