ML022680206

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Supplement to License Amendment Request for Full-Scope Implementation of the Alternate Source Term Technical Specification Change (TSC) Number 2001-07
ML022680206
Person / Time
Site: Oconee  
Issue date: 09/12/2002
From: Mccollum W
Duke Energy Corp
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
TSC 2001-07
Download: ML022680206 (36)
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{{#Wiki_filter:D uke Duke Energy 4kEnergy Oconee Nuclear Station 7800 Rochester Highway Seneca, SC 29672 W. R. Mc~ollum, Jr. (864) 885-3107 OFCE Vice President (864) 885-3564 FAX September 12, 2002 U. S. Nuclear Regulatory Commission Washington, D. C. 20555 Attention: Document Control Desk

Subject:

Oconee Nuclear Station Docket Numbers 50-269, 270, and 287 Supplement to License Amendment Request for Full Scope Implementation of the Alternate Source Term Technical Specification Change (TSC) Number 2001-07 Pursuant to Title 10, Code of Federal Regulations, Part 50, Section 90 (10 CFR 50.90), Duke Energy (Duke) proposes to amend Appendix A, Technical Specifications, for Facility Operating Licenses DPR-38, DPR-47 and DPR-55 for Oconee Nuclear Station, Units 1, 2, and 3. The license amendment requests approval of the Alternate Source Term (AST) analysis methodology for Oconee Nuclear Station that will support simplification of Ventilation System testing requirements during core alterations or movement of irradiated fuel. The License Amendment Request (LAR) was submitted to the Nuclear Regulatory Commission (NRC) on October 16, 2001. Duke received additional questions from the NRC related to the,AST LAR. In a meeting with the NRC on March 21, 2002, Duke discussed the additional questions with the staff. A common understanding of the questions and required responses were obtained, and an initial supplement to the LAR was submitted on May 20, 2002. Follow-up disctssions have indicated that some additional information would be helpful for the NRC review. provides this additional information. provides copies of the revised pages from the previous submittals. A4CMo

U. S. Nuclear Regulatory Commission September 12, 2002 Page 2 Pursuant to 10 CFR 50.91, a copy of this proposed license amendment is being sent to the State of South Carolina. If there are any questions regarding this submittal, please contact Reene' Gambrell at (864) 885-3364. Very truly yours, W. R. McCollum, Oconee Nuclear President

U. S. Nuclear Regulatory Commission September 12, 2002 Page 3 cc: w/attachments 1 & 2(two copies) Mr. L. N. Olshan, Project Manager Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Mail Stop 0-14 H25 Washington, D. C. 20555 cc: w/attachments 1 & 2 Mr. L. A. Reyes, Regional Administrator U. S. Nuclear Regulatory Commission - Region II Atlanta Federal Center 61 Forsyth St., SW, Suite 23T85 Atlanta, Georgia 30303 Mr. M. C. Shannon Senior Resident Inspector Oconee Nuclear Station Mr. Virgil R. Autry, Director Division of Radioactive Waste Management Bureau of Land and Waste Management Department of Health & Environmental Control 2600 Bull Street

Columbia, SC 29201

U. S. Nuclear Regulatory Commission September 12, 2002 Page 4 W. R. McCollum, Jr., being duly sworn, states that he is Vice President, Oconee Nuclear Site, Duke Energy Corporation, that he is authorized on the part of said Company to sign and file with the U. S. Nuclear Regulatory Commission this revision to the Renewed Facility Operating License Nos. DPR-38, DPR-47, DPR-55; and that all the statements and matters set forth herein are true and correct to the best of his knowledge. W. R. McCollum, Jr., V e President Oconee Nuclear Site Subsred Notary. Publ ':My C Smmissi and sworn to before me this day of 2002 [ic -on Expires: -I

ATTACHMENT 1 Duke Energy Corporation Supplement to Response to Request for Additional Information Approval of Alternative Source Term Implementation (October 16, 2001)

U. S. Nuclear Regulatory Commission September 12, 2002 Page 1 I. Aerosol Deposition Duke Energy has evaluated the use of aerosol deposition (natural process) removal of iodine in post-accident containments. Sensitivity studies have shown that dose is not very sensitive to this removal process. Calculated doses without credit for aerosol deposition are only slightly increased (0.1 to 0.2 rem) above doses calculated using the median values from NUTREG/CR-6189. Further reducing the effectiveness of this removal method by using the 10% values from NUREG/CR-6189 provides only a small benefit; therefore, this level of modeling detail with conservatively low deposition rates is not warranted. The LOCA analysis for Oconee Nuclear Station has been revised to conservatively exclude credit for aerosol deposition removal in containment (both sprayed and unsprayed regions). The resulting offsite and control room doses are presented in the table below. All calculated doses remain within the regulatory limits prescribed in Regulatory Guide 1.183. See Attachment 2 of this supplement for changed pages to Duke Energy's response to the Staffs Request for Additional Information (RAI), dated May 20, 2002. LOCA Calculated Doses Containment RBES Model Total TEDE Model (rem TEDE) (rem) (rem TEDE) EAB 8.6 0.2 8.8 LPZ 1.6 0.1 1.7 Control Room 2.6 1 0.6 3.2 I

U. S. Nuclear Regulatory Commission September 12, 2002 Page 2 II. Particulate Spray Removal As Duke examined information provided on particulate iodine removal methods employed in our LOCA containment model, we elected to provide additional information on these methods to supplement our LAR RAI submittal. The following information supplements Duke's response to RAI 7 and 8. Particulate fission products, including aerosol particle forms of iodine, are effectively removed by containment sprays through several mechanisms including Brownian diffusion, diffusiophoresis, interception, and inertial impaction. Estimates of particulate washout are obtained using NUREG/CR-0009 Section 5.3.1, and SRP 6.5.2, Section llI.4.c(4) methodology as follows: Xsp = 3hFt (3.048 ft-) for 0.02 _ C/C -< 1.0 (Eq. 8) 2V

3hF, t

XSp = -h-(0.3048 ) for C/Co < 0.02 (Eq. 9) 2V where: C/CO = Ratio of particulate concentration at time t to the initial concentration at time zero. h = Drop fall height, ft. Ft= Spray flow rate during time step t, ft3/hr. V = Volume of contained gas phase, ft3. Using the equations above along with sprayed containment volume, spray flow rate, and drop height information presented in Xse calculations above, particulate iodine spray removal rate constants are calculated for each post-LOCA time interval. A numerical example is presented below. Time interval: 0 to 96 seconds No spray washout credit is assumed prior to initiation of containment spray flow. Containment' spray flow is assumed to commence 96 seconds after accident initiation. Therefore, the particulate iodine spray removal rate constant, Xsp,, for this time interval is zero: I-sp = 0.0 hf Time interval: injection phase (after 96 seconds prior to recirculation phase)

U. S. Nuclear Regulatory Commission September 12, 2002 Page 3 The particulate iodine spray rerhoval rate constant, X.p, in this time interval is calculated using (Eq. 8) corresponding to the higher removal efficiency presented above. = 3(82ft)(22410ft3 / hr) (3.048 ft1 ) 9.70hr1 2(866,002ft3 ) Time interval: recirculation phase Although SRP 6.5.2, Section llI.4.d guidance indicates that there is no requirement to "cut-off' particulate iodine spray credit, the spray removal rate methodology described above requires a transition to a lower efficiency when the sprayed region particulate concentration reaches 2% of the initial value. In order to estimate the time at which the transition should occur, an expression for sprayed region concentration as a function of time is derived from the general first-order removal process differential equation presented in Section 3.0 of this calculation as follows: dt Revising the equation to indicate a particulate species removal rate problem with only spray removal mechanisms being considered, the general equation is re-written as follows: dCA S dt where: Xýsp = Particulate iodine spray removal rate constant, hfr CP = Time dependent concentration of particulate iodine, ptCi/mil. (Note that a negative sign is added preceding the removal rate constant, Xsp, consistent with the practice of stating this parameter as a positive value.) Rearranging and integrating: fdCp = f .2APCpdt Assuming an initial concentration of C. at time t=O, the solution to the integral is: Cp = C oe- 'APt

U. S. Nuclear Regulatory Commission September 12, 2002 Page 4 Rearranging to solve for time:" t =-In (Cp/Co) / Xsp hr The time at which sprayed region achieves 2% of the initial concentration, C0, is calculated as follows: t = - In (0.02) / 9.70 hr"< = 0.4 hr The injection phase (time interval after 96 seconds and prior to recirculation phase) ends at 25 minutes following initiation of the LOCA event. This implies that the transition to a lower particulate removal efficiency should occur during this time interval. However, mixing between the sprayed and unsprayed regions causes particulate iodine to be removed at a slower rate from the sprayed region than the XP value indicates, because iodine is also being added from the unsprayed region. The time dependent models show that the particulate iodine concentration does not reach 2% of the initial concentration during the injection phase. The transition to a lower particulate removal efficiency occurs during the recirculation phase, at approximately 3.5 hours post-accident. The particulate iodine spray removal rate constant, X-P, early in the recirculation phase is calculated using (Eq. 8) corresponding to the higher removal efficiency presented above. = 3(82ft)(15550ft3 /hr) (3.048 ft-) 6.73 hr4 2(866,002ft3) The lower removal efficiency to be applied in the later stages of the recirculation phase is calculated as follows: = 3(82ft)(15550ft3 / hr) (0.3048 ft 0.673 hf 2(866,002ft 3)

U. S. Nuclear Regulatory Commission September 12, 2002 Page 5 IlI.Elemental Iodine Spray Removal As Duke examined information provided on elemental iodine removal methods employed in our LOCA containment model, we elected to provide additional information on these methods to supplement our LAR RAI submittal. The following information supplements Duke's response to RAI 7 and 8. Numerical Calculation for Elemental Iodine Spray Removal Constant (Injection Phase) A sample calculation for elemental iodine spray removal during the injection phase is detailed in this section. Air-steam mixture property data is extracted for 290' F and a pressure of 58 psia. This condition corresponds roughly to a maximum upper containment temperature and pressure condition over the time period which sprays are expected to be effective. BNWL-1326 air-steam mixture property data required for the calculation of elemental iodine removal efficiencies and Xse values are summarized in the table below as follows: BNWL-1326 Property Table: 2900 F, 58 psia Parameter Value Fl 1.2601E-04 F2 1306.69 ft-micror/min DIMIX 0.149 ft2/hr SCIMIX 1.725 Spray Flow Rate and Instantaneous Elemental Iodine Absorption Coefficient; d, F, hd, and V: d = 0.065 cm F = 2795 gpm = 22410 ft3/hr hd = 82 ft V = 866,002 ft3 Determine Reynolds Number, Re: fDRe 2 = Fl*(d*104)3 = 1.2601E-04*(0.065 cm

  • 104)3

= 34605.5 BNWL-1326, p. 13 relationship between fDRe2 and Re: 6.477Re 1 609 = fDRe2 = 34605.5

U. S. Nuclear Regulatory Commission September 12, 2002 Page 6 => Re = (34605.5/6.477)(1/1609) = 207.4 Determine Terminal Velocity, U: U Re

  • F2 d *104 U = (207.4*1306.69 ft-micron/min)/0.065 cm
  • 10 4 micron/cm U = 416.9 ft/min Determine drop exposure time, te:

t'd *60 U t, (82 ft

  • 60 sec/min)/416.9 ft/min te = 11.80 sec.

Determine gaseous mass transfer coefficient for iodine, kg from (Eq. 5): kg DIMIX [2+0.6Reo°" SCIMIXO03] d* 0.0328 kg = 0.149ft2Thr [2+ 0.6(207.4)05 (1.725)033] 6.5E - 02cm

  • 0.0328ft / cm kg = 862.7 ft/hr Determine elemental iodine spray removal constant, Xse:

Elemental iodine spray removal constant is calculated using the SRP 6.5.2 first order approximation of the Well Mixed Model (Eq. 4) as follows: 5.08E-02kgte F Vd where: 5.08E-02 = 6

  • 8.47E-03 cm-hr/ft-sec

U. S. Nuclear Regulatory Commission September 12, 2002 Page 7 5.08E - 02cm

  • hr/ft
  • sec(862.7ft / hr)(1 1.8 sec)(22410ft3Ahr)

SA,, (866002ft3 )(6.5E - 02cm) se= 205.9 hfr1 The calculated X,,e value of 205.9 per hour satisfies the requirements stated in SRP 6.5.2 for direct application of the first-order approximation of the Well Mixed Model (i.e., as stipulated on p. 10 and 11 of SRP 6.5.2, "the expression is valid for Xse values equal to or greater than ten per hour"). However, the SRP 6.5.2 method limits Xe to < 20 per hour "to prevent extrapolation beyond the existing data for boric acid solutions with a pH of 5" (SRP 6.5.2, p. 11). The iodine spray removal constant for the injection phase (after 96 seconds) is therefore limited below the calculated value of 205.9 per hour to 20 per hour.

U. S. Nuclear Regulatory Commission September 12, 2002 Page 8 Numerical Calculation for Elemental Iodine Spray Removal Constant (Recirculation Phase) The re-volatilization analyses require the calculation of physical property values and the solution of mass balance equations. Each point in time generally requires a new set calculations and solutions to be determined. Thus, the bulk of the calculations performed to support the analyses are accomplished through the use of spreadsheets. Physical Properties ofAir-Steam and Water The calculation of physical properties such as density, viscosity, diffusivity, and vapor pressure are required in order to quantify the mass transfer process. The required physical properties are determined from correlations found in ORNL-TM-1911, "Removal of Elemental Iodine From Steam-Air Atmosphere By Reactive Sprays," and other published literature. Iodine Partitioning The partitioning of elemental iodine between the gaseous and liquid phases at equilibrium is given by the following equation: log PC(I 2) = 6.29 - 0.0149* TK where PC(I2) is measure of elemental iodine volatility represented by the ratio of elemental iodine concentration in the aqueous phase to that in the gaseous phase: PC(12) = [I2]aq / [12]gas The elemental iodine partition coefficient is also represented alternatively by the symbol H and will be used synonymously in this calculation: H = PC(I2) = 1 0 (629-00149T*) The overall mass transfer coefficient is then determined by the following equation: 1I /K,= 1 / kc + 1 / HkL Radiolytic Formation of Elemental and Particulate Iodine NUREG/CR-5950 gives the following relationship between particulate (I-) and elemental (I2) iodine at equilibrium due to radiolysis of water: [H1+]2 * [1-] 2 / ['2] = d + e*[H+] where [H], [f], and [U2] are the individual concentrations expressed as moles per liter. The constants d and e are given as:

4 U. S. Nuclear Regulatory Commission September 12, 2002 Page 9 d = 6.05E-14 e = 1.47E-09 From the pH of the sump, the concentration of the H+ ion is determined by [HI-] = 10"pH The total iodine present in aqueous solution, [I]aq is determined by the concentrations of F and 12 present in the solution: [I]aq = [I] + 2 [I2]aq Rearranging gives: [I2]aq = (/2 [,]aq - [I]1) [H+]2[1] 2/[12]aq = d + e[HF] [H+]2[r] 2 [12]aq {d + e[II]} [HI[I-I] 2 [ 1/2 ( [I]aq - [11 )1 {d + e[1Ir]} [H+]2[112.= 1/2 {d + e[H+]}*[I]aq - /2 {d + e[H+] }[I] [H+]2[1]2 + 1/{2d + e[H+] } [I] - /21{d + e[H+] }[I]aq = 0 [1'] in the above equation can then be determined from the real root of the quadratic equation: [I-] = [-_3 + (p 2-4ccy) 05]/(2oc) where: a = [11+]2 = (1 0-PH)2 S= 1/2{d + e[H+] }= 1/2(d + e*10"PH) y = -/2{d + e[H+] } [Ilaq = -1/2(d + e*10"P)[I]aq Thus, the required inputs for determining the concentrations of elemental and particulate iodine are pH and total amount iodine present in solution. Inputs from the post-accident sump pH analyses and spray washout analyses are used for the determination of sump total iodine concentration and pH. Calculations are performed through spreadsheets use the SUMP Visual Basic program developed to automate these calculations.

U. S. Nuclear Regulatory Commission September 12, 2002 Page1 Time Dependent Re-volatilization Analysis For each of the time dependent re-volatization analyses, the containment compartment will be defined by the following distinct control volumes: "* the containment gas volume in the unsprayed region of the atmosphere; "* the containment gas volume within the sprayed region of the atmosphere; and "* the sump liquid volume Mass must be conserved such that the rate of accumulation in each control volume will be equal to the net rate of input into the control volume plus the rate of generation within the control volume: riput-routput + rgeneration - raccumulation The change in the control volume mass can be determined by integrating over time: t2 t2 An = Jraccumu.ationdt = f(rinput - routput + rgeneration )dt tl tl Thus, the integral can be evaluated as a summation over a sufficiently small time interval: An = Z(r (t,)- rol (t,) + rgn (t,) )t,1+ - t,) i=1

1.

Particulate Iodine Mass Balance The removal process for particulate iodine follows a basic first-order spray removal process. The removal rate equation is defined as follows from Equation 8.3.3-2 of ANSIIANS-56.5: dc, m n Qj, C + n Q dt = XZ 'k,c C- _ I c-C where: c, = concentration of iodine in volume i cj = concentration of iodine in volume j m = number of removal processes in volume i Qj = transfer rate to volume i from volume j = 2 volumes of unsprayed region per hour Qji = transfer rate to volume j from volume i = 2 volumes of unsprayed region per hour Vi = volume of it compartment Xki = removal coefficient of the kt removal process in volume i = Xse and X-sp values for elemental and particulate iodine

U. S. Nuclear Regulatory Commission September 12, 2002 Page 11 Thus for the re-volatilization analyses, the number of removal processes for particulate iodine will be one (i.e., removal by spray) and the number of compartments is two (2) (the sprayed region is one compartment and the unsprayed regions collectively is the other). The mass balance equations developed above assume no particulate iodine releases into the containment atmosphere after the initial instantaneous release at t= 0. Since iodine is assumed to be released gradually over time under the Regulatory Guide 1.183 Alternative Source Term assumptions, the mass balance equations developed above must be modified as follows to account for iodine releases (addition) from the core: 0.95

  • R
  • V'pr"e' dCpart.sprayed

(,,prayed + Vunsprayed) Q + a dt 126.9

  • VKprayed

'partpart'sprayed Vsprayed Cpaflpsprayed - v-"prayed part unsprayed 0.95

  • R* V..prayed dCpart,unsprayed

( (syed unsp+ VQusparted ) + p a dt 126.9

  • Vunsprayed Vunsprayed C partunsprayed Vunsprayed C part sprayed
where, R

= Release rate of iodine from the core, grams per second 0.95 = Fraction of iodine released as particulate iodine 126.9 = Atomic weight of iodine, grams per mole For a small increment in time At, the change in the particulate iodine concentration is determined as: 0.95

  • R
  • Vsprayed Acpart.sprayed (Vsprayed + Vunsprayed)

Q a At 126.9

  • V~prayed 2 partCpart'sprayed Vsprayed Cpart'sprayed V'prayed 0.95
  • R
  • V.nprayd Acpariunsprayed = (Vsprayed + Vunsprayed)

Q -I IQ At 126.9

  • Vuprayed Vunsprayed cpartunsprayed Cparisprayed By multiplying the concentration by the appropriate volume, the change in mass (moles) over time At can be determined for both the sprayed and unsprayed regions by the following equations:

U. S. Nuclear Regulatory Commission September 12, 2002 Page 12 F.95*R*V-'Prayed __(Vspae +Vusryd Anfpartsprayed =126.9 Apart CpartsprayedVsprayed - Qcpart,.sprayed + QCpart.,uprayed 0.95

  • R
  • Vnsprayed Anfpartrunpsrayed "-

(V6praaed +Vunsprayed QCaunsprayed + Qcpart'sprayed At S 126.91 Elemental Iodine Mass Balance During post-accident sump recirculation, the containment sump liquid will be recirculated through the containment atmosphere by the Building Spray (BS) System. The direction of the concentration gradient between the spray drops and the sprayed volume will dictate whether elemental iodine absorption or re-volatilization occurs. Current spray washout models which assume a first order removal process for elemental iodine are valid only when there is no source term by which iodine is added back to the atmosphere. However, since the sump can be a significant source of iodine volatility in the presence of radiation the possibility exists for iodine re suspension or re-volatilization to occur. To account for this possibility, the re volatilization analyses assumes that a state of equilibrium is achieved between the gas and liquid phases as the spray drops are exposed to the containment sprayed gas volume. Under this assumption, the concentration of elemental iodine in the liquid phase and the concentration in the gas phase are related by the elemental iodine partition coefficient (H). Thus, as the spray drops fall through containment atmosphere, the final elemental iodine concentration in the spray droplet is determined as follows: CI,.,d.op (t,) = H1cetemsprayed (td) Since spray is recirculated from the containment sump liquid, the spray drops enter the containment atmosphere with an initial elemental iodine concentration equal to the concentration present in the sump liquid. Thus, the rate of absorption or re volatilization of elemental iodine, due to spray action, can be determined from the spray flow rate and the net change in drop concentration: dnelemspray -action = -Fspray (HCelem sprayed (ti) - Celem,sup (t dt

U. S. Nuclear Regulatory Commission September 12, 2002 Page 13 The net change in mass over a small time increment from t, to t,+, can determined as follows: An,,m.,spray_ acon = Fspray (IHCeen,sprayed(te) - Ceern,sump(t,)*t,+ -- t,) In addition, elemental iodine will also be exchanged between the sprayed and unsprayed regions due to mixing in the containment atmosphere. Using the equation previously developed for particulate iodine, a similar expression can be developed to determine the change in mass due to mixing: Anetem,mmmng Q Ctern,sprayed + Q Celem,unsprayed (t)+1 prayed k V sprayed sprayed J Thus net transfer rate in the sprayed region for elemental iodine from spray action and containment mixing is: Anetem,sprayed "- Afeleer,spray -action + Anletem,rining = Fpray(Hcele..,sprayed(t,) CetemSUmp(t)*t,+i t,) "+- Qsraye Celemsprayed + Celem,unsprayed t sprayed Vrayed prayed For the unsprayed region, the change in elemental iodine mass will be determined by an equation similar to the one developed for particulate iodine: A'eer~n QQ Cnsprayed (t+ 1 - t, )Vunsprayeei lelen,unsprayed "" pd Cunsprayedd Vuprayed Similarly, the follow equation must be satisfied to for the containment elemental iodine mass be conserved in a closed system: Anelem,sprayed +Anelem,unsprayed + Anelemsump = 0 Thus, the net change in elemental iodine in the sump is determined as follows: Anedem,sump z- -(Anelem,sprayed +Anelem,unsprayed) Similar to the previous discussions, the elemental iodine mass balance equations are calculated by the following equations:

U. S. Nuclear Regulatory Commission September 12, 2002 Anelem,sprayed Anelem,unsprayed =

where, R

0.0485 126.9 2 = Release rate of iodine from the core, grams per second = Fraction of iodine released as elemental iodine = Atomic weight of iodine, grams per g-atom = number of gram atoms of iodine per mole of elemental iodine Page 14 Using the removal rate and mass balance expressions developed here, the time dependent re-volatilization analyses are performed by calculating the mass and concentrations over sufficiently small time increments. Determination of Recirculation Phase Spray Lambdas for Use in LOCADOSE: For the recirculation phase, the values for spray lambda entered into LOCADOSE were varied in order to produce node inventories matching the results from the NUREG/CR-5950 methodology described above. method are presented below: The resulting values calculated from this trial-and-error Start Time End Time Elemental Iodine (hours) (hours) Spray Lambda (hr"1) 2.667E-02 0.4167 20.0 0.4167 1.8 0.0 1.8 8.0 0.06 8.0 13.8 0.09 13.8 24 0.13 24 96 0.071 96 112.8 0.002 112.8 720 0.0 0.0485

  • R
  • Vprayed (Vsprayed + Vu.prayed)

Fspray (nCetem sprayed(t) Ceem'sump (t) (t'+1 - t) 2

  • 126.9

+ Qa Celemspraed + Q Celem,unsprayed (ti ti)Vayed + 'rae V )y 0.0485

  • R
  • Vsprayed (Vprayed + Vunsprayed )

Q + Q 2 r 126.9, vuspayed Vunrsprayed Vunsprayed - t, )Vunsprayed

U. S. Nuclear Regulatory Commission September 12, 2002 Page 15 IV. BWST Model Additional information on the method by which the volatile form of iodine (12) is calculated to be accumulated in the BWST and released to the airspace and the environment is presented below. The following information supplements Duke's response to RAI 11. The leakage inlet point into the BWST is through backleakage from the ECCS piping, which is located at bottom of the BWST. The maximum capacity of the BWST is 390,000 gallons. BWST liquid volume and pH are time dependent variables in the model. BWST Concentrations The RB sump solutions conditions are taken from the sump pH calculation for the case that has the earliest swapover to recirculation mode (Case III). The relevant conditions include temperature and the concentrations (in ppm) of boron, lithium, sodium, nitrate, phosphate and chloride. These conditions are assumed to follow the time dependent profiles calculated in the sump pH analysis. From the sump temperature and concentrations, these parameters can be determined for the BWST based on the sump solution backleakage through the ECCS. The density of the BWST solution is (0.01615 ft3/lb) 1 x (0.03531 ft3 / liter) = 2.19 lb/liter. The density of the sump is constant at (0.017174 ft3/lb)"1 x (0.03531 ft3 / liter) = 2.06 lb/liter. The time dependent BWST temperature is discussed in detail in the next section. The time dependent BWST concentrations of each of the components (designated generically as i) contained in the sump solution can be determined by the following equation. Ci, t = [(Ci, to* VB, to

  • 2.19) + (Leak Rate * (t - to)
  • Cis, to
  • 2.06)] / (VB, t
  • 2.19) where:

C, t -" BWST concentration of component i at time t (ppm) (may be substituted by the following for each component): CB = Boron concentration (ppm) C11 = Lithium concentration (ppm) CNa = Sodium concentration (ppm) Cph = Phosphate concentration (ppm) CNO3 = Nitrate concentration (ppm) Cc, = Chloride concentration (ppm) Qs, t= Sump concentration of component i at time t (ppm) (may be substituted using the above notation for each component) VB, t = Volume of BWST at time t (liter)

U. S. Nuclear Regulatory Commission September 12, 2002 Page 16 Leak Rate = Rate of ECCS backleakage to BWST (liter/sec) t = Time (sec) to = Time interval before t (sec) 2.19 = Density of the BWST (lb / liter) 2.06 = Density of the Sump (lb / liter) where VB, t is calculated for each new time step as the volume at the previous time step, plus that added from the ECCS backleakage and BWST refill during the time interval. The volumes added are determined from the backleakage rate or refill rate multiplied by the time interval. BWST Temperature The bulk temperature of the BWST will rise as Reactor Building (RB) sump water is added to the existing inventory. The temperature of the solution in the BWST may be calculated by considering the change in its thermal energy. Let MB(t) and TB(t) be the mass and temperature of the solution in the BWST, respectively. Let Ps and Ts be the density and temperature of the solution in the containment sump. Assume that the specific heat (c,) of the water in the BWST and containment sump is constant at lBtu/(0F-lbm). With no heat transfer to the environment, the rate of change of the energy in the BWST is equal to the rate of energy transported by the ECCS backleakage from the containment sump, where q is the backleakage rate. That is, cp[MB(t)T (t)]" = qpscpTs The mass of the solution in the BWST, initially set to MBO, increases with the addition of water from the sump. Therefore, the mass of the solution in the BWST is given by MB(t) = MBO + qps(t - to) Assuming the sump temperature stays relatively constant over a small incremental, change in time for simplification, the above equation may be simplified to yield [MB(t)TB(t)]" = qpsTs This equation may be solved by direct integration. The initial temperature of the solution in the BWST is TBO. Therefore, the lower limit of integration at to is MBOTBO. Integration yields MB(t)TB(t) - MBOTBO = qpsTs(t - to)

U. S. Nuclear Regulatory Commission September 12, 2002 Page 17 From this, the temperatureo1f the solution in the BWST is found to be TB(t= BoTBO JqTos,(t -to) "MBO + qps(t-to) Similarly, for cases with BWST refill water at a constant temperature (TR) and density (PR) also entering the BWST at a refill rate of q2, the temperature of the BWST solution is found to be TB(t)= MBOTBO + qTsPs(t-to)+ q2TRpR(t-to) M BO + qps(t-to)+ q2p(t -to) The maximum BWST temperature reached is 84.60C. This is well below the boiling point, so boiling of the BWST solution is not a factor in release to the environment. This calculation ignores any heat loss out of the BWST, which is conservative. BWST pH The pH of the solution in the BWST at any time is determined using an EXCEL/Visual Basic program (PHSC) developed by Duke Power Nuclear Chemistry. After the concentrations of boron, sodium, lithium, N0 3, phosphate and CI are determined as a function of time, the pH of the BWST aqueous solution is calculated utilizing this program. The pH at actual BWST temperature (not reference temperature) is used in the calculations below based on NUREG/CR-5950 methodology. Iodine Concentrations in BWST Liquid and Vapor The amount of iodine transferred to the BWST from the ECCS backleakage is based on the concentration of iodine in the sump. This value is determined by assuming all the iodine released to containment is deposited into the sump. The sump concentration changes with time according to the release rates and timing provided in Regulatory Guide 1.183. Iodine is then transferred to the BWST according to the transfer rate by the following equation: C1, t -" [M, t-1 + Leak Rate * (t - to)

  • C1s] / VB, t where:

MI, t= Amount of iodine in BWST at time t (mol) Leak Rate = Rate of ECCS backleakage to BWST (liter/sec) t = time (sec) CIs = Iodine concentration in sump (mol/liter) VB, t = Volume of BWST at time t (liter)

U. S. Nuclear Regulatory Commission September 12, 2002 Page 18 The mathematical model of Beahm, et al. (NUREG/CR-5950) is used to calculate the partitioning of iodine from a solution such as may be found in the BWST with ECCS backleakage. The mathematical model itself is based on the analysis of the reactions of various iodine groups in an irradiated electrolytic solution representative of what might be found in the containment sump or BWST following a postulated MI-A. It was noted that volatile iodine species could form in an irradiated solution, particularly if the solution is acidic (has low pH). It is only the volatile species that will be partitioned; iodide ions will remain in solution. The dominant volatile iodine species is taken to be elemental iodine. Therefore, the iodine in the airspace of the BWST will be predominantly 12. Beahm, et al. state that the formation of 12 in an irradiated electrolytic solution is most likely the result of the reaction of iodide in a post accident solution with hydrogen peroxide as follows: 2F + 2W + H 20 2 -> 12 + 2 H 2 0, and 12 + H 20 2 -> 2F + 2H+ + 02. It has been proposed that for these reactions, the concentrations of H+, 1-, and 12 may be related by the following equations: [H+]2[12 =_d+e[H+] and [I] = [1F] + 2[12]. where: [HI+] is the molar concentration of the hydrogen ion [I] is the total concentration (g-atoms per liter) of iodine in any form [I-] represents the molar concentration of the particulate iodine ions [12] is the molar concentration of elemental iodine Molar concentration is defined as moles/liter. In addition, d and e are constants associated with the reaction between 12 and F in an irradiated electrolytic solution such as may be found in the BWST (or containment sump) following a postulated MHA. NUREG/CR-5950 gives the following relationship between particulate (F) and elemental (12) iodine at equilibrium due to radiolysis of water: [HW] 2 * [1I]2 / [12] = d + e* [H+1

U. S. Nuclear Regulatory Commission September 12, 2002 Page 19 where [HI], [I-, and [12] are the individual concentrations expressed as moles per liter. The constants d and e are given as: d = 6.05E-14 e = 1.47E-09 From the pH of the sump, the concentration of the W ion is determined by [H+] = 10"PH The total iodine present in aqueous solution, [I]aq is determined by the concentrations of F and 12 present in the solution: [I]aq = [I-] + 2 [I2]aq Rearranging gives: [12]aq = 1/2 ( [I]aq - [I) [HW] 2[I'] 2 [I2 ]aq = d + e[H+] [1+] 2['] 2 L [12]aq {d + e[H+] I [H+]2 [-] 2 1/2 ( [I]aq - [1] )* {d + e[H+]} [fl] 2[I]2 = 1/2 {d + e[H+]}*[I]aq - 1/2 {d + e[H] }[r] [Hr]2[i-]2 + 1/2{d + e[H+] }[r] - 1/2{d + e[H+] } [I]aq = 0 [I] in the above equation can then be determined from the real root of the quadratic equation: [I-] = [-13 + (132-4cry')°-'/(2ct) where: CC = [HJ+] 2 = ( 1 0 "PH)2 0 = 1/2{d + e[H+] }= 1/2(d + e*10"pH) Y = -{/2d + e[H] } [I]aq = -1/2(d + e* 10PH)[l]aq Thus, the required input values for determining the concentrations of elemental and particulate iodine are pH and total amount iodine present in solution.

U. S. Nuclear Regulatory Commission September 12, 2002 Page 20 The partitioning of elemental iodine between the gas and liquld phases is given by the following equation (NUREG/CR-5950): PC(12) = [I2]aq / [I2]gas = 10[6 29-0 0149T(K)] Iodine Release Rates from BWST Vapor The iodine release rate from the BWST vapor can be determined by the volume of vapor displaced by the liquid volume additions from ECCS backleakage. The BWST release rate is determined by: Release Rate (mol/sec) = [I2]gas * (Leak Rate) [I2]gas = BWST Vapor Iodine Concentration (mol/liter) Leak Rate = ECCS Backleakage Rate (liter/sec) The release rate in grams per second can be determined by multiplying by the molecular weight of elemental iodine (126.9

  • 2 = 253.8 glmol).

For use in LOCADOSE modeling, the ratio of BWST Release Rate (g Iodine/sec) to ECCS Backleakage Rate (g Iodine/sec) is a useful parameter. ECCS Backleakage Rate is a constant for each case, and BWST Release Rate varies with time. The table below gives the time-weighted average ratio of BWST release rate (g Iodine/sec) to ECCS backleakage rate (g lodine/sec) at various time intervals that will be used for input into LOCADOSE modeling. Average Ratio of Release Rate to ECCS Backleakage Rate Time Intervals 0 min - 25 25 min - 2 hr 2 hr - 8 hr 8 hr - 24 hr '24 hr-1 1 day - 30 min day days 0 4.05E-08 1.02E-06 7.39E-06 1.77E-05 2.58E-05 The average release rates calculated above were modeled by treating the BWST release to the environment as a filtered one with filter efficiencies set to one minus the released ratio. An example of the calculation is as follows for 2 hours to 8 hours: (i-1.02e- )* 100 = 99.99990% filter efficiency [

U. S. Nuclear Regulatory Commission September 12, 2002 Page 21 V. Bounding Fuel Handlintg Accidents The following clarifications supplement Duke's response to RAI 13 and 14. The bounding single assembly fuel handling accident is the fuel assembly accident in the Unit 1 and 2 SFP to the Unit 1 and 2 control room air intake. The bounding cask drop (multiple assembly) accident is the transport cask drop accident in the Unit 1 and 2 SFP to the Unit 1 and 2 control room air intake. VI. Clarification of Fuel Handling Accident Cases Evaluated The following information supplements Duke's response to RAI 18. See Attachment 2 of this supplement for changed pages to Duke Energy's response to the Staff's Request for Additional Information (RAI), dated May 20, 2002. The accident cases shown in the Table on page 68 of the Response to RAI were chosen to evaluate the range of scenarios postulated to occur during fuel handling operations. It is postulated that a single fuel assembly could be dropped either in the SFP or inside containment, and that a transport cask or ISFSI cask could be dropped into a SFP and damage multiple assemblies in the pool. For single assembly accidents (Cases 1 through 6) the event could occur in either of the two SFPs or in any of the containments, and the source term is the same. All assume the entire gap of one assembly is released. For transport cask and ISFSI drop accidents (Cases 7 through 12), the accident scenarios are specific to a particular SFP. This is due to different number of assemblies contained in each pool that are postulated to be damaged by the cask drop; therefore, the source term is Unit specific. The Table on page 68 has been revised to reflect this information for clarity. For a fuel handling accident occuring in a SFP, the release of radioactivity could occur either through the Unit Vent of a particular unit, or through the SFP Roll-up Doors. To narrow the number of cases that needed to be evaluated, and to ensure conservative results, ARCON96 was used to calculate X/Q values (atmospheric dispersion values) for all combinations of release points and intake locations. These are shown in the table below. The maximum bounding set of X/Q values for a particular release type (e.g., unit vent, roll-up door, etc.), reduced by a factor of 2 for dual CR intake credit, was used in the dose calculations. This allows the calculations to bound an event occuring in any of Oconee's three units.

U. S. Nuclear Regulatory Commission September 12, 2002 Page 22 Summary of Oconee Nuclear Station X/Q Values Example Case Designations: VENT1 NE Unit 1 vent releases to NE control room air intake VENT2NE Unit 2 vent releases to NE control room air intake VENT3NE Unit 3 vent releases to NE control room air intake VENTISE Unit 1 vent releases to SE control room air intake VENT2SE Unit 2 vent releases to SE control room air intake VENT3SE Unit 3 vent releases to SE control room air intake Vent Releases Reduced by Maximum a Factor of 2 X/Q for Dual Intakes VENT1NE VENT2NE VENT3NE VENT1SE VENT2SE VENT3SE 0 to 2 hr 8.21 E-04 3.45E-04 1.44E-04 1.58E-04 3.02E-04 8.70E-04 8.70E-04 4.35E-04 0 to 8 hr 6.69E-04 2.49E-04 9.68E-05 1.22E-04 2.18E-04 6.23E-04 6.69E-04 3.35E-04 8 to 24 hr 2.54E-04 7.92E-05 2.98E-05 3.50E-05 6.35E-05 2.20E-04 2.54E-04 1.27E-04 1 to 4 days 1.99E-04 7.04E-05 2.50E-05 3.39E-05 6.02E-05 1.65E-04 1.99E-04 9.95E-05 4 to 30 days 1.61 E-04 5.81 E-05 2.22E-05 2.83E-05 4.70E-05 1.29E-04 1.61 E-04 8.05E-05 Fuel Handling Building Roll-up Door Releases FHB12NE FHB3NE FHB12SE FHB3SE 0 to 2 hr 2.88E-04 1.39E-04 1.32E-04 2.71 E-04 2.88E-04 1..44E-04 0 to 8 hr 2.44E-04 1.07E-04 1.07E-04 2.13E-04 2.44E-04 1.22E-04 8 to 24 hr 9.70E-05 3.89E-05 3.15E-05 7.31 E-05 9.70E-05 4.85E-05 1 to 4 days 7.45E-05 3.03E-05 3.05E-05 6.27E-05 7.45E-05 3.73E-05 4 to 30 days 5.66E-05 2.50E-05 2.33E-05 5.30E-05 5.66E-05 2.83E-05

U. S. Nuclear Regulatory Commission September 12, 2002 Page 23 VII. Clarification of WO/ Parameters The following information supplements XIQ values presented on Page 23 of Duke's LAR submittal. See Attachment 2 of this supplement for changed pages to Duke Energy's License Amendment Request, dated October 16, 2001. The values for atmospheric dispersion factors (X/Qs) shown in the table in page 23 of the ONS LAR submittal are the bounding maximum X/Qs for the limiting combination of a particular release point type and either control room intake. For example, for the unit vent releases, we evaluated releases from Unit 1, 2 and 3 unit vents to each of the two control room intake locations. This yielded six different X/Q values for a unit vent release (see table above supplementing RAI 18). We then chose the limiting (maximum) X/Q value to apply to any unit vent release scenario. This reduces the number of LOCADOSE runs that need to be performed, and ensures the most conservative dose result, regardless of which ONS Unit experiences the postulated event.

ATTACHMENT 2 Duke Energy Corporation Revised Pages to License Amendment Request Approval of Alternative Source Term Implementation (October 16, 2001) Revised Pages (p. 23) and Response to Request for Additional Information Approval of Alternative Source Term Implementation (May 20, 2002) Revised Pages (p. 1, 2, 44, 45, 46, 68) 1

U. S. NRC October 16, 2001 Page 23 The bounding X/Q values below were determined by calculating a X/Q value for each pair of release points and control room intakes possible (i.e., six Unit Vent X/Qs were calculated for releases from all three unit vents to each of the two control room intakes). The maximum X/Q from these six values was chosen for the dose analysis to ensure bounding results regardless of which unit is postulated to experience the fuel handling event. Parameter Bounding Value Factor from X/Q of 2 Reduction Calculation Unit Vent Atmospheric Dispersion 8.70E-04 4.35E-04 Factor-X/Q (0-2 hour) sec/m3 sec/m3 Equipment Hatch Atmospheric Dispersion 6.35E-04 3.18E-04 Factor-X/Q (0-2 hour) sec/mr3 sec/m3 SFP Roll-up Door Atmospheric Dispersion 2.88E-04 1.44E-04 Factor-X/Q (0-2 hour) sec/m3 sec/m3 In conclusion, the TEDEs to Control Room operators are summarized in the table below for each case described using the AST. The alternative source term TEDE is calculated in accordance with guidance provided in Regulatory Guide 1.183. According to the table below, the maximum TEDE calculated for either Control Room following a fuel handling event is 2.0 rem. The event is the transport cask drop in the Unit 1&2 spent fuel pool with transport from the Unit 2 vent to the Unit 1 & 2 Control Room. This value satisfies the 5-rem occupational TEDE limit.

pt .1 U. S. Nuclear Regulatory Commission April 30, 2002 Page 1 Duke Energy Corporation Response to Request For Additional Information Approval of Alternative Source Term Implementation Introduction Duke personnel met with NRC staff on March 21, 2002 to discuss proposed responses to the Request for Additional Information. Based on the discussions at this meeting, both the Loss of Coolant Accident (LOCA) and Fuel Handling Accident (FHA) analyses have been revised to incorporate improvements and conservative simplifications in features and input parameters. The LOCA analysis for Oconee Nuclear Station has been revised to conservatively exclude credit for aerosol deposition removal in containment (both sprayed and unsprayed regions). Duke also agreed to revise the FHA analysis to credit an overall effective Decontamination Factor (DF) of 200, instead of an elemental DF of 500 and organic DF of 1. LOCADOSE Version 6.0 was used inthe revision of these analyses. This version provides improved logic and treatment of a calculation using the ONS model input features. Version 6.0 of the code adds some modeling techniques to Version 5.0 that are especially suited to the ONS Alternative Source Term (AST) application. The calculated LOCA and FHA doses are shown in the tables below. All calculated doses remain within the regulatory limits prescribed in Regulatory Guide 1.183. LOCA Calculated Doses Containment Model RBES Model Total TEDE (rem TEDE) (rem TEDE) (rem) EAB 8.6 0.2 8.8 LPZ 1.6 0.1 1.7 Control Room 2.6 0.6 3.2

U. S. Nuclear Regulatory Commission April 30, 2002 .1,, Calculated Doses to Control Room Operators due to Fuel Handling Events Spent Fuel Pool (SFP) and Containment Case Group Source Release Control Room Unit TEDE Point Destination (rem) 1 1 Fuel Assembly Accident in Any Unit Unit 1&2 1.8 Either SFP Vent 2 1 Fuel Assembly Accident in Either Roll-Unit 1&2 0.6 Either SFP Up Door 3 1 Fuel Assembly Accident in Any Unit Unit 3 - 1.2 Either SFP Vent 4 1 Fuel Assembly Accident in Either Roll-Unit 3 0.4 Either SFP Up Door 5 2 Fuel Assembly Accident in Any Unit Unit 1 &2 1.0 Any Containment Vent 6 2 Fuel Assembly Accident in Any Unit Unit 3 0.7 Any Containment Vent 7 3 Transport Cask Drop in Any Unit Unit 1 &2 2.8 Unit 1&2 SFP VentI 8 3 Transport Cask Drop in Any Unit Unit 3 1.9 Unit 1&2 SFP Vent 9 3 ISFSI Cask Drop in Any Unit Unit 1&2 1.2 Unit 1&2 SFP Vent 10 3 ISFSI Cask Drop in Any Unit Unit 3 0.8 Unit 1&2 SFP Vent 11 3 ISFSI Cask Drop in Either Roll-Unit 1 &2 0.4 Unit 3 SFP Up Door 12 3 ISFSI Cask Drop in Either Roll-Unit 3 0.3 L Unit 3 SFP Up Door Calculated Offsite Doses due to Fuel Handling Events Spent Fuel Pool (SFP) and Containment Case Group Source Release Point EAB LPZ TEDE (rem) TEDE (rem) 1 1 Fuel Assembly Accident Any Unit Vent 1.2 0.1 in Either SFP 2 1 Fuel Assembly Accident Either Roll-Up 1.2 0.1 in Either SFP Door 5 2 Fuel Assembly Accident Any Unit Vent 0.7 0.1 in Any Containment 7 3 Transport Cask Drop in Any Unit Vent 1.2 0.2 Unit 1&2 SFP 9 3 ISFSI Cask Drop in Any Unit Vent 0.8 0.1 Unit 1&2 SFP 11 3 ISFSI Cask Drop in Either Roll-Up 0.8 0.1 Unit 3 SFP Door I I Page 2 I ý, -,ý,

U. S. Nuclear Regulatory Commission April 30, 2002 Page 44

7.

NRC Request In pages 15 and 16 of Attachment 3, you describe fission product release model from the containment. Provide the following additional information: "* Values and calculation of aerosol deposition (natural processes) rates in reactor building "* Calculation of particulate removal coefficients by containment spray and the bases for 25 minute turnover of the rate "* Time for reaching a particulate decontamination factor of 50 by containment spray "* Time for reaching an elemental iodine decontamination factor of 200 by containment spray "* Duration of the worst 2 hour period used in Exclusion Area Boundary (EAB) dose calculation

Response

Duke Energy has evaluated the use of aerosol deposition (natural process) removal of iodine in post-accident containments. Sensitivity studies have shown that dose is not very sensitive to this removal process. Calculated doses without credit for aerosol deposition are only slightly increased above doses calculated using the median values from NUREG/CR-6189. Further reducing the effectiveness of this removal method by using the 10% values from NUREG/CR-6189 provides only a small benefit; therefore, this level of modeling detail with conservatively low deposition rates is not warranted. Particulate fission products, including aerosol particle forms of iodine, are effectively removed by containment sprays through several mechanisms including Brownian diffusion,' diffusiophoresis, interception, and inertial impaction. Estimates of particulate washout are obtained using NUREG/CR 0009 (Reference 3) and SRP 6.5.2 (Reference 4) methodology as follows: 3hF1 Xsp = (3.048 ft-) for 0.02:< C/Co < 1.0 .sp = 3V (0.3048 ft-) for C/Co < 0.02

U. S. Nuclear Regulatory Commission April 30, 2002 Page 45 where: C/Co = Ratio of particulate concentration at time t to the initial concentration at time zero. h = Drop fall height, ft., Ft = Spray flow rate during time step t, ft3/hr. V = Volume of contained gas phase, ft3 The particulate iodine spray removal rate constant, Xsp, early in the recirculation phase is calculated corresponding to the higher removal efficiency presented above. The lower removal efficiency is applied in the later stages of the recirculation phase. At 25 minutes post-accident, the recirculation phase begins, and a new lambda is calculated based on the recirculation spray flowrate. "* The time for reaching a particulate decontamination factor of 50 by containment spray is approximately 3.5 hours post-accident. The elemental iodine decontamination factor does not reach 200 by containment spray in the calculation based on NUREG/CR-5950 methodology. See RAI 8 for description of calculations. The elemental iodine DF reached is approximately 82 when equilibrium elemental iodine concentration is reached. Spray removal is terminated at 4.7 days post accident, once equilibrium is reached. ' "* The maximum 2 hour dose for the EAB occurs during the time period from 0.6 to 2.6 hours following accident initiation.

U. S. Nuclear Regulatory Commission April 30, 2002 THIS PAGE INTENTIONALLY LEFT BLANK Page 46

4 k U. S. Nuclear Regulatory Commission April 30, 2002 Page 68 Calculated Doses to Control Room Operators due to Fuel Handling Events Spent Fulel Pool (SFP) and Containment Case Group Source Release Control Room Unit TEDE Point Destination (rem) 1 1 Fuel Assembly Accident in Any Unit Unit 1 &2 1.8 Either SFP Vent 2 1 Fuel Assembly Accident in Either Roll-Unit 1 &2 0.6 Either SFP Up Door 3 1 Fuel Assembly Accident in Any Unit Unit 3 1.2 Either SFP Vent 4 1 Fuel Assembly Accident in Either Roll-Unit 3 0.4 Either SFP Up Door 5 2 Fuel Assembly Accident in Any Unit Unit 1 &2 1.0 Any Containment Vent 6 2 Fuel Assembly Accident in, Any Unit Unit 3 0.7 Any Containment Vent 7 3 Transport Cask Drop in Any Unit Unit 1&2 2.8 Unit 1 &2 SFP Vent 8 3 Transport Cask Drop in Any Unit Unit 3 1.9 Unit 1&2 SFP Vent 9 3 ISFSI Cask Drop in Any Unit Unit 1&2 1.2 Unit 1&2 SFP Vent 10 3 ISFSI Cask Drop in Any Unit Unit 3 0.8 Unit 1&2 SFP Vent 11 3 ISFSI Cask Drop in Either Roll-Unit 1 &2, 0.4 Unit 3 SFP Up Door 12 3 ISFSI Cask Drop in Either Roll-Unit 3 0.3 Unit 3 SFP Up Door Calculated Offsite Doses due to Fuel Handling Events Spent Fuel Pool (SFP) and Containment Case Group Source Release Point EAB LPZ TEDE (rem) TEDE (rem) 1 1 Fuel Assembly Accident Any Unit Vent 1.2 0.1 in Either SFP 2 1 Fuel Assembly Accident Either Roll-Up 1.2 0.1 in Either SFP Door 5 2 Fuel Assembly Accident Any Unit Vent 0.7 0.1 in Any Containment 7 3 Transport Cask Drop in Any Unit Vent 1.2 0.2 Unit 1&2 SFP 9 3 ISFSI Cask Drop in Any Unit Vent 0.8 0.1 Unit 1&2 SFP 11 3 ISFSI Cask Drop in Either Roll-Up 0.8 0.1 Unit 3 SFP Door}}

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