ML051080486
| ML051080486 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 03/17/2005 |
| From: | Nuclear Management Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| L-MT-05-013, TAC MC3299 CA-05-072, Rev 13 | |
| Download: ML051080486 (30) | |
Text
ENCLOSURE 4 MONTICELLO NUCLEAR GENERATING PLANT CALCULATION CA-05-072 Effect of Reduced Pool Water Levels on Fuel Handling Accident Consequences This enclosure consists of a copy of the Monticello Nuclear Generating Plant Calculation CA-05-072, uEffect of Reduced Pool Water Levels on Fuel Handling Accident Consequences." This calculation is current as of the date of this letter. Future revisions to this calculation, if any, will be available onsite for staff review.
MONTICELLO NUCLEAR GENERATING PLANT 3494 TITLE: CALCULATION COVER SHEET Revision 13
. TlTPage 1 of 5 Page 1 of CALCULATION COVER SHEET Title EffectofReducedPoolWaterLevelson CA- 05 - 072 Add. 0 Fuel Handling Accident Consequences PART A - (Not Applicable to Vendor Calcs)
Assigned Personnel Name (Print) Signature Title Initials Record of Issues Total Last Approval Rev Description Sheets Sheet Preparer Verifier Approval Date Verification Method(s) 2 Review E Alternate Calculation D Test LI Other EI Technical Review (per 4 AWI-05.08.07 (FP-E-MOD-07))
3087 (DOCUMENT CHANGE, HOLD AND COMMENT FORM) incorporated: : .
I:F.OR:ADIUItSTRATIVESAiN ReSD SUDV: CNSTP I Assoc Ref 4 AWI-05.01.25 l SR: N l Frec: 0 vrs USE ONLY ARMS: 3494 Doc Tvoe: 3042 Admin initials- Date:
Approved (Signatures available in Master File)
MONTICELLO NUCLEAR GENERATING PLANT 3494 TITLE: CALCULATION COVER SHEET Revision 13
. Page 2 of 5 Page 2 of CA 072 PART B - (Applicable to Vendor Calculations Only)
Sargent &
Vendor Name Lundy Vendor Calc No: 2004-09840 Vendor Approval Date: 03/07/2005 En Form 3345 or QF-0547 attached.
l Safety related? If checked, attach DIA or reference here. SLMON-2005-017 I Melissa LimbeckI Reviewed by: Kathy Shriver ly A.8A X & /'442 Print Name -i" iture Date Accepted by: Al Williams _ 3/I Print Name Eng. Supv. Signature Date Record of Issues Revision Description Total No. of Sheets Last Sheet Number 0 Initial Calculation 29 23 of 23 PART C - Design Basis Tracking Data (Complete for all Calculations) 10 CFR50.59 Screening or Evaluation No: N/A Associated Reference(s): ASTPhase 1 (FHA) LAR Does this calculation:
Calc. No(s)
Yes L No l: Rev.s &
Supercede another calculation? Add No(s):
Calc. No(s)04-041, Rev Augment (credited by) another Yes No L:Revs & Add 1;05-081, calculation? No(s): Rev 0 Caic. No(s)
Derive inputs from another Yes i No l: Revs & Add calculation? INo(s):
Affect the Fire Protection Program Yes (Form 3765)? (attach L No Fonn 3765)
Approved (Signatures available in Master File)
MONTICELLO NUCLEAR GENERATING PLANT 3494 TITLE:l CALCULATION COVER SHEET Revision 13 IPage 3 of 5 Approved (Signatures available in Master File)
MONTICELLO NUCLEAR GENERATING PLANT 3494 TITLE: CALCULATION-COVER SHEET Revision 13 A ~Page 4 of 5 Page 3 of CA 072 Does this calculation (Cont'd):
Affect IST. Program Valve or Pump Yes LI No 1 Reference Values, and/or Acceptance Criteria?
If yes, inform the ]ST Coordinator and provide with copy of revised calculation.
List all documents/procedures that are based on this calculation:
USAR, Rev 21 (Section 14.7.6); AST Documents/ProcedureS (include revision): Phase 1 (FHA) LAR
- List all plant procedures used to ensure inputs/assumptions/outputs are maintained:
Procedures (include revision): USAR, Rev 21 (Section 14.7.6)
- What Systems or components are affected?
System Code(s): (See Form 3805 (DESIGN BASIS INFORMATION SYSTEM (DBIS)
CODES FOR SYSTEMS, STRUCTURES AND TOPICS) for available codes)) FPC, RCH Component ID's (CHAMPS Equip) N/A DBD Section (if any): B.02.01: Table 1.1; Sections 1, 2, 4 Topic Code: (See Form 3805 for available codes) DBAE Structure Code: (See Form 3805 for available codes) N/A Other Comments: See next page.
Approved (Signatures available in Master File)
MONTICELLO NUCLEAR GENERATING PLANT 3494 TITLE: CALCULATION COVER SHEET Revision 13 Page 5 of 5 Future needs (reference EWR024927): Update USAR and DBD relevant sections. Determine procedural controls to ensure consideration and update of this calculation if fuel design or refuel bridge and mast configuration change.
This calculation was completed in accordance with the approved project work plan (DIA Equivalent) S&L Letter# SLMON-2005-017, and P502378.
- This calculation provides the basis for amendment to TS 3.1 O.C, Spent Fuel Pool Water Level, as part of the AST Phase I (FHA) LAR.
This calculation also provides the basis for pool water level specifications for the ITS submittal. Procedures referencing required spent fuel pool water level will be based on the TS itself rather than this calculation, and will be revised as part of the TS or ITS implementation process. These procedures include: 0000-J; 9007; 9007-B; 9009; D.2-02; D.2-05.
Approved (Signatures available in Master File)
MONTICELLO NUCLEAR GENERATING PLANT 3345 TITLE: VENDOR CALCULATION REVIEW CHECKLIST Revision 3 Page 1 of 1 CA- 05 - 072 The purpose of the review is to ensure that the vendor calculation or analysis complies with the conditions of the purchase order and is appropriate for its intended use. The purpose of the review is not to serve as an independent verification. Independent verification of the calculation or analysis by the vendor should be evident in the document.
The reviewer should use the criteria below as a guide to assess the overall quality, completeness and usefulness of the calculation or analysis. The reviewer is not required to check the vendor's calculations in detail.
See 4 AWI-05.01.25 (CALCULATION/ANALYSIS CONTROL) for guidance. Place initials by items reviewed.
REVIEW
- 1. Form 3544 (PIPING AND SUPPORT NUMBERING) completed for calculations affecting piping or supports. N/A
- 2. Design inputs correspond to those which were transmitted to the vendor. kw
- 3. Assumptions are described and reasonable. Basis for assumptions identified. /au
- 4. Applicable codes, standards and regulations are identified and met: _ _
- 5. Applicable construction and operating experience is considered.
- 6. Applicable structure(s), system(s),.and component(s) are listed.
- 7. Formulas and equations documented and unusual symbols are defined.
- 8. Acceptance criteria are identified, adequate and satisfied. A
- 9. Results are reasonable compared to inputs.
- 10. Source documents are referenced.
- 11. The calculation is appropriate for its intended use. /iArA,
- 12. The calculation complies with the terms of the Purchase Order.
- 13. Inputs, assumptions, outputs, etc. which could affect plant operation are enforced by adequate procedural controls. List any affected procedures.
Reference FWR024927 Futuire Needs for CA-05-072.
1Ž7L2-Completed By: I /e-/' ii'{-IK' Date: 3- I- ZDoS 3087 (DOCUMENT CHANGE, HOLD AND COMMENT FORM) incorporated:
Resp Supv: CN5TP l Assoc Ref: 4 AWI-05.01.25 l SR: N Freg: 0 yrs FORADMINISTRATI I:lARMS: 3345 l Doc TvDe. 3042 l Admin initials: l Date:
Approved (Signatures available in Master File)
41 11 ISSUE
SUMMARY
Form SOP-0402-07. Revision 6 DESIGN CONTROL
SUMMARY
CLIENT: Nuclear Management Company UNIT NO.: 1 Page No.: 1 PROJECT NAME: Monticello PROJECT NO.: 11163-063 E NUCLEAR SAFETY- RELATED CALC. NO.: 2004-09840 1 NOT NUCLEAR SAFETY-RELATED TITLE: Effect of Reduced Pool Water Levels on Fuel Handling Accident Consequences EQUIPMENT NO.:
IDENTIFICATION OF PAGES ADDED/REVISEDISUPERSEDEDNOIDED & REVIEW METHOD Original Calculation (23 pages)
INPUTS/ ASSUMPTIONS
.0 VERIFIED o UNVERIFIED REVIEW METHOD: Detailed .REV. 0 STATUS: Approved -DATE FOR REV.: PREPARER W. J.Johnson 'DATE: - 7fl9
- REVIEWER B. J. Andrews C b DATE. 3_ 7- '.ove' DAl.
APPROVER S. R. Raup2 ODATE- 7A i,';
IDENTIFICATION OF PAGES ADDED/REVISED/SUPERSEDEDNOIDED & REVIEW METHOD INPUTS/ ASSUMPTIONS 0 VERIFIED o UNVERIFIED REVIEW METHOD: REV.
STATUS: - DATE FOR REV.:
PREPARER DATE:
REVIEWER DATE:
APPROVER fDATE.
IDENTIFICATION OF PAGES ADDED/REVISEDISUPERSEDEDNOIDED & REVIEW METHOD INPUTS/ ASSUMPTIONS o VERIFIED D UNVERIFIED REVIEW METHOD: REV.
STATUS: DATE FOR REV.:
PREPARER DATE.
REVIEWER DATE.
APPROVER DATE.
NOTE: PRINT AND SIGN IN THE SIGNATURE AREAS SOP040207.DOC Page 1 of 1 Rev. Date: 08-31-2004
1%,
Calcs. For -Effect of1Reduced Pool Water Levels on Fuel I Caic No. 2004-09840 n J Handling Accident Consequences ;Rev. 0 Date X ] Safety Related Non-Safety Related Page 2 of ,3 l
Client Nuclear Management Company Project Monticello IProj. No 11163-063 Equip. No. y TABLE OF CONTENTS 1 PURPOSE AND SCOPE ....................................... 3 2 DESIGN INPUT ............ ; ; . 4 3 ASSUMPTIONS ........... 6 4 METHODOLOGY AND ACCEPTANCE CRITERIA ....................................... ; . 7 5 CALCULATIONS..............................12 RE S . . ......
.. ........... 20.
6 RESULTS S ........... 2 7 REFERENCES ...... 21 Form'GO-3.i8.1 Rev.2
,file G0308-ZDOC Fiie G0306-2.tG0C Formr70-3:08.1 Rev.2
Calcs. For Effect of Reduced Pool Water Levels on Fuel I ]Calc No. 2004-09840- ]
3t . 1
. Lr e j Handling Accident Consequences Rev. 0 IDate X Safety Related Non-Safety Related Page 3 of. 9.3 Client Nuclear ManagementCompany Prepared by Date Project Monticello Reviewed by Date ProJ. No 11163-063 Equip. No. Approved by Date I PURPOSE AND SCOPE The purpose of this calculation is to determine the effect of reduced water level in the Refueling Pool (RP) and Spent Fuel Storage Pool (SFSP) on radiological consequences following a Fuel Handling Accident (FHA). The depth of water affects the radiological consequences because credit is taken for removal by the pool water of airborne iodine nuclides, thereby reducing the amount of activity released from the plant and the resulting radiation doses. As described in the Section 14.7.6.1 of the MNGP USAR [Reference 7.5], the limiting FHA is one resulting from the accidental dropping of a fuel assembly into the reactor vessel onto the top of the core. The objective of this calculation is to demonstrate that the current design basis (limiting) FHA is bounding for an FHA that involves the drop of an assembly in the RP (such as on the Reactor Pressure Vessel (RPV) flange) or in the SFSP.
The scope of this calculation involves three separate analyses:
- Effect of Water Level on Iodine Removal. A generic analysis is performed to demonstrate the effect of reduction of pool water level on the amount of iodine activity released from the pool.
. Evaluation of the Drop of an Assembly in the RP. The drop of an assembly in the RP (rather than in the reactor vessel) is .evaluated to demonstrate that the consequences are bounded by the current design basis FHA.
. Evaluation of the Drop of an Assembly in the SFSP. The current design basis FHA is a drop of an assembly into the reactor vessel. A drop of an assembly in the SFSP is evaluated to demonstrate that it is bounded by the current design basis FRA.
The current design basis analysis for the FHA uses the assumptions outlined in Regulatory Guide 1.25 (RG 1.25) [Reference 7.1], which includes assumptions concerning removal of iodine by the pool water.
An alternative method of analysis for the FHA is outlined in Regulatory Guide 1.183 (RG 1.183)
[Reference 7.2], which includes different assumptions concerning removal of iodine by the pool water.
Calculation 2004-02104 [Reference 7.7] documents the FHA analysis performed for MNGP using RG 1.183 (AST) methodology. This analysis addresses both sets of assumptions.
'File G030B-2.DOC Form GO-3.08.1 Rev.2
1 Caics. For Effect of Reduced Pool Water Levels on Fuel Calc No. 2004-09840 Mr 0 Handling Accident Consequences Rev. 13fDate X I Safety Related Non-Safety Related Page 4 of as Client Nuclear Management Company Prepared by Date Project Monticello Reviewed by Date Proj. No 11163-063 Equip. No. Approved by Date 2 DESIGN INPUT 2.1 The minimum depth of water in the refueling pool in the fuel movement path is the depth to the cattle chute floor. With a refueling water height at elevation 1026'-2", this depth is 21'-8" and the minimum water depth above the RPV flange is 22'-2". [Reference 7.3, Itemn 1]
2.2 With a refueling water height at elevation 1026'-2", the minimum depth of water in the reactor pressure vessel above the core during fuel movement is 45'-10". {Reference 7.3, Item 2]
2.3 With a refueling water height at elevation 1026'-2", the minimum depth of water in the SFSP above spent fuel during fuel movement is 21'-9". The minimum depth of water above damaged fuel (dropped bundle) is 21'-4". [Reference 7.3, Item 11] - .. ;. ....
2.4 Maximum height of fuel assembly above the RPV flange during fuel movement is 3'-2".
[Reference 7.3, Item 3]
2.5 Maximum height of a fuel assembly above core during fuel movement is 26'-10". {Reference 7.3, Item 4]
2.6 Maximum height of a fuel assembly in the SFSP above the spent fuel during fuel movement is 2'-10". [Reference 7.3i Item 12]
2.7 The inorganic iodine species fraction for RG 1.25 analysis is 99.75%, and the inorganic iodine species fraction for RG 1.183 analysis is 99.85%. [Reference 7.3, Item 5]
2.8 The pool decontamination factor for inorganic iodine for a pool depth of 23' is 133 for the RG 1.25 analysis and 500 for the RG 1.183 analysis. [Reference 7.3, Item 6]
2.9 The decontamination factor for organic iodine is 1.0 for both the RG 1.25 and RG 1.183 analysis.
[Reference 7.3, Item 7]
2.10 The number of fuel rods in the fuel assembly considered in the design basis FHA analysis is .60, based on an 8x8 fuel assembly. [Reference 7.3, Item 8]
2.11 The length of a GE14 BNVR/3 fuel assembly is approximately 14.3'. This is the difference between the height of the fuel assembly above the bottom of the SFSP and the height of the bottom of the fuel racks: 185.24"-8.99"-5.19" = 171.06" _ 14.3'{Reference 7.3, Item 11]
2.12 The number of failed fuel rods assumed in the design basis FHA is 125, based on an Sx8 fuel assembly. [Reference 7.3, Item 10]
Form GO-3.08.1 Rev.2 FiIe00308-2.DOC
'File G0308-2.DOC form 130-3.08.1 Rev.2
Caics. For Effect of Reduced Pool Water Levels on Fuel Calc No. 2004-09840 Handling Accident Consequences .Rev. 0 ]Date X I Safety Related Non-Safety Related Page 5 *of Z3 Client Nuclear ManagementCompany ] Prepared by Date Project Monticello 4Reviewed by Date Proj. No 11163-063 Equip. No. Approved by Date 2.13 The weight of the GE14 BWR/3 fuel assembly dropped in the FHA is 553 lbs. {Reference 7.3, Item 13]
2.14 The weight of the NF-400 refueling mast that is dropped with the fuel assembly is 350 lbs.
[Reference 7.3, Item 14]
2.15. The fraction of GE14 BWRI3 fuel assembly weight~not including fuel) that is associated with the fuel cladding is 0.525. [Reference 7.3, Item 15]
2.16 Current MNGP core and refuel loads include GEII BWR/3 and GE14 BWR/3 fuel.
[Reference 7.3, Item 13]
2.17 The equiivalent number of full length fuel rods in7aE14 BWRJ3fuel assembly is 7.33. -
[Reference 7.8, pg 90]
Form GQ.3.08.1 Rev.2 File G0308-2.DOC
'Fite G0308-2.DOC Forrn GQ-3.08.1 Flev.2
Caics. For Effect of Reduced Pool Water Levels on Fuel I IC-31c No. 2004-ID9840 --
I I wyl ... IHandling AccidentConsequences I]1Rev 0 ~1Date X I Safety Related I I Non-Safety Related 113age 6 Of a3 lClient Nuclear Management Company I IPrepared by
]Project I.
Monticello 1 1Reviewed by IProj. No 11163-063 Equip. No. I IApproved by 3 ASSUMPTIONS 3.1 All iodine that is not inorganic is assumed to be organic. The-organic species fraction specified in RG 1.25 is 0.25% and the organic species fraction specified in RG 1.183 is 0.15%.
3.2 As specified in RG 1.25 and RG 1.183, the retention of noble gas activity in the pool water is negligible, i.e., the decontamination factor for noble gas activity is 1.
3.3 The method for deterhining the number of failed fuel rods resulting from the drop of a fuel assembly is based on the assumptions described in MNGP USAR Section 14.7.6 Reference 7.5]
and GESTAR II [Reference 7.6]. These assumptions are summarized in Section 4.2 of this calculation. - . -.- - - .. -:
Frm 0 -3.8.1 ev.
Fi~e Q30-2.O C FileG0308-2DOC Form GO-3.OB.1 Rev.2
Calcs. For Effect of Reduced Pool Water Levels on Fuel I lCalc No. 2004-09840.
I
-yangandlingAccidentConsequences H.- , Rev. 0 ]Date X I Safety Related l lNon.Safety Related Page 7 of ;1?
Client Nuclear Management Company Prepared by Date Project Monticello Reviewed by Date Proj. No 11163-063 Equip. No. Approved by Date 4 METHODOLOGY AND ACCEPTANCE CRITERIA 4.1 Effect of Pool Water Depth on Pool Decontamination Factor Both RG 1.25 and RG 1.183 indicate that for pool water depths less than 23' the decontamination factor will have to be determined on a case-by-case method. The pool water depth is defined as the depth of water above the damaged fuel. RG 1.183 cites as a reference for the methodology Reference 7.4.
Inspection of this reference indicates it is the basis for the guidance provided in RG 1.25. Therefore, the methodology used in this calculation to determine the decontamination factor for pool water depths less than 23' is based on Reference 7.4.
TI'e decontaminiation factor (DF) is defined as the ratio of the initial to final-concentrations of species of:
interest in a bubble of gas that passes through the pool of water. The DF differs for the inorganic and organic species of iodine and must be evaluated separately for each one. For organic iodine the DF is assumed to be 1, i.e., there is no absorption in the pool water. For inorganic species, the DF is defined by the following expression.
(6 H' DFInr =exp ~keff J()
-rs db Vb.
The parameters in this equation are defined as follows:
db = bubble diameter keyf = mass transfer coefficient vb = bubble velocity H = bubble rise height, or the effective depth of the water in the pool The bubble diameter is affected by the size of the orifice through which the leak occurs, which will not change as a function of pool depth. Similarly, since the bubble velocity is directly related to bubble volume it is also not expected to be sensitive to small changes in pool depth. The mass transfer coefficient is determined largely by the characteristics of the water, which will also not change with pool depth. Therefore, these three parameters can be held constant, and the equation for the DF can be rewritten as follows.
DFInor = exp(CH) (2)
In this expression, C is defined as follows.
Form GO-3.08.1 Rev.2 File G0308-2.DOC File G0308-2.130C Form GO-3.08.1 Rev.2
16 Calms.lFor Effect of'Reduced Pool Water Levelson Fuel Calc No. 2004-09840
. .....J Handling Accident Consequences
. Rev. 0 IDate X Safety Related l Non-Safety Related Page 8 of A?
Client Nuclear Management Company Prepared by Date Project Monticello Reviewed by Date Proj. No 11163-063 Equip. No. Approved by Date C =-kff (3) db Vb If the DF is known for some specific pool water depth, Ho, the DF for any other water'depth 1H, can be determined by the following expression.
PDF.Orgo exp(CHO)
To remove the constant (C) from this equation, the natural logarithm of both sides is taken, resulting in A.he following equation. '. .' .;;- - - - . . -
If(~n Ij)nDF~~0 = CH C'H, -
7-Ho H (5) 0 From equation (2) above, the following expression for C results from taking the -natural logarithm of both sides of the equation.
ln(bFinor,.O)
C = )(6)
Ho Substituting this definition for C and rearranging results in the following expression.
ln(DF,nor.) = H 0 n(DFnor.O (7)
Using the relationship a ln(x) = ln(xa ), this can be rewritten as follows by exponentiating both sides of the equation.
DFnorg., =DFnorg.c)7 (8)
The overall DF for the pool is a function of both the DF for the iodine species and the species fraction.
Since the DF for organic iodine is ], the pool DF is given by the following expression.
D F fraction inorganic fraction organic 19)
DFInorg Form .GQ-3.08.1 Rev.2 FiIeGQ30B-2:DOC File GQ308-2:DOC Form-GQ-3.08.1 Rev.2
Calcs. For -Effect of Reduced Pool Water Levels on Fuel ICalc No. 2004-09840 L.LGJ L9 Handling Accident Consequences Rev. 0 Date X SafetyRelated l Non-Safety Related .Page 9 of Q3
.Client Nuclear Management Company ;Prepared by Date Project Monticello Reviewed by Date
- Proj. No 11163-063 Equip. No. Approved by Date It should be noted that the use of the species fractions and DFs from RG 1.25 results in an effective DF of 100, as stated in the regulatory guide. However, using the species fractions and DFs from RG 1.183 results in an effective DF of 286, which is Jarger than the DF of 200 specified in the regulatory guide.
This'discrepancy is a deliberate effort by the NRC to reduce the uncertainty in the pool DF. To maintain this margin, the DF calculated using theRG 1.183 species fractions and DF is normalized to a DF of 200 at a pool depth of 23'.
4.2 Effect of Pool Depth on Fuel Failure The amount of activity released following an FHA is characterized by two parameters. The first is the amount of activity in the void space of a fuel rod, which is commonly called the gap activity and is specified on a nuclide basis as afractionof the total core inventory. The second parameter is the number of failed fuel rods 'aused by he 'accident sinice it is*assumed thatall of the gp activity' ineach faild fuel rod is released. The gap activity is a characteristic of the operation of the reactor and is not affected by pool depth. The number of failed fuel rods, however, is affected by pool depth since the pool depth determines, in part, how far the fuel assembly will fall before the impact that -causes the fuel rods to fail.
The FHA is described in Section 14.7.6 of the Monticello USAR [Reference 7.5]. The method for determining the number of failed.rods is taken from GESTAR II [Reference 7.6]. GESTAR II contains a detailed analysis of the drop of a 9x9 fuel assembly (GEl I or GEI3). The following assumptions are made in the determination of the total number of failed fuel rods.
- 1. The fuel assembly is assumed to be dropped from a height of 34', which is based on a drop into the reactor vessel onto the top of the core from the highest point that the fuel assembly can be raised.
- 2. The refueling mast and grapple head are also assumed to drop and impact the dropped assembly and the assemblies in the core.
- 3. The entire amount of potential energy, including the .energy of the entire assemblage falling to its side from a vertical position, is available for application to the fuel assemblies involved in the accident.
- 4. None of the energy associated with the dropped fuel assembly is absorbed by the fuel material (uranium dioxide).
- 5. The dropped fuel assembly is assumed to impact at a small angle from the vertical, subjecting all.
the fuel rods in the dropped assembly to bending moments. The fuel rods are expected to absorb little energy prior to failing as a result of bending. For this reason, it is assumed that all of the rods in the dropped assembly fail. For the 9x9 assembly, this is a total of 74 fuel rods (seven of the fuel rods are displaced by water rods).
File-GO308-2.DOC Form4GO-3.08.1 Rev.2
P.
Caics. VFor Effect of Reduced Pool Water Levels on Fuel I JCaIc No. 2004-09840 1 I I Handling Accident Consequences lRev. 0 IDate LLX-ye' O I1 X I Safety Related . Non-Safety Related Page 10 of 3
- 1 Client Nuclear Management Company Prepared by Date
. Project Monticello Reviewed by Date Pro]. No 11163-063 Equip. No. Approved by Date
- 6. One half of the energy is considered to be absorbed by the falling assembly and one half by the four impacted assemblies.
- 7. The energy'available for clad deformation is considered to be proportional to the mass ratio. For the assembly analyzed, the mass ratio is equal to a maximum of 0.510.
S. Each rod that fails is expected to absorb approximately 200 ft-lb before cladding failure, based on uniform 1% plastic deformation of the cladding.
- 9. Based on the assumptions above and using a fuel assembly weight of 562 pounds and a grapple mast and head weight of 619 pounds, 51 fuel rods will fail in the impacted fuel assemblies.
- 10. The dropped assembly"is assumed 'to tip over aid'impact horizontally on the-top-bf the core from a height of one bundle length, approximately 160 inches. This second impact results in the failure of 15 additional fuel rods.
- 11. Based on items 5, 9 and 10 above, the total number of failed fuel rods for the drop of the 9x9 assembly is 74+51+15 = 140.
As this discussion indicates, the number of failed fuel rods will depend on the height of the drop because the total potential energy of the dropped assembly is the product of the weight and drop height. The assumption of the failure of all of the rods in the dropped fuel assembly is reasonable and conservative for a drop of any significant height. However, the number of fuel rods that fail in the impacted assemblies will be reduced if the drop height is decreased.
4.3 Effect of Pool Depth on Dose Consequences The general expression for the thyroid dose due to inhalation of iodine released during a FHA is given by the following expression from RG 1.25.
D=Fg I FPBR (y/Q)
D=Fsl~pB alQ)(10)
DFp DFf The variables in this expression are defined as follows.
D = thyroid dose (rads)
Fg = fraction of fuel rod iodine inventory in the fuel rod void space, or gap activity fraction I = core iodine inventory at time of accident (Ci)
F = fraction of core damaged so as to release void space iodine, or failed fuel fraction P = fuel peaking factor B = breathing rate (m 3 /sec)
Form-GO-3.08.1 Flev.2 File G0308-2.DOC File G0308-2.DOC Form-GO-3.08.1 Rev.2
r.
. Calcs. For Effect of Reduced Pool Water Levels on Fuel 016 CHandling Accident Consequences SfX Related ISafety lI Non-Safety Related 1Client Nuclear Management Company I FPrepared by Project Monticello IReviewed by Date I
JProj. No 11163-063 Equip. No. Approved by Date DFp = effective iodine decontamination factor for pool water DFf = effective iodine decontamination factor for filters by/? = atmospheric diffusion factor at receptor location (sec/r 3 )
R = adult thyroid dose conversion factor for the iodine isotope of interest (rad/Ci)
In this expression the only variable that is a function of pool water depth is DFp, which is the same as DFEff defined in equation (9). There is a potential that the pool water depth would also affect the effectiveness of the charcoal filters in the ventilation system, which is incorporated in DFf. This effect occurs if different filter efficiencies are assumed for the iodine species. Since the relative mix of inorganic and organic iodine in the air above the pool will change as the pool water depth changes, the amount of activity removed by the filter system would also change if the filter efficiencies for inorganic
+ andorgani ideadfrn.owev~rer, as shown in Table 14.7-22 of the USAR [Reference 7.5],-only andoranic iodine are different. H 'e,- e 1 erce.oy one filter efficiency is used for the exhaustffriter so the change in species distributiohn Vll fnbt 'affect iodine removal. For the AST analysis in Reference 7.7, it is assumed that there is no removal of iodine by either the exhaust filters or the control room intake filters. Therefore, DFf would default to 1 and not be affected by pool water depth.
The only variable affected by the amount of fuel damage that occurs during the drop is F. Therefore, an expression for the dose that results from a drop occurring in a pool with an effective pool depth of HI is as follows.
Di = FDF.I FFF,1D (I 1)
In this expression, D is the thyroid dose as calculated using equation (10) (assumed to be with a 23' pool water depth), and the other variables are defined as follows.
DI = thyroid dose(rads) for a pool depth of HI FDF.1 = DF adjustment factor =-wee Ff
=sDF__ where DFEf I is determined using equation (9) for a pool water depth of HI and DF, is the effective DF for a pool water depth of 23' FFF.I = fuel failure adjustment factor F, F
F, = fraction of core damage for a drop in a pool with effective height HI As this expression indicates,'the thyroid dose will increase with decreasing pool water depth because the DF will become smaller as the pool water depth becomes smaller. Also, as expected, the thyroid dose will be directly proportional to any change in failed fuel fraction. For analyses based on RG 1.1 83, the results are reported in total effective dose-equivalent (TEDE) rather than thyroid dose. The TEDE includes contributions from both noble gases and iodines. Since the pool water depth has no effect on the Form GQ-3.08.1 Rev2 FileGO3O8-2D0C FileaiQ38-2.1D0C Form Gi-3.08.1 Rev.2
e.
-Calcs. For Effect of Reduced Pool Water Levels on Fuel
,Ln" f Handling Accident Consequences X Safety Related Non-Safety Related Client Nuclear ManagementCompany Prepared by 1Date Project Monticello Reviewed by Date Prol. No 11163-063 Equip. No. Approved by Date noble gas activity released, the decrease in TEDE will be even larger than the decrease in thyroid dose because the noble gas contribution to the TEDE is only affected by the fuel failure adjustment factor.
4.4 Acceptance Criteria There are no acceptance criteria for the decontamination factor as a function of pool water depth.
For the evaluation of the drop of an assembly in the RP or SFSP, the acceptance criterion is that the control room and offsite doses are bounded by the current design basis FHA; 5 CALCULATIONS 5.1 Decontamination Factor Vs. Pool Water Depth The calculated inorganic and effective DF for various pool water depths are listed in Table 1. The inorganic DF is calculated using equation (8). For example, consider a pool water depth (HI) of 20'. For the RG 1.25 assumptions, the inorganic DF is given by 20 DFInorS20fi = (133)23 = 70.279.
The effective DF is then calculated using equation (9) with the appropriate species fractions. For the RG 1.25 species fractions and a pool water depth of 20', the effective DF is given by:
0.9975 D 0.0025 70.279 1 The DF adjustment factor for determining the effect of the change in pool water depth on the dose is calculated as defined in Section 4.3. For the RG 1.25 parameters and a pool water depth of 20', the DF adjustment factor is FDFI = 100 =1.669.
'59.904 Similarly, for the RG 1.1 83 assumptions, the inorganic DF is calculated for a pool water depth of 20' using equation (8) as 20
.DFlno.rg 20f, = ('500)23 = 222.295 .
Re.2
'Fil 0008.2DocFormGQ-~O8.
FileGQ308-2.D0C Form GO-3.08.1 Rev.2
-7 r
-Calcs. For Effect of Reduced Pool Water Levels on Fuel .Calc N
. Handling Accident Consequences Rev.
X I Safety Related Non-Safety Related Page Client Nuclear Management Company Prepared by Project Monticello Reviewed by IProj. No 11163-063 Equip. No.
The calculation of the effective DF is slightly different because it has to be normalized to an effective DF of 200 at 23'. Since the effective DF using a DF of 500 for inorganic iodine is DFEff RGI.183 ° 9985 + 0.05 285.959,
'500 1 the normalization factor for the effective DF is
- * .. Normalization Factor= 200 = 0.6994.
.-25.959 -.. -
Therefore, the calculated effective DF using the RG 1.183 parameters and a pool water depth of 20' is given by:
0.6994 DFEf 2 yOft= 09985 0.001 = 116.727.
+
222.295 1 Using the DF adjustment factor as defined in Section 4.3, the RG 1.183 parameters and a pool water depth of 20', the DF adjustment factor is FDFI = 2 -1.713.
116.727 5.2 Evaluation of an Assembly Drop in the RP The current design basis FHA assumes a drop on to the reactor core that results in failure of 125 fuel -rods assuming 8x8 fuel assemblies (Design Input (DI) 2.12). The RG 1.183 analysis fReference 7.7] also assumes failure of 125 fuel rods of 8x8 fuel. Based on the methodology in GESTAR II and since there are 60 fuel rods in the dropped assembly (DI 2.10), the number of fuel rods that are assumed to fail in the fuel assemblies in the core is 65 (125-60) assuming 8x8 fuel assemblies. The drop height of the-fuel assembly is 34', and the pool DF used in the analysis is 100, which implies a minimum pool water depth of 23' (see Section 4.2).
For a drop in the RP that does not impact the core, three items change. First, since .only one assembly is involved in the drop, the maximum number of fuel rods that can be damaged is limited to the number of fuel rods in the single assembly. Although an 8x8 fuel assembly is assumed in this evaluation to be Forn'i'GQ-3.08.1 Rev.2 GO3OB-2.DOC
'File
-Fle G0308-2.DO:C Forrn¢GO-3.08.1 flev.2
rI Calcs. For Effect of Reduced Pool Water Levels on Fuel Calc No. 2004-09840
, A s d Handling Accident Consequences Rev. sO IDate Safety Related sX Non-Safety Related Page 14 of P 3 Client Nuclear Management Company Prepared by Date Project Monticello Reviewed by Date Prol. No 11163:063 Equip. No. Approved by Date consistent with the current design basis analysis, this evaluation is applicable to all fuel types in use at MNGP.
The second item that changes is the drop height of the fuel assembly. The maximum height of a fuel assembly above the RPV flange is 3'-2" MDI 2.4). This means that the distance the assembly falls before impact is smaller than the 34' assumed in the design basis analysis by about a factor of 10. Therefore the amount of energy absorbed by the fuel assembly at impact will also be smaller by a factor of 10.
However, the fuel assembly is assumed to impact at a small angle to the vertical, subjecting all the fuel rods in the dropped assembly to bending moments. The.fuel rods absorb little energy prior to failing as a result of bending. Therefore, all of the fuel rods in the dropped assembly are assumed to fail.
The third item that changos is height of water above the dropped fuel assembly at the point ofimpact. As indicated in DI 2.1, the highest point in the RP in the path of fuel movement is actually the cattle chut&7 The minimum depth of water above the cattle chute is 21'-8" (DI 2.1), which is smaller than the 23' assumed in the design basis analysis. Therefore, the DF for the pool will decrease.
To determine the overall effect of these three items, equation (11) is evaluated. The DF adjustment factor is determined by first calculating the inorganic DF for Hi of 21'-8". Using the RG 1.25 DF, the inorganic DF is given by the following.
DFinn,rl = (133) 23' =100.168 The effective DF is then calculated using'equation (9) with the RG 1.25 species fractions.
IFf =80.268 DFzf 0.9975 0.0025 100.168 1 The DF adjustment factor is then calculated as follows.
FDF.1 = 100 = 1.246 80.268-The fuel failure adjustment factor is simply the ratio of the number of fuel rods that fail following the drop in the RP to the number of fuel rods that fail following the drop in the design basis FHA.
FFFI = 12- = 0.48 125
'file G0308-2.DOC FormGQ-3.08.1 Rev.2
-Calcs. For Effect of Reduced Pool Water Levels on Fuel Calc No. 2004-09840 So e; t u...asniye Handling Accident Consequences Rev. 0 iDate X lSafety Related Non-Safety Related Page 15 of : 3 Client Nuclear Management Company . Prepared by Date Project Monticello Reviewed by Date Pro]. No 11163-063 Equip. No. Approved by Date The thyroid dose due to a drop in the RP relative to the dose due to a design basis FHA, using the RG 1.25 assumptions, is calculated using equation (11).
DIRG 12 = (1.246)(0.48)D = 0.598D The analysis using the R-G 1.183 assumptions for iodine species fractions and DFs is similar.
DFsInorg I (500) 23' 348.745 0.6994
.- Fff= 099&Y Ool5160.298 .
348.745 1 FDF I 200
= 1.248 DJ160.298 The fuel failure adjustment factor is the same as the RG 1.25 evaluation. Therefore the thyroid dose due to a drop in the RP relative to the thyroid dose due to a design basis FHA, using the RG 1.183 assumptions, is:
DIoRG 1183 =((1.248)(0.4S)D = 0.599D As this evaluation indicates, for both the RG 1.25 and RG 1.183 assumptions, the thyroid doses resulting from a fuel assembly drop in the RP are 40% lower than the thyroid doses resulting from a design basis FHA. This is because the 25% increase in iodine activity released from the pool due to the decrease in pool water depth is offset by the 52% decrease in iodine activity released from the fuel assemblies because of a smaller number of failed fuel rods. For analyses based on RG 1.183 assumptions that are reported in TEDE, such as Reference 7.7, the portion of the TEDE from iodine activity will also decrease by about 40%. Since the noble gas contribution to the TEDE will decrease by 52%, the net effect will be an even larger decrease in the TEDE. Therefore the current design basis FHA, which is a drop of a fuel assembly onto the reactor core, is bounding for a drop of an assembly in the RP.
As these results indicate, there is still considerable margin between the thyroid dose calculated for the design basis event and the thyroid -dose that results from a drop in the RP. This implies that the water level could be even lower than the current design and result in thyroid doses that are still bounded by the design basis FHA. The minimum water level is that water level that produces a DF adjustment factor that is the inverse of the fuel failure adjustment factor so that the product of the two adjustment factors is 1.
For the drop in the RP, the limiting DF adjustment factor is given by the following expression.
Fie GQ308,2.DOC eForm GO-3.08.1 Rev.2
-Caics."For -Effect of Reduced Pool Water Levels on Fuel Calc No. 2004-09840 A L.ry"A G Handling Accident Consequences ev 0 Date X I Safety Related Non-Safety Related . Page 16 of 23 Client Nuclear Management Company Prepared by Date Project Monticello Reviewed by Date Proj. No 11163-063 Equip. No. Approved by Date FDF=X -- 82.08 0.48 Inspection of Table 1 indicates that, for the RG 1.25 analysis, the minimum height of water above the dropped assembly is 19' since it is the minimum height of water that results in a DF adjustment factor less than 2.08. Table 1 indicates that for the RG 1.183 assumptions the DF adjustment factor for 19' is 2.112, which is slightly larger than the maximum estimated value. However, the 19' value is still considered applicable to RG 1.183 analyses because the portion of the TEDE'from noble gas will be reduced by about 52% and the total TEDE will be less than the TEDE calculated for the design basis FHA.
The depthtf water'used in this analysis is 21'-8'", which is 2'-8" deeper than the minimum value of 19'.
This implies that the minimum elevation of the refueling water level can be reduced from 1026'-2" to 1023'-6" without exceeding the design basis FHA.
5.3 Evaluation of an Assembly Drop in the SFSP The drop of a fuel assembly in the SFSP is similar to the design basis accident in that-it is thedrop of a single fuel assembly onto other fuel assemblies. There are two major differences between the drop in the SFSP and the design basis analysis. The first is that the depth of water above the fuel in the SFSP is considerably less than the depth of the water above the core, which will result in less removal of iodine by the pool water. The second is that the drop height of the fuel assembly is shorter in the SFSP than in the reactor. This results in less energy in the dropped fuel assembly and therefore fewer failed rods.
Current MNGP core and refuel loads consist of GEl1 BWR/3 and GE14 BWRJ3 fuel (DI 2.16). The GE14 BWR/3 fuel assembly, which is a 10x10 assembly containing partial length fuel rods that contains the equivalent to 87.33 full length fuel rods (DI 2.17), is the heaviest of the two assemblies. The GE14 BWR/3 assembly is therefore used in this analysis because it generates more energy and results in more damage than the GEl 1 BWR/3 fuel assembly. Note that the number of damaged fuel rods assuming GE14 BWRI3 fuel can be converted to an equivalent number of damaged fuel rods for Mx8 fuel assemblies by multiplying by the ratio of the number of fuel rods per assembly (60/87.33).
The effect of the smaller pool water depth is evaluated by calculating the DF adjustment factor as described in Section 4.1. As indicated in DI 2.3, the minimum depth of water above a damaged fuel assembly is 21 '-4". Using the RG 1.25 species fractions and DFs, the DF adjustment factor is calculated as follows.
21'-4 DFmnorgl (133) 23' = 93.315 File G0308-2.DOC Form G3O-3.08.1 Rev.2
.Calcs. For Effect of Reduced -Pool Water Levels on Fuel Cale No. 20O4-09840 dwL. "yLL Handling Accident Consequences X I Safety Related I I Non-Safety Related Page 17 of 23 Client Nuclear ManagementCompany Prepared by Date Project Monticello Reviewed by Date Proj. No 11163-063 Equip. No. Approved by Date DFE. = 0.9975 0.0025 93.315 1 100 FDF.2,RGIV = 75 =1.319 75.817 Similarly, the evaluation of the DF adjustment factor using the RG 1.183 assumptions is as follows.
2P DFinorgri =M500) 23 =318.708 DFEff = 0.9 94 =150.962 0.9985 0.0015 318.708 1 200 FDFGI19 == 1.325 150.962 Estimation of the number of failed fuel rods is done by balancing the energy of the dropped assemblage against the energy required to fail a fuel rod. The weight of the dropped assemblage is the sum of the weight of the fuel assembly (553 lbs, DI 2.13) and the weight of the NF-400 refueling mast (350 Ibs, DI 2.14). The drop height of the fuel assembly in the pool is the maximum height of a fuel assembly above the spent fuel during fuel movement. From DI 2.6, this height is 2'-]0". Therefore, the potential energy to be dissipated by the first impact is (553 lbs +350 lbs)(2'-10") = 2558 ft-lbs.
One half of the energy is considered to be absorbed by the falling assembly and one half by the impacted assemblies. In addition, the energy available for clad deformation is considered to be proportional to the mass fraction of cladding in the fuel assembly, which is 0.525 (DI 2.15). Therefore, the energy absorbed by the cladding in the impacted fuel assemblies is (2558 ft - lbs)(0.5)(0.525) = 672 ft - lbs.
Each rod that fails is expected to absorb approximately 200 ft-lbs before cladding failure. 'Therefore, the number of rods that fail in the impacted assemblies following the initial impact is
.672 ftt-lbs = 3.36 rods.
200 ft - lbs Form GQ-3.08.1 Rev.2 File GOSOS-2.DOC File GQ308-2.DO0C Form GO-3.08.1 Rev.2
- Calcs. For Effect-of Reduced Pool Water Levels on Fuel
"' Handling Accident Consequences 1 X Safety Related Non-Safety Related r
Client Nuclear Management Company Prepared by Date Project Monticello . Reviewed by ,lDate Proj.No 11163-063 Equip. No. Approved by Date The dropped assembly is assumed to impact at a small angle from vertical, subjecting all the fuel rods in the dropped assembly to bending moments. The fuel rods are expected to absorb little energy prior to failure as a result of bending. For this reason, it is assumed that all the rods in the dropped assembly fail (87.33 fuel rods, DI 217). The total number of failed fuel rods on initial impact is 87.33 rods dropped assembly-+3.36 rods impacted assemblies = 90.69 rods.
The assembly is assumed to tip over and impact horizontally on the top of the fuel rack-sfrom a height of one bundle length, approximately 14.3' pD 2.11). The energy available for this second impact is calculated by assuming a linear weight distribution in the assembly with a point load at the top of the as'sernblyto repretsentthefuel grapple weight.- - : .. .
(350 lb)(143')+-(5531b)(14.3') = 8960 ft -lbs 2
As before, the energy is considered to be absorbed equally by the falling assembly and the impacted assemblies. The fraction available for clad deformation of 0.525. Therefore, the energy absorbed by the clad in the impacted fuel assemblies from the second impact is (8960 ft - lbs)(0.5)(0.525) =2352 ft - lbs, and the number of failed fuel rods in the impacted assemblies is 2352 ft-lbs = 11.76 rods.
200 ft - lbs Since all of the rods in the dropped assembly are assumed to fail on initial impact, there are no additional failed rods in the dropped assembly on the second impact. The total number of failed rods is therefore 90.69 rods initial impact + 11.76 rods second impact = 102.45 rods.
This is the number of fuel rods that fail in the GE14 BWR/3 fuel assembly. To convert this to the equivalent number of rods in an Ex8 assembly, the following expression is used.
102.45 rods in 8x8 71 rods.
87.33 rods in GE14 Using this number of failed rods, the failed fuel adjustment factor is calculated as follows.
Form GQ.3.08.1 1ev.2 File O3O6-2.DOC File GQ306-2.00C ForrnGQ-3.08.1 Rev.2
Calcs. For Effect-of Reduced Pool Water Levels on Fuel
... I Handling Accident Consequences IRev.
X l Safety Related I Non-Safety Related Client Nuclear Management Company . Prepared by
. Project Monticello JReviewed by Prol. No 11163-063 Equip. No. Approved by jDate 0 I 71F = 0.568 The impact on the thyroid dose due to a drop in the SFSP relative to the thyroid dose -due to a design basis FHA is evaluated using equation (11).
DIsRGI 25 = (1.319)(0.568)D =0.749D D1,RG 1 .1 3 = (1.325)(0.568)D = 0.753D As this evaluation-indicates, for.both theRG 1.25 and RG 1.183 assumptions, thethyroid doses resulting I "
from a fuel assembly drop in the SFSP are 25% lower than the thyroid doses risulting fromn a design basis FHA. This is because the more than 30% increase in i6dine activity released due to the decrease in pool water depth is offset by the 40% decrease in iodine activity released from the fuel assemblies because of a smaller number of failed fuel rods. For analyses based on RG 1.1 83 assumptions that are reported in TEDE, such as Reference 7.7, the portion of the TEDE from iodine activity will also decrease by about 25%. Since the noble gas contribution to the TEDE decreases by 40%, the net effect will be an even larger decrease in the TEDE. Therefore the current design basis FHA, which is a drop of a fuel assembly onto the reactor core, is bounding for a drop of an assembly in the SFSP.
As these results indicate, there is still considerable margin between the thyroid dose calculated for the design basis event and the thyroid dose that results from a drop in the SFSP. This implies that the water level could be even lower than the current design and result in thyroid doses that are still bounded by the design basis FHA. The minimum water level is that water level that produces a DF adjustment-factor that is the inverse of the fuel failure adjustment factor so that the product of the two adjustment factors is 1.
For the drop in the RP, the limiting DF adjustment factor is given by the following expression.
FDF6 = -=1.761 Inspection of Table 1 indicates that, for both the RG 1.25 and RG 1.183 analysis, the minimum height of water above the dropped assembly is 20' since it is the-minimum height of water that results in a DF adjustment factor less than 1.761.
The depth of water used in this analysis is 21'-4", which is 1'4" deeper than the minimum value of 20'.
This implies that the minimum elevation of the refueling water level can be reduced from 1026'-2" to 1024'-10" without exceeding the design basis FHA.
Form GO-3.08.1 Rev.2 File GOSOB.2.DOC File G0308-2.,DOC Form GO-3.08.1 Rev.2
Calms. For Effect of Reduced Pool Water Levels on Fuel I ,W a.%SI Handling Accident Consequences X I Safety Related I I Non-Safety Related V
Client Nuclear Management Company I IPrepared by
]Project Monticello I IReviewed by lProj.No 11163-063 Equip. No. I ]Approved by 6 RESULTS A generic analysis of the change in pool DF as a function of pool water depth is summarized in Table 1 and Figure 1.
An evaluation of the drop of a .GE14 BWR/3 fuel assembly in the RP and the SFSP was performned.
Since MNGP current core and refuel loads include only GEI I BWR/3 and GE14 3BWR/3 fuel, and since the GE14 BWR/3 fuel is heavier and therefore will cause more damage, this evaluation is bounding for both types of fuel. The results of this evaluation are summarized below.
- -.- l i
.. I .
Parameter Evaluated
- - . ." . - . ; ^;'
5-. A _-. ". - - .. . .- *-.
Drop in the RP Drop in the SFSP Fraction of the Current Design Basis RG 1.25 0.598 0.749 FHA Dose for a Refueling Water .. .
Elevation of 1026'-2" RG 1.183 0.599 0.753 Minimum Elevation of the Refueling Water that 1023'-6" 1024'-1]01 Bounds the Design Basis FHA These results indicate that the design basis EHA, which involves a drop of a fuel assembly onto the reactor core, is bounding for a'drop in the RP and the SFSP. This is because the increase in iodine activity released from the pool due to lower water height is offset by a smaller number of failed rods, which reduces the iodine activity available for release. For doses reported as TEDE rather than thyroid dose, the decrease will be even larger because the noble gas contribution to the TEDE will decrease more ihan the iodine contribution.
File-G0308-2.DOC Form GO-3.08.1 Rev.2
Calcs.For Effect of Reduced Pool Water Levels on Fuel Cale No. 2004-09840. 1 L-LIrsly Handling AccidentConsequences Rev.
I X I Safety Related Non-Safety Related age 21 of ,23 l
- Client Nuclear Management Company Prepared by Project Monticello Reviewed by I
IProj. No 11163-063 Equip. No. Approved by 7 REFERENCES 7.1 Regulatory Guide 1.25, "Assumptions Used for Evaluating the Potential Radiological Consequences of a Fuel Handling Accident in the Fuel Handling and Storage Facility for Boiling and Pressurized Water Reactors," USNRC, March 23, 1972 7.2 Regulatory Guide 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors," USNRC, July 2000 7.3 NMC DIT No. 18933, "Alternate Source Term Implementationflmproved Tech Specs," for MNGP dated 02/09/05.
7.4 U. Burley, "Evaluation of Fission Product Release and Transport," Staff Technical Paper, USNRC, 1971 -
7.5 Monticello Updated Safety Analysis Report, Rev. 21 7.6 GNF Licensing Topical Report NEDE-2401 1-P-A-14-US, "General Electric Standard Application for Reactor Fuel (Supplement for United States)," June 2000 (Referred to as GESTAR II) 7.7 Calculation 2004-02104, "MNGP AST - FHA Radiological Consequence Analysis," Rev. 0 7.8 GNF Report NEDC-32868P, "GE14 Coxhpliance With Amendment 22 of NEDE-2401 1-P-A (GESTAR II)," Revision 1, September 2000 Form 430-3.08.1 Rev.2
'File 60308-2.DOC Pile G0308-2.DOC Form 430-3,08.1 Rev.2
Caics. For Effect of Reduced Pool Water Levels on Fuel 1 ksni, JL Handling Accident Consequences X Safety Related I INon-Safety Related lClient Nuclear Management Company I IPrepared by I.I 1 iProject Monticello I IReviewed by Proj. No 111'63-063 Equip. No. I IApproved by Table 1. Decontamination Factor (DF) and DF Adjustment Factor (FDF) for Various Pool Water Depths Pool Water -Reg. Guide 1.25 Reg. Guide 1.183 Depth DF DF Ft -In Inorganic Effective FDF Inorganic Effective FDF 23 -0 133 100 1 500 200 1 22 -11 130.66 98:68- 1.013 488.87 197.43 1.013 22 -10 128.37 97.37 1.027 477.98 194.87 1.026 22-9 126.11 96.07 1.041 467.34 192.32 1.040 22 - 8 123.90 94.78 1.055 456.94 189.79 i.o054 22 -7 121,72 - --93.50- 1.069 446.76:- z187.26-. -1.068 22 - 6 119.59 92.24 1.084 436.81 184.74 1.083 22 -5 117.49 90.99 1099 427.09 182.23 1.097 22-4 115.42 89.75 1.114 417.58 179.74 1.113 22 -3 113.40 88.52 1.130 408.28 177.26 1.128 22 -2 111.40 87.31 1.145 399.19 174.79 1.144 22 -1 109.45 86.10 1.161 390.30 172.34 1.160 22 - 0 107.53 84.91 1.178 381.61 169.90 1.177 21 -11 105.64 83.73 1.194 373.12 167.48 1.194 21 -10 103.78 82.57 1.211 364.81 165.07 1.212 21 -9 101.96 81.41 1.228 356.69 162.67 1229 21 - 8 100.17 80.27 1.246 348.74 160.30 1.248 21 -7 98.41 79.14 1.264 340.98 157.94 1.266 21 - 6 96.68 78.02 1.282 333.39 155.59 1.285 21 - 5 94.98 76.91 1.300 325.97 153.27 1.305 21 -4 93.31 75.82 1.319 318.71 150.96 1.325 21 -3 91.68 74.73 1.338 311.61 148.67 1.345 21 -2 90.07 73.66 1.358 304.67 146.40 1.366 21 -1 88.48 72.60 1.377 297.89 144.15 1.387 21 -0 86.93 71:56 1.397 291.26 141.92 1.409 20 -0 70.28 59.90 1.~669 222.30 116.73 1.713 19-0 56.82 49.86 . 2.006 169.66 94.70 2.112 18 -0 45.94 41.30 2.422 129.49 75.93 2.634 17 - 0 37.14 34.06 2.936 98.83 ^60.28 3.318 16 - 0 30.02 27.99 3.572 75.43 47.46 4.214 15 - 0 24.27 22.94 4.360 57.57 37.12 5.389 Form GQ-3.08.1 Rev.2 File .G0308-2.DOC File GQ308-2.DOC Form-GOQ-3.408.1 Rev.2
'Calcs. For Effect of Reduced Pool %Vater Levels on Fuel
~ F-" t- rye Handling Accident Consequences
.1 X -I Safety Related INon-Safety Related lClient Nuclear Management Company I Prepared by IProject Monticello Reviewed by I Pro]. No 11163-063 Equip. No. I IApproved by Fiie-GQ308-2.00C Form GO-3.08.1 Rev.2