ML031960415
| ML031960415 | |
| Person / Time | |
|---|---|
| Site: | Susquehanna  |
| Issue date: | 06/27/2003 |
| From: | Susquehanna |
| To: | Document Control Desk, Office of Nuclear Security and Incident Response |
| References | |
| 28401 EP-PS-136, Rev 0 | |
| Download: ML031960415 (26) | |
Text
Jun. 27, 2003 Pagel1 of 1 MANUAL HARD COPY DISTRIBUTION DOCUMENT TRANSMITTAL 2003 -30998 USER INFORMATIO N GERLAC OSE M EM.[PL#: 28401 CA*:0363 dress UCSA2 ho 254 TRANSMITTAL INFORMATION:
TO: :zz:.:-Ilrr.:!. rs 06/27/2003 LOCATION: DOCUMENT CONTROL DESK .
FROM: NUCLEAR RECORDS DOCUMENT CONTROL CENTER (NUCSA-2)
THE FOLLOWING CHANGES HAVE OCCURRED TO-THE HARDCOPY OR ELECTRONIC MANUAL ASSIGNED TO YOU:
136 - 136 CORE THERMAL HYDRAULICS ENGINEER REMOVE MANUAL TABLE OF CONTENTS DATE:
ADD MANUAL TABLE OF CONTENTS DATE: 06/26/2003 CATEGORY: PROCEDURES TYPE: EP ID: EP-PS-136 ADD: REV: 0 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.
PROCEDURE COVER SHEET PPL SUSQUEHANNA, LLC NUCLEAR DEPARTMENT PROCEDURE CORE THERMAL HYDRAULICS ENGINEER: EP-PS-136 Emergency Plan-Position Specific Instruction Revision 0 Page 1 of 3 QUALITY CLASSIFICATION: APPROVAL CLA6t1lCION:
( ) QA Program (X) Non-QA Program ( ) PlaRt? / "j ) Non-Plant (X In4) t on EFFECTIVE DAT& '
PERIODIC REVIEW FREQUENCY:. Two Years PERIODIC REVIEWV0U4 IAT 4 G-24:f RECOMMENDED REVIEWS:.4 - '4 ALL Procedure Owner: Nuclear Emergency Planning
('4 .,, #ij Responsible Supervisor: Manager-Station Engineering Responsible FUM: Supervisor-Nuclear Emer. Planning Responsible Approver: V.P.-Nuclear Operations FORM NDAP-QA-0002-1, Rev. 3, Page 1 of 1
EP-PS-136 Revision 0 Page 2 of 3 CORE THERMAL HYDRAULICS ENGINEER: Emergency Plan Position Specific Procedure WHEN: When the TSC is activated HOW NOTIFIED: Paged/Telenotifications System WHERE TO REPORT: TSC REPORT TO: Technical Support Coordinator OVERALL DUTY:
Provide fuel assessment and status information to the Technical Support Coordinator in support of dose projections, emergency classifications, protective action recommendation, and information dissemination.
MAJOR TASKS: TAB: REVISION:
Perform fuel damage calculations in TAB A 0 support of dose projections, emergency classifications, protective action recommendation process, and information dissemination.
EP-PS-1 36 Revision 0 Page 3 of 3 SUPPORTING INFORMATION: TAB:
Fuel Damage Worksheet Tab 1 Core Damage Estimate I Tab 2 (Primary System Breach Inside Containment)
Core Damage Estimate II Tab 3 (Small or No Primary System Breach Inside Containment)
REFERENCES:
NRC RTM 92, Nuclear Regulatory Commission Response Team Ma-nual SSES Emergency Plan NUREG 0654, Planning Standards and Evaluation Criteria NUREG 0731, Guidelines for Utility Management Structure and Technical Resources, Sept. 1980 NUREG 0696, Functional Criteria for Emergency Response Facilities NEDO 22215, Procedure for the Determination of the Extent of Core Damage Under Accident Conditions
TAB A EP-PS-1 36-A Revision 0 Page 1 of 2 MAJOR TASK:
Perform fuel damage estimates in support of dose projections, emergency classifications, protective action recommendation process, and information dissemination.
SPECIFIC TASK: HOW:
- 1. Notify the Technical Support 1a. Determine the amount and type of fuel Coordinator, the RPC, the TSC Dose damage using the Core Damage Calculator, the Dose Assessment Estimate "HELP" tabs and complete the Supervisor in the EOF, and the Fuel Damage Worksheet.
Engineering Support Supervisor in the EOF of initial fuel damage estimate. NOTE:
Core Damage Estimate I should be used when there Is a primary system breach Inside containment. Core Damage Estimate II should be used when there Is a small or no primary system breach Inside containment.
HELP CORE DAMAGE ESTIMATE I (Primary Syistem Breach Inside Containment)
See TAB 2 HELP CORE DAMAGE ESTIMATE II (Small or No Primary System Breach Inside Containment)
See TAB 3 HELP FUEL DAMAGE WORKSHEET See TAB 1 NOTE:
Work with the responding engineering staff In the TSC as required to complete Fuel Damage Worksheet.
TAB A EP-PS-1 36-A Revision 0 Page 2 of 2 SPECIFIC TASK: HOW:
- 2. Refine the fuel damage estimate as 2a. Use the instructions located in the Core more data becomes available. Damage Estimate "HELP" tabs.
Provide this information to the Technical Support Coordinator, the 2b. If PASS data becomes available, contact RPC, the TSC Dose Calculator, EOF Nuclear Fuels Engineering to perform a Engineering Support Supervisor, and fuel damage calculation and transmit EOF Dose Assessment Staffer. PASS data.
- 3. Ensure Dose Assessment Staffer and 3a. Work with the TSC Engineering Staff to TSC Dose Calculator has the fuel obtain needed information.
damage estimates and other needed information required to perform Dose Calculations.
- 4. Provide technical information to the NRC Response Team and MOC Technical Briefers as required.
Tab 1 EP-PS-1 36-1 FUEL DAMAGE WORKSHEET General The following information should be kept current at all times after facility activation. This information is used by Dose Calculators to perform dose projections for support of Protective Action Recommendations required within fifteen minutes of a General Emergency classification.
Engineering Support is required to provide an estimate of percent fuel damage to the TSC Dose Calculator and the EOF Dose Assessment Staffer in a timely manner, aflowing sufficient time for a dose projection to be performed.
The following information is the best estimate possible within the time and using available data.
If no information is provided to the Dose Calculator, default values will be used to determine dose projections. This may result in more severe conditions prompting a non-conservative protective action recommendation.
1.0 ISOTOPIC DETERMINATION: (choose one)
UNKNOWN MIX (Containment Rad <SR hr)
NORMAL COOLANT LEAK (Containment Rad <5R/hr)
LOCA No Fuel Damage (Iodine Spike, Containment Rad <10R/hr)
LOCA CLAD FAILURE (Containment Rad 1.5E+02 - 5.0E+04 R/hr)
LOCA FUEL MELT (Containment Rad 8.OE+03 - 1.OE+06 R/hr)
FUEL HANDLING ACCIDENT EP-AD-000-454, Revision 3, Page 1 of 2
Tab 1 EP-PS-1 36-1 2.0 CORE CONDITION: (choose one)
Gap Release (Core uncovered for 15-30 minutes)
In Vessel Severe Damage (Core Uncovered >30 Minutes)
Vessel Melt Through EP-AD-000-454, Revision 3, Page 2 of 2
( ( (
Tab 3 EP-PS-1 36-3 DOSE PROJECTION WORKSHEET GENERAL:
After Initial activation, the following information should be kept current at all times. This information is used by the Health Physics Dose Calculator to perform dose projections in support of Protective Action Recommendations required to be made within 15 minutes of a General Emergency classification.
The Engineering Support function will provide the following information to the Dose Calculator in a timely manner to allow sufficient time to perform a dose projection.
The following information is the best estimate possible considering the time and information available. If the information is not provided, the Dose Calculator will use default values which may yield dose projections much more severe than actual conditions prompting a non-conservative Protective Action Recommendation.
1.0 GENERAL INFORMATION RX SHUTDOWN TIME: J RELEASE START TIME: j RELEASE STOP TIME: j 2.0 TYPE OF RELEASE (select one) i MONITORED &
MONITORED: UNMONITORED: UNMONITORED:
EP-AD-000-459, Revision 2, Page 1 of 4
C C ~~~~~~~~~~~~~~~~~~~~~~~~~~(
Tab 3 EP-PS-1 36-3 3.0 DBA ACCIDENT TYPES [ISOTOPIC DETERMINATION (CHOOSE ONE)]
UNKNOWN MIX (Containment Rad <5 R/hr)
NORMAL COOLANT LEAK (Containment Red <5 Rehr)
LOCA (No Fuel Damage, iodine Spike, Containment Rad <10 R/hr)
LOCA CLAD FAILURE (Containment Red 1.5E+02 - 5.0E+04 R/hr) __% (Give Estimate)
LOCA FUEL MELT (Containment Red 8.0E+03 -1 .OE+06 R/hr) =_ % (Give Estimate)
FUEL HANDLING ACCIDENT NOTE: Quick methods to determine the isotopic mix/type .of fuel damage and estimate percentages are located in HELP tabs entitled Core Damage Estimate I (Primary System Breach Inside Containment) and Core Damage Estimate 11(Small or no Primary System Breach Inside Containment).
EP-AD-000-459, Revision 2, Page 2 of 4
( (C (
Tab 3 EP-PS-1 36-3 4.0 RELEASE TYPES (Select release type and go to release selected) 4.1 Drvwell Release (circle one In each column)
Core Condition (Choose one, provide estimate In '%) Sprays Hold-up Time Treatment Release Rate Gap Release (Core uncovered for 15-30 minutes) On <1 Hour Filtered 100%/o/hr In Vessel Severe Damage (Core uncovered > 30 minutes) Off 2-12 Hours Unfiltered 100%/o/day Vessel Melt Through 24 Hours Design (10/./day) 4.2 Wetwell Release (circle one In each column)
Core Condition Water Hold-up Time Release Rate (Choose one, provide estimate In I%) Conditions Treatment Gap Release (Core uncovered for 15-30 minutes) Subcooled <1 Hour Filtered 100%/o/hr In Vessel Severe Damage (Core uncovered > 30 minutes) Saturated 2-12 Hours Unfiltered 100 0/o/day Vessel Melt Through 24 Hours Design (1%/./day)
EP-AD-000-459, Revision 2, Page 3 of 4
( ( (
Tab 3 EP-PS-1 36-3 4.3 Secondary Containment Bypass Release (circle one In each column)
Core Condition (Choose one, provide estimate in %)Treatment Release Rate Gap Release (Core uncovered for 15-30 minutes) Filtered 100%/o/hr In Vessel Severe Damage (Core uncovered > 30 minutes) Unfiltered 100%/o/day Vessel Melt Through Design (1%/o/day) 4.4 Spent Fuel Pool Release (circle one In each column)
Core Condition Hold-up Release (Choose one) Accident Type Time Treatment Rate Gap Release Zircaloy Fire in One 3 Month Batch <1 Hr Filtered/ 100%/o/hr (Core uncovered for 15-30 minutes) Sprays on In Vessel Severe Damage Gap Release from One 3 Month Batch 2-12 Hrs. Unfiltered/ 100%/o/day (Core uncovered > 30 minutes) Sprays off Vessel Melt Through Gap Release from 15 Batches EP-AD-000-459, Revision 2, Page 4 of 4
Tab 4 EP-PS-136-4 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 backgroundradiation 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 andlor 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 4 EP-PS-1 36-4 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 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 4 EP-PS-1 364 Containment Radiation Monitor Response Direct Release Path to Drv well I.E+07 (Sprays Off)
I.E+06 100%
-50% .
100%
1.E+05 10% 50% 100% r t5% ==-50%- 100%
10%
I.E+04 1t% 5% _10% -_50%
ffi [.~~~1% _.--_ -10%_
'= 1.Ee03 :1% _ 5%__
C I.E+03 E
C
° 1.E+02 0)50 o 1.E+OI1s 1%1e%
It 1.E+O _ _0 CD I~1.E-002 0 L 00s%_0_110 50%~~~~~~~~~~~~~0 I.E-03 0 ~ ~ ~~~~~~~~~~~~~~~~~0 110% 1%%
I.E-03 %20 I.E-04 1%.
I.E-05 In 24h In 2h 1 4 h 24i 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-000-457, Revision 7, Page 3 of 10
Tab 4 EP-PS-1 364 Containment Radiation Monitor Response Direct Release Path to Dir well (Sprays On)
I.E+07 I.E+06 _
1.E+05 50%
-10% 10%
100%
I .E+04 _% _ 50%
1% 10% ~50%
i2 --- ~~~~~10% 100
- 1.E+03 c=%50%
E
- C 1.1% 1
-1%-%10%-
° 1.E+02 o -~~~~~~~~~1%
'a 1.E+O0 0* 1.E+O Q
° 1.E-O1-- 0~~~~~~~~~~~~~0
- 0~~~~~~~~~~~~~~~10 1.E-02 100%
-50%-
1t% _ -10% _ _1%
1.E-03 10% _-
10%
1I.E-04 ==_ -s f s I.E-05 1h 24h 1h 24h 1h 24h 1h 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 3to determine If fuel melt occurred (core uncovered or fuel blockage).
ATTACHMENT 1B EP-AD-000-457, Revision 7, Page 4 of 10
Tab 4 EP-PS-1 364 Containment Radiation Monitor Response Direct Release to Wetwell and Not to Drvwell 1.E.06
.100%
1.E+05 50%
-10%
- 100%
1.E+04 -:5% -=50%
100%_
50%
1%_ = 10% _ -100%
1 .EI03 - - 5%-_ 10% _=50%
1 0- _,_5%
i2 1 .E+02 1% 10%_
S 1.E+01 E
0 .+0
'E 1I.E+01-- ° -T°°
- t -~~~~~~~~~50%'
_50%-
U~~~~~~~~~~~~~~~~~~~~~~~
°T 1.E _0-%_10%.
1I.E-03 -_1% = 1%, 100% _.10
_-5% ' 50% - _
I.E-05 _ -% _ 1% *M_
1.E-06 1h 24h 1h 24h 1h 24h 1h 24h In-Vessel Melt Gap Spiked Coolant Nonnal 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 IC EP-AD-000-457, Revision 7, Page 5 of 10
Tab 4 EP-PS-136-4 CONTAINMENT HYDROGEN VS CORE DAMAGE
% MdaIWats IacIon&COM~re amgeSt 40 80 '4. ponlfmeMum 20 <.hSSIft Unoolabl ONrhe, IStud~
10 S= M*hu 0
0.1 1 10 H2 % InContainment
-*GNM k &I1 Saw=, NUREG=-M&26 p.4-3; daxge mal NREUG4524. VaL S.;
ThUpe~ ge IURBG-1370- NUREGICR 04OI: NUREW M-5S6. Tabl 4.9. p.71.
cai~ ~ TYab = C ATTACHMENT 2 EP-AD-000-457, Revision 7, Page 6 of 10
Tab 4 EP-PS-136-4 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 0F.
ATTACHMENT 3 (Page 1 of 4)
EP-AD-000-457, Revision 7, Page 7 of 10
Tab 4 EP-PS-1 36-4 WATER INJECTION REQUIRED TO COOL CORE BY BOIUNG aLile the top of the active core Ls uncovered, assuie that the fuel viii heat uat t -2*P/sc.
fuel pi damage as shown below.
The increased a tperature vill result In NOTE:~~~~~~~~~is "MsA eztimtes are reasonable (factor of .n d WvW6OJ)
- 2) If the core is unovered v a zitin - W tey hours of shutdown (Inld faIxIe to gA d as scram) . f there Is suffacIqat nj.ectIon core heatup may be stopped or slowed due Pa VMtbn d can to Aste coalng.
steam coolng may not Oarr pre-Vent. core dmg under &ccldent Foowd d "
Ar~~~~~~~Vts. hod mc_~w VWYlipid ssun - SWcfoln -
dditg a bum ed oan hJI pinmih gw goue woo-09OO. O/-434, UMG0-CsS 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 7, Page 8 of 10
Tab 4 EP-PS-136-4 WATER INJECTION REQUIRED TO COOL CORE BY BOILING nWETIM (mm) To aZPTAC WATER LT
.5r BOL3NG MO DECAY BEAT FM a sooo VK(t)
PLAN C1/2-24 )UM IJZ SBUTDOW) e00 t
- CI j ISM sso fig I 3 2 4 o SO0. , A, . Shtew nmowN (gpa) RZQunD To RIMACE WALR OST By BOILING DE TO DECAY MA FM a 3000 KW(i)
PIS= (1 to 30 DAYS A?= SHUTDONN) ontsa Cc enJectec SO~~~~~~~~~~~~~~~~~~~~~~~IC 20 . ' C 3
- 2 8 4 1 I0 *7
- tO0 0 2*
0 Aft Utwn ATTACHMENT 3 (Page 3 of 4)
EP-AD-000-457, Revision 7, Page 9 of 10
Tab 4 EP-PS-136-4 WATER INJECTION REQUIRED TO COOL CORE BY BOILING Core dnge vs. thae that reactor care Is mcovered Tiin PWR or 205 of C~ CP=
BWR aczive cute is (ii) (OF) CC) Possibl core damag 0 >600 >315' None 0.5 so 0.75 1800-2400 M8-1300
- Loes fumel eiht
- SDomizi of claddin vfb stara
-roow (inwtkuinic ZER~O winod whh rapid E6 genration)
- Rapid Wd clddi failme (ga rielen from die core')
0.5 to 1.3 2400-4200 1300-230D
- Rapid rlem of volatile fission
- relyeas~e f eco
- Possible rlocsatim (ubup) of molten cme
- Possbibl urecolble cane Iso3+ >4200 >2300 bbhdehmxgh of vessel with possible cmmaim fhatlr ead release of additinllssvltl fissoti Aommst NUREGfcR-424S. NUREGIMR4624. NUJEGICR4629. NUREGfCR-S374..NUREO4900.
NUREG196. NiUREG I5. and NURBG-1465.
ATTACHMENT 3 (Page 4 of 4)
EP-AD-000-457, Revision 7, Page 10 of 10
Tab 5 EP-PS-1 36-5 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 significantry 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 RESPONSIBILIES 2.1 The Nuclear Fuels Engineer. Lead Technical Support Enaineer, 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 2200 0F, but the Containment Post-Accident Radiation Monitor readings have increased.
3.1.2 Fuel Overheat is expected If peak cladding temperature exceeds 2200 0F, but the maximum volume-averaged fuel pellet temperature remains less that 45000 F.
3.1.3 Fuel Melt is expected If any volume-averaged fuel pellet temperature exceeds 45000F.
EP-AD-000-270, Revision 6, Page 1 of 3
Tab 5 EP-PS-1 36-5 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 Attachment 3 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:
FFF= CPARM Reading Expected 100% Fuel FailureCPARM Reading EP-AD-000-270, Revision 6, Page 2 of 3
Tab 5 EP-PS-1 36-5 FIGURE '1 CONTAINMENT HIGH RANGE RADIAIBON MONITOR READINGS THAT ARE EXPEClD WITH 100% OF THE FUEL PALED FOR AN EVENT WITH NO PRIMARY SYSTEM BREACH INSIDE OONTAINMENT
~~~~~~.......... ............. . ... . . ....... . . ....... ....
. ........ 1~~1 !.. ..... . . .. . .... ... .., .. .----........--'-
...... . ......... _., . ............. .O-V I .. : NON-LOCA
. .. r... . .........
FUELMELT.
- 1. ..... . ..
W 7..... .............
C~~ ~~ ~ ~ ~~~ ........... ..-- .. . ..... .......
~ 10 -..
............. 9.99.9
. . .r ....
. -. P. .
= ............................... ................... ....;;;
C. .......................... ......
_~~~~~~~~~~~~~~~~~~~~~ ------;7 -----
EL OVERHEAT -.
U-o._. ~~~~~....... ....... I .. ,......
j.,.,..
.. 9.......
.9.D..........~... * ............
U----------- --------- ....... ...... ..... ............ ..... ............... _.:
10 5ts . zo Post-Shutdown Time (HRS)
EP-AD-000-270, Revision 6, Page 3 of 3