ML071580364

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Calculation H21C-102, Rev 00, U2 FHA, AST Methodology.
ML071580364
Person / Time
Site: Nine Mile Point Constellation icon.png
Issue date: 05/29/2007
From: Berg M, Pustulka H, Stinson G
Engineering Services Co
To:
Office of Nuclear Reactor Regulation
References
H21C-102, Rev 00
Download: ML071580364 (22)
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Text

Nine Mile Point Unit 2 Alternative Source Term Calculation H21C-102 "U2 FHA, AST Methodology"

Engineering ige 1(Next 2 .

AM 21 Services ist Attachment 2 Prolect: NINE MILE POINT NUCLEAR STATION Unit (1,2 or O=Both): 2 Discipline: CR Title Calculation No. H21 C-102 U2 FHA, AST Methfodology (Sub)system(s) Building Floor Elev. Index No.

N/A N/A N/A N/A Originator(s) H. Pustulka IReviewer(s)/Approver(s) M. Berg NMP Acceptance: GLM 347 ., ' / . .s/a e .. Desciption ChangNoý PrepaoWd By Date Revlewed by .- ate Aate 00 OinafTl Issueý NYA h~ v5,29/07 _____ /2§17, Computer Output/Microfilm separately filed? (Yes/No/N/A) No_ Safety Class: (*SR.NSR/Qxx): S If SR, attach or reference the associated Design Verification Report.

Superseded Document(s):

Document Cross Reference(s) - For additional references see page(s) 6 Output provided? 14 If yes, group(s)

(Y/N)

Ref Ref No. Document No. Type Index Sheet Rev No. Document No. Type Index Sheet Rev General

References:

-E-- -PA6 9' Remarks:

Confirmation Required (Yes/No): t0o Final Issue Status Turnover See Page(s):___ Req'd (Yes/N/A): Y, 10 CFR50.59 Evaluation Number(s): Component ID(s)(As shown inMEL):

Copy of Applicability Determination er 0.5 Sfeen Attached? Yes Z No*O N N/A O *lf"No", location of AD/Screen?

Key Words: Fuel Handling Accident, FHA, Design Basis, Dose, Accident

ENGINEERING SERVICES CALCULATION CONTINUATION SHEEP Page 2

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Project: Nine Mile Point Nuclear Station Unit: 2 Disposition:.

Originator/Date Reviewer/Date Calculation No, R io H.Pustulka 5/29/07 M.Berg 5/29/07 H21C-102 0 R*ef.I List of Effective Pages Page Latest Page Latest Page Latest Page Latest Page Latest Page Latest No. Rev. No. Rev. No. Rev. No. Rev. No. Rev. No. Rev.

1 0 Al-A5 0 2 0 Attach 1 0 3 0 Attach 2 0 4 0 5 0 6 0 7 0 8 0 9 0 11 0 12 0 13 0 14 0 Total Number of Calculation Pages 21

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Project.- Nine Mile Point Nuclear Station Unit: -2 Disposition:-

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Table of Contents CALCULATION COVER SHEET ................................................................................ 1 List of Effective Pages........................................................................................... 2 Table of Contents................................................................................................ 3 Purpose........................................................................................................... 4 Summary of Results..............................................................................................

Methodology ..................................................................................................... 5 Assumptions..................................................................................................... 6 References....................................................................................................... 6 Design Inputs ...................................................................................................-7 Calculation.......................................................................................................9 Results .......................................................................................................... 10 Conclusions .................................................................................................... 14 Appendix A: A Spreadsheet for the Calculation of Offsite Control Room Doses (5Pages)

Attachment 1:Design Verification Report (1Page)

Attachment 2: Design Verification Checklist (1Page)

ENGINEERING SERVICES CALCULATION CONTINUATION SHEET Page 4 r ~ (Next -5 Project.-Nine Mile Point Nuclear Station Unit: 2 Disposition: -

Originator/Date Reviewer/Date Calculation No. Revision H.Pustulka 5/29/07 M.Berg 5/29/07 H21 C-1 02 0 Ref.

Purpose The purpose of this calculation is to provide an analysis of the Fuel Handling Accident (FHA) for Nine Mile Point Nuclear Station. This analysis provides (1) implementation of the Reference 1 (AST) source terms and (2) both offsite and control room doses. This calculation analyzes the following three cases:

  • Effect of an FHA in U2 on the U2 Control Room Effect of an FHA in U 1 on the U2 Control Room
  • Effect of anFHA in U2 on the U1 Control Room

ENGINEERING SERVICES CALCULATION CONTINUATION SHEET q; Page 5 (Next 6)

Project: Nine Mile Point Nuclear Station Unit: 2 Disposition: __

0Oginator/Date Reviewer/Date. Calculation No. Revision H,Pustulka 5129/07 M.Berg 5/29/07 H21C-102 Ref.

Summary of Results Table 1 shows the results of this calculation:

Table 1: Dose Results U2 on U2 U1 on U2 U2 on U1 Limit TEDE TEDE TEDE (rem)

(rem) (rem) (rem)

Control Room 3.15E+00 2.22E-01 5.1OE-01 5 EAB 4.50E-01 6.3 LPZ 6.13E-02 6.3

1. Limit from Reference I From the above table it can be seen that applicable limits are met at all locations.

It should be noted that the U2 doses are based on a power level of 4067MWt. (Current licensed power is 3467 MWt, with a proposed Extended Power Uprate (EPU) to 120% of original power to 3988 MWt. This value is increased by 2% to 4067MWt to account for power measurement uncertainties in accordance with Reference 1.)

Methodology This dose analysis fully complies with NRC Regulatory Guide 1.183 [Ref 1]. Following accident initiation (24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after shutdown), the radionuclide inventory from the damaged fuel pins is assumed to leak out to the environment instantaneously (even though releases to the environment could be assumed to occur over a 2-hour period [Ref 1]).

In no case is RBEVS filtration credited. Due to these simplifying, conservative assumptions, a spreadsheet is used to calculate the control room, EAB, and LPZ doses.

Releases account for:

  • a 1.02 multiplier on licensed power,
  • U2 makes use of a proposed EPU of 120% where the MWt is increased from 3467 to 3988.
  • *a radial peaking factor of 1.8,
  • 5% gap activity (except 10% for Kr85 and 8% for 1131),
  • a failure fraction of 0.262% corresponding to 2 out of 764 assemblies,
  • an overall iodine DF of 200 for the refueling pool (where elemental iodine has a DF of 237 for U2, and 268 for Ul) and an infinite DF for other radionuclides except for noble gas.

The TEDE values obtained from the revised analysis are compared with the 6.3 rem FHA TEDE limit for offsite doses and the 5 rem TEDE limit for the control room [Ref 1].

ENGINEERING SERVICES CALCULATION CONTINUATION SHEET . Page 6 I(Next.7-)

Project: Nine Mile PointNuclear Station Unit: Disposition: __

Originator/Date Reviewer/Date Calculation No. Revision H.Pustulka 5/29/07 M.Berg 5/29/07 H21C-102 0 ef.

Assumptions Assumption 1: The accident is assumed to occur 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after shutdown. Consequently, core inventories were calculated for that decay time.

Justification: Reference 2, Item 1.4 Assumption 2: The release to the environment from the refueling floor occurs within two hours.

Justification: Reference 1 Assumption 3: The DF in the refueling pool does not exceed 200 for iodine, (DF of 268 for Ul and 237 for

,U2 is used for elemental iodine as shown in the Design Inputs section). No DF is applied to noble gas, and the DF for other radionuclides is assumed to be infinite.

Justification: Reference 1 Assumption 4: No credit is needed (or taken) for RBEVS filters.

Justification: Conservative References

1. ."Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors", US NRC Regulatory Guide 1.183, Revision 0, July 2000
2. PSAT 3101.CF.QA.03, "Design Data Base for Application of the Revised DBA Source Term to Nine Mile Point U2", Revision 0
3. S. L. Humphries et al, "RADTRAD: A Simplified Model for Radionuclide Transport and Removal and Dose Estimation", NUREG/CR-6604, Sandia National Laboratories, December 1997.
4. PSAT4026CF.QA.03, "Design Data Base for Application of the Revised DBA Source Term to Nine Mile Point Ul", Revision I

ENGNEENGSERVCS CALCULATION CONTINUATION SHEET Page_7__

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OriginatodDate Reviewer/Date Calculation No. Revision H. Pustulka 5/29/07 M.Berg 5/29/07 H21C-102 0 Ref.

Design Inputs Design Input Data (Reference 4- item numbers given in brackets, Reference 2-item numbers given in double brackets)

Power Level of Unit 1:1887 MWt [1.1]

(A 2% multiplier on licensed power is used to account for power measurement uncertainties in accordance with Reference 1.)

Power Level of Unit 2: 4067 MWt ((1.1))

(Licensed power is increased to account for a EPU of 20% and a 2% multiplier on the increased power level is used to account for power measurement uncertainties in accordance with Reference 1.)

Core inventory at shutdown for U1 and U2 ((1.2))

Nuclide Ci/MWt Nuclide WtCiMWt de Ci/MWt Kr83m 3.27E+03 Rul06 1.76E+04 Cs134 7.29E+03 Kr85 3.93E+02 Rhl05 2.84E+04 Cs136 2.28E+03 Kr85m 6.82E+03 Sb127 3.01E+03 Cs137 4.35E+03 Kr87 1.30E+04 Sb129 8.91E+03 Bal37m 4.12E+03 Kr88 1.83E+04 Te127 3.OOE+03 Ba139 4.89+04 Kr89 2.22E+04 Te127m 4.05E+02 Bal40 4.71+04 Rb86 7.29E+01 Te129 8.76E+03 Lal40 5.12+04 Sr89 2.45E+04 Tel29m 1.30E+03 Lal4l 4.45+04 Sr90 3.14E+03 Tel31m 3.97E+03 La142 4.29+04 Sr9l 3.10E+04 Te132 3.85E+04 Cel4l 4.47+04 Sr92 3.38E+04 1131 2.71E+04 Ce143 4.11+04 Y90 3.24E+03 1132 3.92E+04 Ce144 3.70+04 Y91 3.18E+04 1133 5.51E+04 Pr143 3.97+04 Y92 3.40E+04 1134 6.03E+04 Nd147 1.80+04 Y93 3.96E+04 1135 5.16E+04 Np239 5.78E+05 Zr95 4.46E+04 Xel31m 3.04E+02 Pu238 1.45E+02 Zr97 4.51 E+04 Xe133 5.27E+04 Pu239 1.34E+01 Nb95 4.48 E+04 Xel33m 1.63E+03 Pu240 1.89E+01 Mo99 5.13 E+04 Xe135 1.91E+04 Pu241 5.49E+03 Tc99m 4.49 E+04 Xel35m 1.09E+04 Am241 7.48E+00 RulO3 4.29 E+04 Xe137 4.80E+04 Cm242 1.85E+03 RulO5 3.01 E+04 Xe138 4.50E+04 Cm244 1.23E+02

ENGINEERING SERVICES Pg ENINERNGSEVIESCALCULATION CONTINUATION SHEET >,Page -8

  • (Next -9 Project. Nine Mile Point Nuclear Station Unit: _2_ Disposition: __

Originator/Date Reviewer/Date Calculation No. Revision H.Pustulka 5/29/07 M.Berg 5/29/07 H21C-102 0 Ref.

Total number of fuel assemblies in U1 core: 532 assemblies [2.5]

Total number of fuel assemblies in U2 core: 764 assemblies ((2.1))

Number of damaged assemblies in U 1: 2 assemblies (1 dropped and one struck) [2.11]

  • 8x8 assembly is bounding Number of damaged assemblies in U2: 2 assemblies (1 dropped and one struck) ((2.8))
  • 8x8 assembly is bounding Gap release fractions:

Radio-nuclide Release Fraction from Group Gap to Coolant Kr-85 10%

Other NG 5%

1-131 8%

Other lodines 5%

((1.5))

Peaking factor for Ul: 1.8 [1.3]

Peaking factor for U2: 1.8 ((1.3))

3 Control Room Free Volume: 1.35E+05 ft [3.9]

Control Room Free Volume: 3.81E+05 ft3 ((3.2))

Ratio of Occupied Volume to Gross Volume: 0.529 ((3.3))

3 Offsite X/Q values in sec/m :

EAB Ul : 1.90E-04 [5.1]

EAB U2: 1.19E-04 ((5.1))

LPZ UI: 1.63E-05 [5.2]

LPZ U2: 1.62E-05 ((5.2))

U 1 CR X/Q values in sec/mr3 from the specified release point:

U1 RB Blowout Panel: 4.82E-04 ((5.3))

U2 Combined Radwaste & Reactor Vent: 1.77E-04 ((5.3))

U2 CR X/Q values in sec/m 3 from the specified worst case release point:

UI RB Blowout Panel: 1.26E-04 ((5.5))

U2 Combined Radwaste & Reactor Vent: 1.09E-03 ((5.5))

Breathing Rate in m3/s (from start of release for CR): 3.5E-4 ((5.6))

Iodine Species: 99.85% elemental, 0.15% organic** ((4.4))

    • Iodine chemical form not critical since control room filters are not used. Elemental iodine DF adjusted to obtain overall iodine of DF of 200 per Reference 1. No DF applied to organic iodine Elemental Iodine DF for Water Pool for U 1: 268 [4..5]

Elemental Iodine DF for Water Pool for U2: 237 ((4.4))

ENGINEERING SERVICES CALCULATION CONTINUATIO HPe (Next-10-)

Project: Nine Mile Point Nuclear Station Unit: 2 " Disposition:

Originator/Date Reviewer/Date Calculation No. Revision H.Pustulka 5/29/07 M.Berg 5/29/07 H21C-102 0 Ref.

Calculation Core inventories at 24hrs after shutdown are calculated using a spreadsheet methodology.

The starting point of the calculation was the t =.0 shutdown inventories (Ci/MWt) from Reference 2, Item 1.2. To get the total curies of the isotope of interest one must add the curies resulting from its direct decay plus the curies resulting from decay in chains in which it is a daughter product. The final activities are shown in Table 2.

Table 2 - Core Inventories (per MWt) for FHA Shutdown Adjusted' Branching 24 Hours3 K Nuclide (Ci/MWt) (Ci/MWt) Fraction 2 (Ci/MWt) Decay /sec Kr85m 6.82E+03 same N/A 1.66E+02 4.30E-05 Kr85 3.93E+02 7.86E+02 N/A 7.86E+02 2.05E-09 Kr87 1.30E+04 same N/A 2.81E-02 1.51E-04 Kr88 1.83E+04 same N/A 5.23E+01 6.78E-05 Tel3lm 3.97E+03 * *

  • 6.42E-06 1131 2.71E+04 4.34E+04 N/A 4.OOE+04 9.98E-07 Xel3lm 3.04E+02 same 0.011 3.03E+02 6.74E-07 Tel32 3.85E+04 * *
  • 2.46E-06 1132 3.92E+04 same N/A 3.21E+04 8.37E-05 1133 5.51E+04 same N/A 2.48E+04 9.26E-06 Xel33m 1.63E+03 same 0.029 1.48E+03 3.67E-06 Xe133 5.27E+04 same 0.971 5.09E+04 1.53E-06 1135 5.16E+04 same N/A 4.18E+03 2.91E-05 Xe135m 1.09E+04 * *
  • 7.56E-04 Xe135 1.91E+04 same 0.835 1.23E+04 2.12E-05

'*' denotes where a nuclide is considered as a parent only

1. The adjusted column multiplies Kr85 by 2 and 1131 by 1.6 to account for the gap release fractions (Kr85=l 0%, I131=8%, all other noble gases and iodines=5%)[Ref 1]. 'Same' refers to the fact that it is the same as the value listed for shutdown.
2. Branching fractions 'F,', multiplier applied to the iodine parent of the specific Xe radionuclide.
3. Calculated using the decay expression :a 1 ao*e"Kt 1131, Xel3lm, 1132, Xel33m, Xel33 and Xe135 also consider parent nuclide decay in determining their total (Ci/MWt). The following equation is used for these radionuclides:

FB .apo" K (e-KP - eKt) a, = a 0 *e-Kt +

K -Kp

'p' denotes parent value, and 'a' is activity in(Ci/MWt) and K is the decay constant It should be noted that the above equation is expanded for Xe133 and Xe135 to accommodate the presence of two parent nuclides.

Note:

" For Kr83m, Kr89, Xe135m, Xe137, Xe138 and 1134, the activity left after 24 hrs is negligible.

" No credit is taken for the RBEVS filtration.

ENGINEERING SERVICES CALCULATION . SHEET Page 10.

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Offsite and control room doses are analyzed using a spreadsheet methodology as discussed in Appendix A. The spreadsheet inputs are described below.

Control Room Occupied Volume (Row 3):

  • U1 CR Volume is 1.35E+05ft3 per Reference 4.

U2 CR Volume is 3.8 1E+05 ft 3*0.529=2.02E+05 ft3 .

Scaling Factors (Rows 4, 5 & 6):

Scaling Factor 1 is the Power Level in MW(t), used to convert the core inventory concentration to total activity. Scaling Factor 2 is the peaking factor. Scaling Factor 3 is the gap fraction, multiplied by the failure fraction.

DF (Row 7):

DF for Elemental I is 268 for U1, and 237 for U2, and the DF for Alkali metals is 1.

Source in Ci/MW(t) (column 2):

Values are taken from the core inventories adjusted for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, presented in column 3 of Table 2.

Nuclide Specific Scaling Factor (column 3):

No additional Nuclide Specific Scaling Factor is needed, the values in this column are set to unity for this calculation.

Table 3: Case In ut Summary U2 FHA U1 FHA U2 FHA on U2 CR on U2 CR on U1 CR EAB X/Q (sec/m 3 ) 1.19E-04 1.90E-04 1.19E-04 LPZ X/Q (sec/m 3) 1.62E-05 1.63E-05 1.62E-05 CR X/Q (sec/mr) 1.09E-03 1.26E-04 1.77E-04 Scaling Factor 1 (MWt) 4067 1887 4067 Scaling Factor 2 1.8 1.8 1.8 Scaling Factor 3 0.05"(2/764) 0.05*(2/532) 0.05*(2/764)

DF of Elemental 1 237 268 237 DF of Alkali Metals 1 1 1 The results of this calculation can be seen in Table 4.1, 4.2 and 4.3 for cases U2 on U2, Ul onU2 and U2 on U1 respectively.

Results An EXCEL spreadsheet calculation has been carried out to obtain the dose results for a FHA. . The spreadsheet methodology used is described in Appendix A with the following notable exceptions:

o The Source values have been decayed 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the results of which are presented in Table 2.

o The Nuclide Specific Scaling Factor has been set to I for all nuclides because the scaling has already been credited, as can be seen in the 'Adjusted' column of Table 2.

o Calculated decayed inventories from Table 2 for Iodine have been partitioned into elemental and organic forms per the design input section. (99.85% elemental, 0.15%

organic).

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Originator/Date Reviewer/Date Calculation No. Revision H. Pustulka 5/29/07 M.Berg 5/29/07 H21C-102 0 Ref.

TABLE 4.1: Doses after a Fuel Handling Accident in U2 on U2 CR NMP2 FHA (U2onU2) EAB LPZ CR Dispersion (X/Qs) = 1.19E-04 1.62E-05 1.09E-03 sec/m3 CR Vol = 2.02E+05 ft3 w/ finite volume gamma correction = 0.052911 Scaling Factor 1 = 4067 MW(t)

Scaling Factor 2 = 1.8 Peaking Factor*

Scaling Factor 3 = 1.31E-04 (0.05 gap fract x 2 of 764assemblies)

DF for Elemental I 237 DF for Alkali Metals = I Nuclide- WB CEDE TEDE CR EAB LPZ CR Source: Specific DCF DCF DCF DCF TEDE TEDE TEDE Units >>> Ci/MW(t) Scaling rem-m3 rem/Ci rem-m3 rem-m3 rem rem rem Nuclide Factor Ci-sec Ci-sec Ci-sec Kr83m 0 5.55E-06 0 5.55E-06 2.94E-07 0.OOE+00 0.OOE+00 0.OOE+00 Kr85m 1.66E+02 0.0277 0 0.0277 0.001466 5,24E-04 7.14E-05 2.54E-04 Kr85 7.86E+02 0.00044 0 0.00044 2.33E-05 3.94E-05 5.37E-06 1.91E-05 Kr87 2.81E-02 0.152 0 0.152 0.008042 4,87E-07 6.63E-08 2.36E-07 Kr88 5.23E+01 0.501 8.36E+01 0.53026 0.055768 3116E-03 4.30E-04 3.05E-03 Kr89 0 0.323 0 0.323 0.01709 0.OOE+00 0.OOE+00 0.OOE+00 Xel3 lm 3.03E+02 0.00144 0 0.00144 7.62E-05 4.98E-05 6.77E-06 2.41E-05 Xe133m 1.48E+03 0.00507 0 0.00507 0.000268 8.56E-04 1.16E-04 4.15E-04 Xe133 5.09E+04 0.00577 0 0.00577 0.000305 3.35E-02 4.56E-03 1.62E-02 Xe135m 0 0.0755 0 0.0755 0.003995 0.OOE+00 0.OOE+00 0.OOE+00 Xe135 1.23E+04 0.044 0 0.044 0.002328 6.17E-02 8.40E-03 2.99E-02 Xe137 0 0.0303 0 0.0303 0.001603 0.OOE+00 0.OOE+00 0.OOE+00 Xe138 0 0.213 0 0.213 0.01127 0.OOE+00 0.OOE+00 0.OOE+00 1131Org 6.OOE+01 0.0673 3.29E+04 11.5823 11.51856 7.92E-02 1.08E-02 7.22E-01 11320rg 4.8 1E+01 0.414 3.8 1E+02 0.54735 0.155255 3.OOE-03 4.09E-04 7.80E-03 I133Org 3.71EE+l01 0.109 5.85E+03 2.1565 2.053267 9.12E-03 1.24E&03 7.96E-02 11340rg 0.OOE+00 0.481 1.3 1E+02 5.27E-01 0.0713 0.OOE+00 0.OOE+00 0.OOE+00 I135Org 6.26E+00 0.307 1.23E+03 0.7375 0.446744 5.26E-04 7.17E-05 2.92E-03 113 1Elem 4.OOE+04 0.0673 3.29E+04 11.5823 11.51856 2.23E-01 3.03E-02 2.03E+00 I132Elem 3.20E+04 0.414 3.8 1E+02 0.54735 0.155255 8.43E-03 1.15E-03 2.19E-02 I133Elem 2.47E+04 0.109 5.85E+03 2,1565 2.053267 2.56E-02 3 .49E-03 2.23E-01 I134Elem 0.OOE+00 0.481 1.31E+02 0.52685 0.0713 0,OOE+00 0.OOE+00 0.OOE+00 I135Elem 4.17E+03 0.307 1.23E+03 7.38E-01 0.446744 1.48E-03 2.01E-04 8.21E-03 Rb86 0 0.0178 6.62E+03 2.3348 2.317942 0.OOE+00 0.OOE+00 0.OOE+00 Cs134 0 0.28 4.63E+04 16.485 16.21982 0.OOE+00 0.OOE+00 0.OOE+00 Cs136 0 0.392 7.33E+03 2.9575 2.586241 0.OOE+00 0.OOE+00 0.OOE+00 Cs137 0 0.101 3.19E+04 11.266 11.17034 0.OOE+00 0.OOE+00 0.OOE+00 Cs138 0 0.4255 1.15E+02 0.465904 0.062918 0.OOE+00 0.OOE+00 0.OOE+00 Total TEDE 4.50E-01 6.13E-02 I 3.15E+00

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Originator/Date Reviewer/Date Calculaton No. Revision H. Pustulka 5/29/07 - M. Berg 5/29/07 H21C-102 0 Tfef.

TABLE 4.2: Doses after a Fuel Handling Accident in U1 on U2 OR NMP2 FHA (UI onU2) EAB LPZ CR Dispersion (X/Qs) = 1.90E-04 1.63E-05 1.26E-04 sec/m3 CR Vol = 2.02E+05 ft3 w/ finite volume gamma correction = 0.052911 Scaling Factor 1 = 1887 MW(t)

Scaling Factor 2 = 1.8 Peaking Factor Scaling Factor 3 = 1.88E-04 (0.05 gap fract x 2 assy out of 532)

DF for Elemental I = 268 DF for Alkali Metals = I Nuclide- WB CEDE TEDE CR EAB LPZ CR Source: Specific DCF DCF DCF DCF TEDE TEDE TEDE Units >> Ci/IMW(t) Scaling rem-m3 rem/Ci rem-m3 rem-m3 rem rem rrem Nuclide Factor Ci-sec Ci-sec Ci-sec Kr83m 0 5.55E-06 0 5.55E-06 2.94E-07 0.OOE+00 0.OOE+00 0.OOE+00 Kr85m 1.66E+02 0.0277 0 0.0277 0.001466 5.58E-04 4.79E-05 1.96E-05 Kr85 7.86E+02 0.00044 0 0.00044 2.33E-05 4.20E-05 3.60E-06 1.47E-06 Kr87 2.8 1E-02 0.152 0 0.152 0.008042 5.18E-07 4;45E-08 1.82E-08 Kr88 5.23E+01 0.501 8.36E+01 0.53026 0.055768 3.36E-03 2.89E-04 2.35E-04 0

Kir89 0.323 0 0.323 0.01709 0.00E+00 0.OOE+00 0.00E+00 Xel31m 3.03E+02 0.00144 0 0.00144 7.62E-05 5.29E-05 4.54E-06 1.86E-06 Xel33m 1.48E+03 0.00507 0 0.00507 0.000268 9.1OE-04 7.81E-05 3.19E-05 Xe133 5.09E+04 0.00577 0 0.00577 0.000305 3.56E-02 3.06E-03 1.25E-03 Xe135m 0 0.0755 0 0.0755 0.003995 0.OOE+0O O.OOE+00 0.OOE+00 Xe135 1.23E+04 0.044 0 0.044 0.002328 6.57E-02 5.63E-03 2.30E-03 Xe137 0 0.0303 0 0.0303 0.001603 0.OOE+00 0.OOE+00 0.00E+00 Xe138 - 0 0.213 0 0.213 0.01127 0.OOE+00 0.OOE+00 0.OOE+00 1131Org 6.OOE+01 0.0673 3.29E+04 11.5823 11.51856 8.43E-02 7.23E-03 5.56E-02 11320rg 4.8 1E+01 0.414 3.81E+02 0.54735 0.155255 3.19E-03 2.74E-04 6.01E-04 11330rg 3.7 1E+01 0.109 5.85E+03 2.1565 2.053267 9.71E-03 8.33E-04 6.13E-03 I1340rg 0.OOE+00 0.481 1.31E+02 5.27E-01 0.0713 0.OOE+00 0.OOE+00 0.OOE+00 1135Org 6.26E+00 0.307 1.23E+03 0.7375 0.446744 5.60E-04 4.81E-05 2.25E-04 113 lElem 4.OOE+04 0.0673 3.29E+04 11.5823 11.51856 2.10E-01 1.80E-02 1.38E-01 I132Elem 3.20E+04 0.414 3.8 1E+02 0.54735 0.155255 7.93E-03 6.80E-04 1.49E-03 I133Elem 2.47E+04 0.109 5.85E+03 2.1565 2.053267 2.41E-02 2.07E-03 1.52E-02 I134Elem 0.OOE+00 0.481 1.31E+02 0.52685 0.0713 0.OOE+00 0.OOE+00 0.OOE+00 1135Elem 4.17E+03 0.307 1.23E+03 7.38E-01 0.446744 1.39E-03 1.19E-04 5.59E-04 Rb86 0 0.0178 6.62E+03 2.3348 2.317942 0.OOE+00 0.OOE+00 0.OOE+00 Cs134 0 0.28 4.63E+04 16.485 16.21982 0.OOE+00 0.OOE+00 0.OOE+00 Cs136 0 0.392 7.33E+03 2.9575 2.586241 0.OOE+00 0.OOE+00 0.OOE+00 Cs137 0 0.101 3.19E+04 11.266 11.17034 0.OOE+00 0.OOE+00 0.OOE+00 Cs138 0 0.4255 1.15E+02 0.465904 0.062918 0.OOE+00 0.00E+00 0.OOE+00 4.47E-O1 3.84E-02 I 2.22E-Ol Total TEDE Total .TEDE 4.47E-01 3.84E-02 2.22E-01

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TABLE 4.3: Doses after a Fuel Handling Accident in U2 on U1 CR NMP2 FHA (U2onUl) EAB LPZ CR Dispersion (X/Qs) = 1.19E-04 1.62E-05 1.77E-04 sec/m3 CR Vol = 1.35E+05 ft3 w/ finite volume gamma correction = 0.046212 Scaling Factor I = 4067 MW(t)

Scaling Factor 2 = 1.8 Peaking Factor Scaling Factor 3 = 1.3 1E-04 (0.05 gap fract x 2 of 764 assemblies)

DF for Elemental I = 237 DF for Alkali Metals = I i

Nuclide- WB CEDE TEDE CR EAB LPZ CR Source: Specific DCF DCF DCF DCF TEDE TEDE TEDE Units >> Ci/MW(t) Scaling rem-m3 rem/Ci rem-m3 rem-m3 rem rem rem Nuclide Factor Ci-sec Ci-sec Ci-sec Kr83m 0 5.55E-06 0 5.55E-06 2.56E-07 0.OOE+00 0,OOE+00 0.OOE+00 Kr85m 1.66E+02 0.0277 0 0.0277 0.00128 5.24E&04 7.14E-05 3.60E-05 Kr85 7.86E+02 0.00044 0 0.00044 2.03E-05 3.94E-05 5.37E-06 2,71E-06 Kr87 2.81E-02 0.152 0 0.152 0.007024 4.87E-07 6.63E-08 3.35E-08 Kr88 5.23E+01 0.501 8.36E+01 0.53026 0.052412 3,16E-03 4.30E-04 4.65E-04 Kr89 0 0.323 0 0.323 0.014926 0.OOE+00 0.OOE+00 0.OOE+00 Xel31m 3.03E+02 0.00144 0 0.00144 6.65E-05 4,98E-05 6.77E-06 3.42E-06 Xel33m 1.48E+03 0.00507 0 0.00507 0.000234 8.56E-04 1.16E-04 5.88E-05 Xe133 5.09E+04 0.00577 0 0.00577 0.000267 3.35E-02 4.56E-03 2.30E-03 Xel35m 0 0.0755 0 0.0755 0.003489 0.OOE+00 0.OOE+00 0.OOE+00 Xel35 1.23E+04 0.044 0 0.044 0.002033 6.17E-02 8.40E-03 4.24E-03 Xel37 0 0.0303 0 0.0303 0.0014 0.OOE+00 0.OOE+00 0.OOE+00 Xe138 0 0.213 0 0.213 0.009843 0.OOE+00 0.OOE+00 0.OOE+00 1131Org 6.OOE+01 0.0673 3.29E+04 11.5823 11.51811 7.92E-02 1.08E-02 1.17E-01 11320rg 4.8 1E+01 0.414 3.81 E+02 0.54735 0.152482 3.OOE-03 4.09E-04 1.24E-03 1133Org 3.71E+01 0.109 5.85E+03 2.1565 2.052537 9.12E-03 I .24E-03 1I29E-02 11340rg 0.OOE+00 0.481 1.31E+02 5.27E-01 0.068078 0.OOE+00 0.OOE+00 0.OOE+00 I135Org 6.26E+00 0.307 1.23E+03 0.7375 0.444687 5.26E-04 7.17E-05 4.72E-04 I13lElem 4.OOE+04 0.0673 3.29E+04 11.5823 11.51811 2.23E-01 3.03E-02 3.30E-01 I132Elem 3.20E+04 0.414 3.81 E+02 0.54735 0.152482 8.43E-03 1.15E-03 3.49E-03 I133Elem 2.47E+04 0.109 5.85E+03 2.1565 2.052537 2.56E-02 3.49E-03 3.63E-02 I 34Elem 0,O0E+00 0.481 1.31E+02 0.52685 0.068078 0.OOE+00 0.OOE+00 0.OOE+00 113 5Elem 4.17E+03 0.307 1.23E+03 7.38E-01 0.444687 1.48E-03 2.01E-04 1.33E-03 Rb86 0 0.0178 6.62E+03 2.3348 2.317823 0.OOE+00 0.OOE+00 0.OOE+00 Cs134 0 0.28 4.63E+04 16.485 16.21794 0.OOE+00 0.OOE+00 0.OOE+00 Cs136 0 0.392 7.33E+03 2.9575 2.583615 0,OOE+00 0.OOE+00 0.OOE+00 Cs137 0 0.101 3.19E+04 11.266 11.16967 0.OOE+00 0.OOE+00 0.OOE+00 Cs138 0 0.4255 1.15E+02 0.465904 0.060067 0.OOE+00 0.OOE+00 0.OOE+00 Total TEDE 4.50E-01 6.13E-02 5.10E-01

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The control room dose limit is 5 rem TEDE and that the offsite dose limit for the FHA is 6.3 rem TEDE

[Ref 1]. The results from Tables 4.1, 4.2 and 4.3 can be compared to these limits. Note that there is considerable margin for the doses.

Conclusions The FHA control room and offsite doses for all cases are well within their Reference 1 limits.

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Appendix A A Spreadsheet for the Calculation of Offsite and Control Room Doses Background/Methodology It is desirable for simplicity in many cases to calculate a bounding radiation dose for a given accident using several basic assumptions. These are as follows:

o It is assumed that the release of activity may be defined at the outset (i.e., there are no time-dependent mechanisms that modify the amount of activity that's released; e.g., no delayed filtration or holdup).

o It is assumed that the release is instantaneous and complete, and the transport to the receptor is instantaneous, as well. Therefore, no radioactive decay needs to be considered. Note that the activity release, A, may, in fact, occur over a given time duration, t, at a rate A/t. As long as the exposure time is equal to duration of the release, time cancels out of the integrated dose analysis.

o It is assumed that the release is limited to coolant and/or gap activity (i.e., only a limited number of radionuclides are included in the sheet).

o It is assumed that the chemical/physical form of the iodine as it is released is limited to organic and elemental.

o No credit for control room emergency ventilation (i.e., filtration) is assumed.

o It is assumed that the atmospheric dispersion for the duration of the release may be characterized by a single value of X/Q for each location (EAB, LPZ, and control room).

o It is assumed that the exchange rate of the control room with the environment is infinite so that the concentration of activity inside the control room is equal to that in the atmosphere.

o It is assumed that the breathing rate of exposed individuals is a constant 3.5E-4 m 3/sec.

Effectively, this means the release actually must occur over a period of no more than eight hours in order for the LPZ dose not to be overstated.

o It is assumed that the control room occupancy factor is unity.

In addition, for the spreadsheet to be consistent with Reference I, Dose Conversion Factors (DCFs) based on References 2 and 3 must be used. These are taken from the default TID.INP and FGR60.INP default files of Reference 4. Breathing rates and occupancy factors are taken from Reference 1.

The following section describes the development of such an Excel spreadsheet.

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Spreadsheet Development The spreadsheet is displayed at the end of this section, just before the references.

At the top of the spreadsheet (in the first row) is the title. An example might be "NMP2 MSLB". In the second row may be found the EAB, LPZ, and control room X/Qs in units of seconds/m 3. The control room volume in ft3 is given in the third column. It is included to provide the basis for the finite volume correction factor for gamma shine dose provided by Reference 1 (calculated to the right of the control room volume).

The next three rows provide scaling factors that apply equally to all of the radionuclides listed and to all of the calculated doses (EAB, LPZ, and control room). For example, in an FHA analysis, if the core-wide activity available for release is expressed as Ci/MWt, one scaling factor may be the power of the core, a second may be the peaking factor to account for the fact that the specific activity in the affected fuel bundles may be greater than the core average, and the third may be the fraction of the core's activity that is released from the damaged bundles (i.e., the fraction of the core activity assumed to be in the gap multiplied by the fraction of the core fuel bundles that are damaged by the drop). Space is available next to each scaling factor to annotate what each value represents.

DFs are specifically provided in the next row after the scaling factors. One DF is provided for elemental iodine and one for alkali metals (i.e., Cs and Rb).

The "Source" column (i.e., the second column) has already been mentioned. One space is provided under "Source" to identify the units of "Source". For each of the coolant and/or gap release radionuclides identified in the first column, a "Source" entry may be made.

In the third column, there is a place for scaling factors unique to individual radionuclides. For example, gap fractions that differ from the general gap fraction may be accommodated using these radionuclide-specific scaling factors. If the 1-131 gap fraction is 8% vs. the general value of 5%, then the "Source" for 1-131 would have to be increased by a factor of 1.6 to account for that difference.

That factor may be entered in the third column.

In the fourth column, the DCFs for immersion dose are provided. As noted previously, these are taken from Reference 4 TID.INP and FGR60.INP with the multiplication of"Cloudshine-Effective" by 3.7E12 to convert Sv-m 3/Bq-sec to rem-m 3/Ci-sec. In the fifth column, the "Inhaled-Chronic-Effective" values from FGR60.INP have been multiplied by the same 3.7E 12 to convert Sv/Bq to rem/Ci. Note that these DCFs include short-lived decay daughters as long as (1) the daughter has a half-life less than 90 minutes and (2) the daughter has a half-life less than 0.1 times the parent. One exception has been made to this rule. Because of its importance as a decay daughter, the DCFs for Rb-88 have been added to those for Kr-88 even though the half-life of Rb-88 (17.8 minutes) is

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slightly greater than 10% of its parent Kr-88 (170.4 minutes).

In the sixth column, a TEDE DCF is prepared which is the sum of the immersion DCF and the inhalation DCF times the assumed breathing rate of 3.5E-4 m 3/sec.

In the seventh column, a control room DCF is defined which is similar to the TEDE DCF. However, the immersion DSF is diminished by the finite volume correction factor defined as the following in Reference 1:

DDEfiite - DEV'a338 DD~i~e =DBE 0 .V 1173 For a control room volume of 135,000 ft3, for example, the factor is 0.0462. Note that this factor appears next to the control room volume at the top of the spreadsheet. It is -unity for a control room volume of 1.2E9 ft3.

The eighth column is the EAB dose, the product of Columns 2, 3, and 6, the three general scaling factors, and the EAB X/Q. Note that if a release of the activity, A, in Column 2 occurs over time, t, the release rate is A/t assuming a unit scaling factor in Column 3. When multiplied by the X/Q, the product is the concentration present at the X/Q location for the time, t (i.e., for the duration of the release). When multiplied by the DCF (Column 6) in units of rem-volume/Ci-time, the result is a dose rate for the duration, t. As long as it is assumed that the exposure duration, t', is the same as release duration, t, then the immersion + inhalation dose is simply the product as just described. In the last row of Column 8, the EAB dose is summed for all radionuclides in Column 1. Note that in calculating the EAB dose, the elemental iodine dose is reduced by the DF for elemental iodine and the alkali metal dose is reduced by the DF for alkali metals.

In Column 9, the Column 8 results are adjusted by the ratio of the LPZ X/Q to the EAB X/Q to obtain the LPZ dose.

Finally, in Column 10, the Column 8 results are adjusted by the ratio of the control room X/Q to the EAB X/Q and by the ratio of the control room DCF to the TEDE DCF to obtain the control room dose contribution for each radionuclide. As with the EAB and the LPZ doses, these are summed at the bottom of column to obtain the total control room TEDE.

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Spreadsheet for Simplified Dose Evaluation TITLE EAB LPZ CR Dispersion (X/Qs) = x.xxE-xx x.xxE-xx x.xxE-xx sec/m3 CR Vol = 1.20E+09 ft3 w/ finite volume gamma correction 0.999 Scaling Factor 1 = I Scaling Factor 2 = I Scaling Factor 3 = I DF for Elemental I 1 DF for Alkali Metals = 1 Nuclide- WB CEDE TEDE CR EAB LPZ CR Source: Specific DCF DCF DCF DCF TEDE TEDE TEDE Units >> Scaling rem-m3 rem/Ci rem-m3 rem-m3 rem rem rem Nuclide Factor Ci-sec Ci-sec Ci-sec Kr83m 0., 5.55E-06 0 5.55E-06 5.54E-06 0.OOE+00 0.OOE+00 0.OOE+00 Kr85m 0 1 0.0277 0 0.0277 0.027666 0.OOE+00 0.OOE+00 0.OOE+00 Kr85 0 0.00044 0 0.00044 0.000439 0.OOE+00 0.OOE+00 0.OOE+00 Kr87 0 0.152 0 0.152 0.151813 0.OOE+00 0.OOE+00 0.OOE+00 Kr88 0 0.501 8.36E+01 0.53026 0.529643 0.OOE+00 0.OOE+00 0.OOE+00 Kr89 0 1 0.323 0 0.323 0.322603 0.OOE+00 0.OOE+00 0.OOE+00 Xel31m 0 0.00144 0 0.00144 0.001438 0.OOE+00 0.OOE+00 0.OOE+00 Xe133m 0 0.00507 0 0.00507 0.005064 0.OOE+00 0.OOE+00 0.OOE+00 Xe133 0 1 0.00577 0 0.00577 0.005763 0.OOE+00 0.OOE+00 0.OOE+00 Xe135m 0 0.0755 0 0.0755 0.075407 0.00E+00 0.OOE+00 0.OOE+00 Xe135 0 0.044 0 0.044 0.043946 0.OOE+00 0.OOE+00 0.OOE+00 Xe137 0 I 0.0303 0 0.0303 0.030263 0.OOE+00 0.OOE+00 0.OOE+00 Xe138 0 0.213 0 0.213 0.212738 0.OOE+00 0.OOE+00 0.OOE+00 I131Org 0 1 0.0673 3.29E+04 11.5823 11.58222 0.OOE+00 0.OOE+00 0.OOE+00 11320rg 0 0.414 3.81 E+02 0.54735 0.546841 0.OOE+00 0.00E+00 0.OOE+00 11330rg 0 0.109 5.85E+03 2.1565 2.156366 0.OOE+00 0.OOE+00 0.OOE+00 11340rg 0 1 0.481 1.31E+02 5.27E-01 0.526258 0.OOE+00 0.OOE+00 0.OOE+00 11350rg 0 1 0.307 1.23E+03 0.7375 0.737122 0.OOE+00 0.OOE+00 0.OOE+00 I131Elem 0 0.0673 3.29E+04 11.5823 11.58222 0.OOE+00 0.OOE+00 0.OOE+00 II32Elem 0 0.414 3.8 1E+02 0.54735 0.546841 0.OOE+00 0.OOE+00 0.OOE+00 I133Elem 0 0.109 5.85E+03 2.1565 2.156366 0.OOE+00 0.OOE+00 0.OOE+00 I134Elem 0 0.481 1.31 E+02 0.52685 0.526258 0.OOE+00 0.OOE+00 0.00E+00 I135Elem 0 0.307 1.23E+03 7.38E-01 0.737122 0.OOE+00 0.OOE+00 0.OOE+00 Rb86 0 0.0178 6.62E+03 2.3348 2.334778 0.OOE+00 0.OOE+00 0.OOE+00 Cs134 0 0.28 4.63E+04 16.485 16.48466 0.OOE+00 0.OOE+00 0.OOE+00 Cs136 0 0.392 7.33E+03 2.9575 2.957018 0.OOE+00 0.OOE+00 0.OOE+00 Cs137 0 0.101 3.19E+04 11.266 11.26588 0.OOE+00 0.OOE+00 0.00E+00 Cs138 0 0.4255 1.15E+02 0.46590z 0.46538 0.OOE+00 0.OOE+00 0.OOE+00 Total TEDE 0.OOE+00 O.OOE+O0 O.OOE+O0

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Project. Nine Mile Point NuclearStation Unit: 2 Disposition: __

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References A-I Regulatory Guide 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors", July 2000 A-2 K.F. Eckerman et al., "LimitingValues of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion," Federal Guidance Report 11, EPA-520/1-88-020, Environmental Protection Agency; 1988.

A-3 K.F. Eckerman and J.C. Ryman, "External Exposure to Radionuclides in Air, Water, and Soil," Federal Guidance Report 12, EPA-402-R-93-081, Environmental Protection Agency, 1993 A-4 NUREG/CR-6604, "RADTRAD: A Simplified Model for RADionuclide Transport and Removal And Dose Estimation", December 1997

"ITACHMEt4T 1: ESIGAR YE1IAl~REP~ORT Document being design-verified: 0 DCP IElCalc El Spec Ul NER U DBD E] Other Doc#, Rev and

Title:

H21C-102, Revision 0: U2 FHA, AST Methodology Extent of Design Verification (Briefly describe):

This calculation was design verified by 1) validating all input with respect to the input database making sure the appropriate input values were used; 2) validating that all assumptions are conservative and conform to RG 1.183 AST requirements; 3) validating the calculation methodology and calculation tools (i.e. spreadsheet) as being acceptable for the task; and 4) validating final results to make sure that they are as expected. Additional check calculations were also performed.

Method of Design Verification:

IE Design Review E] Qualification Testing U3 Alternate Calculations UI Applicability of Proven Design Results of Design Verification:

El Fully acceptable with no issues identified UJ Fully acceptable based on the following issues identified and resolved:

All inputs were appropriate and all assumptions valid (no further validation of assumptions are required). The calculation methodologies were appropriate for the task. All calculated values conform to expected results. The calculation made several assumptions which simplified the analysis, and also added siqnificant conservatism.

Among these conservatisms is the control room being essentially open to the environment, so that no Habitability Zone protections were taken into account (such as filtration, delayed inflow, etc). Check calculations were performed and the results were within 4% of the values in the text. Minor issues were commented upon and corrected prior to the final draft of the calculation.

U Continuation Page Follows Discipline Involvement and Approvals:

Lea Design Verifier!: M, Berg

. :5129/2001 Discipline Design Verifiers, if required:

'N/A "_*_*

DiD~clpInS Name Signature, Dat

AT,AZ HIWEN T 2: bE 14& -ERIEICAf TI-ON t HERckLI The following questions are required to be addressed based on the Nine Mile Point commitment to NQA-1 (1983) for design verification activities. This checklist is intended to assist when using the Design Review method of design verification to ensure relevant items are addressed in the verification effort. Each "No" answer will require correction or resolution by the originator of the document being verified prior to full acceptance by the design verifier(s).

Doc#: H21C-102, Rev 0 Lead Design Verifiers M.Berg Name:

~~ Review~ Ch~eck Items~~~~

~ ~ eiw4se

~ Adrse ~hBsso Yes ~No -N/A~

1. Were the inputs correctly selected ?

x

2. Are assumptions necessary to perform the design activity adequately described and reasonable ? Where necessary, are the assumptions identified for subsequent re-verifications when the detailed activities are completed ? X
3. Was an appropriate design method used?

x

4. Were the design inputs correctly incorporated into the design ?

x

5. Is the design output reasonable compared to design inputs ?

x

6. Are the necessary design input and verification requirements for interfacing organizations specified in the design documents or in supporting procedures or instructions ? X
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