ML030660549

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Revised Emergency Plan Position Specific Procedure EP-PS-325
ML030660549
Person / Time
Site: Susquehanna  Talen Energy icon.png
Issue date: 02/27/2003
From:
Document Control Desk
To:
Susquehanna
References
Download: ML030660549 (11)
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Feb. 27, 2003 Page 1 of 1 MANUAL HARD COPY DISTRIBUTION DOCUMENT TRANSMITTAL 2003-9336 FLAIM*LA R L B EMPL#:23244 CA#: 0363 Address: CSA2 Phone#: 2 4- 658 TRANSMITTAL INFORMATION:

Tv: PLAIULAUI *-w---" 02/27/2003 LOCATION: DOCUMENT CONTROL DESK

-ROM: NUCLEAR RECORDS DOCUMENT CONTROL CENTER

  • NUCSA-2)

ýHE FOLLOWING CHANGES HAVE OCCURRED TO THE HARDCOPY OR ELECTRONIC MANUAL ASSIGNED TO YOU:

325 - 325 - SYSTEMS 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-325 REPLACE: REV:5 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 7 EP-PS-325-7 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 7 EP-PS-325-7 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.

"* Ifthe core experienced an overpower/pressure transient, a gap release may have occurred.

"* Ifthe core experienced a mechanical failure, which could produce flow blockage, there may be localized fuel melt.

"* Ifthe 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.

"* Ifthe 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 7 EP-PS-325-7 Containment Radiation Monitor Response Direct Release Path to Dry well (Sprays Off) 1.E+07 1.E+06 1.E+05 1.E+04 f.-.

1.E+03 E

C 1.E+02 0

00 1.E+01 CD 4t 1.E+00 O,

1.E-01 1.E-02 1.E-03 1.E-04 1.E-05 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 7 EP-PS-325-7 Containment Radiation Monitor Response Direct Release Path to Dry well (Sprays On) 1.E+07 1.E+06 100%

1.E+05 -50%

10% -- 100%

100%

1.E+04-- 5%%_ 50%o "1% 10% 50%

100%

1

-10%

1.E+03 5 0- 50%

"5% S~1%

E 10%

o 1.E+02 0

21.E+00 cc 0100%

0 1.E-02

  • w 10%0" 1.E-02  %- .s%.,3o
  • --- 10% -10 1.E-04 5%

1.E-04 I Ih h 4h 2h I 24h 1I hh 24 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 7 EP-PS-325-7 Containment Radiation Monitor Response Direct Release to Wetwell and Not to Drvwell 1.E+06 1.E+05 1.E+04 1.E+03 1.E+02 E

"E 1.E+01 00 0 1.E+00 C)

C)

"1 1.E-01 o1.E-02 0

1.E-03 1.E-04 1.E-05 1.E-06 lh 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 1C EP-AD-000-457, Revision 6, Page 5 of 10

Tab 7 EP-PS-325-7 CONTAINMENT HYDROGEN VS CORE DAMAGE

%Mtap-Waw~ Rleaction &Com Damnage State s0 40 30 40. Pove,Mel Thrig 2D 10 Clad~Failsel 0

0.1 1 10 100 H2 % In Contanment

-*BWRMk I&I11 saw=s~ NUREG/CR(-2726. p. 4-3:. damage stts NUREG-i4S2 Vol.5.;.

ThfJ pmemmge, NUREG-1370:- NUR~EGICR-4041: NUREGICR-S557, Table 4.9. p. 71.

valMr

'zmdiY' ATTACHMENT 2 EP-AD-000-457, Revision 6, Page 6 of 10

Tab 7 EP-PS-325-7 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 212 0 F.

ATTACHMENT 3 (Page 1 of 4)

I-kv EP-AD-000-457, Revision 6, Page 7 of 10

Tab 7 EP-PS-325-7 WATER INJECTION REQUIRED TO COOL CORE BY BOILING While the top of the active core is uncovered, assume that the fuel vili heat up at 1-2"F/sec. The increased core temperature will result in fuel pin damage as shown below.

NOTE:

These estimates are r asonable (factor of

2) if the core is uncovered vithin a -4WCF few hours of shutdown (including failure to R5a'sn ofm oJsfasnpo scram). If there in fro tu ,, "r sufficient injection, core heatup may be stopped or sloved due to steam cooling.

steam cooling may not prevent core damage under accident conditions. - .,,iF

'WY=Pd isa f b~ftmn cowiim

- a w4nobi gem*s

  • **oF V- arf *s - - - mWn po~at cldg bumu - raisi of om pin gMP 4

SOUrce: NIORG-0900, 2=G/J-4524, NURZG-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: 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 6, Page 8 of 10

Tab 7 EP-PS-325-7 WATER INJECTION REQUIRED TO COOL CORE BY BOILING 110=710TN (gpu) MEQURiD To REPIAcE WATER LOST

.3? BOILING M7E TO DECAY BEAT FOR A 3000 101(t)

PLPMI (1/2-2 4 HOURS AFTER SHUTDOWN) 200 250 2=O "150 "100 ISO so.

0.5 2 3 -4 5 10 15 24 N@Luu- Altev- ShUCkW,, IDA INJECTION (gp=) REQUIRED TO REPL.CE WATER LOST BY BOILING DUE TO DECAY HEAT FOR A 3000 10(t)

PlAM (I to 30 DAYS AFTER SHUTDOWN) ow Do) lc IC InJecteco "10O 80 sO 40 40 20 20 0 S 9 9 9 9 t IJ

"*1 2 3 4 5 6 7 a 9 1o 20 3ad yA" Al'%w Shutckmn ATrACHMENT 3 (Page 3 of 4)

EP-AD-000-457, Revision 6, Page 9 of 10

Tab 7 EP-PS-325-7 WATER INJECTION REQUIRED TO COOL CORE BY BOILING Core damage vs. time that reactor core is meovered "Ti PWR or20% of BWR amce core is e unovered (h) (IF) (c) Possible core damage 0 >600 >315 & None 0.5 to 0.75 180D-2400 980-1300

  • Locaf fael meltin
  • Burming of claddin with steam produbtia (exoftmic Zr-H.0 reaction with apid E6 generation)
  • apid fbl claddin faglure (gap release from the Come')

0.5 to 1.5 2400-4200 1300-2300 e Rapid r-lease of volatile fission p-duets ('t-ves seve= cot damage release fom core) o Possible relocation (slump) of molten core

  • Possible uncoolable core I to 3+ >4200 >2300

NUREG-006. NUREG-t 150. and NUREG-1465.

ATTACHMENT 3 (Page 4 of 4)

EP-AD-000-457, Revision 6, Page 10 of 10

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