Text

License Amendment Request to Modify the Licensing Basis Alternate Source Term
	ML16259A169
Person / Time
Site:	Robinson
Issue date:	09/14/2016
From:	Glover R; Duke Energy Progress
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	RNP-RA/16-0057
	Download: ML16259A169 (69)
	v • d • e

(_~ DUKE R. Michael Glover ENERGY~ H. B. Robinson Steam Electric Plant Unit 2 Site Vice President Duke Energy Progress 3581 West Entrance Road Hartsville, SC 29550 0 . 843 857 1701 F: 843 857 1319 Mike.Glovet@duke-energy.com 10 CFR 50.90 Serial: RNP-RA/16-0057 SEP _i *4 2016 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 H. B. ROBINSON STEAM ELECTRIC PLANT, UNIT NO. 2 DOCKET NO. 50-261/RENEWED LICENSE NO. DPR-23 LICENSE AMENDMENT REQUEST TO MODIFY THE LICENSING BASIS ALTERNATE SOURCE TERM

Dear Sir or Madam:

Pursuant to 10 CFR 50.90, Duke Energy Progress, Inc., hereby requests an amendment to the H. B. Robinson Steam Electric Plant, Unit No. 2 (HBRSEP) renewed facility operating license DPR-23. The proposed license amendment will modify the current HBRSEP current licensing basis to adopt a revised alternate source term. The revision is needed to support the transition from an 18-month to a 24-month fuel cycle. The amendment will require changes to the Technical Specifications Bases, the Update Final Safety Analysis Report, and the Technical Requirements Manual. A license amendment request to provide Technical Specifications changes for the extended fuel cycle will be submitted under a separate cover.

The Enclosure provides the basis for the proposed change, including a detailed description, technical and regulatory evaluations, environmental considerations, and the Duke Energy Progress, Inc. determination that the proposed change does not involve a significant hazards consideration. The proposed marked-up pages for the Technical Specifications Bases, the Update Final Safety Analysis Report, and the Technical Requirements Manual are provided in .

Approval of the proposed amendment is requested by August 30, 2018. Once approved, the amendment shall be implemented within 120 days.

This proposed change has been reviewed by the HBRSEP Plant Nuclear Safety Committee.

This letter contains no new Regulatory Commitments.

United States Nuclear Regulatory Commission Serial: RNP/RA/16-0057 Page 2 of 2 In accordance with 10 CFR 50.91(b), a copy of this application is being provided to the State of South Carolina.

If you should have any questions regarding this submittal, please contact Mr. Tony Pilo, Acting Manager - Regulatory Affairs at (843) 857-1409.

I declare under penalty of perjury that the foregoing is true and correct. Executed On: C\-1t.t-z.ou..

Sincerely, R. Michael Glover Site Vice President RMG/jk Enclosures c: NRC Administrator, Region II Mr. Dennis Galvin, NRC Project Manager, NRR NRC Resident Inspector, HBRSEP2 Ms. S. E. Jenkins, Manager, Infectious and Radioactive Waste Management Section (SC)

Attorney General (SC)

United States Nuclear Regulatory Commission Enclosure I to Serial: RNP-RA/16-0057 23 Pages (Including the cover sheet)

EVALUATION OF PROPOSED CHANGE 1.0

SUMMARY

DESCRIPTION 2.0 DETAILED DESCRIPTION

3.0 TECHNICAL EVALUATION

3.1 Source Term Analysis 3.2 Gap Release Analysis 3.3 Design Basis Accident Dose Consequences

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria 4.2 Precedent 4.3 Significant Hazards Consideration 4.4 Conclusions

5.0 ENVIRONMENTAL CONSIDERATION

6.0 REFERENCES

1.0

SUMMARY

DESCRIPTION This technical evaluation supports a request to amend the H. B. Robinson Steam Electric Plant Unit No. 2 (HBRSEP2) Renewed Facility Operating License DPR-23.

The proposed changes would revise the dose consequences for the facility as described in the Updated Final Safety Analysis Report (UFSAR) as follows:

1. transition from an 18-month cycle length source term to a 24-month cycle length source term

2. provide gap release fractions for high-burnup fuel rods (i.e., greater than 54 GWD/MTU) that exceed the 6.3 kW/ft linear heat generation rate (LHGR) limit detailed in Table 3 in Regulatory Guide 1.183 (Reference 1)

3. provide a new design basis accident dose analysis for the rod ejection scenario 2.0 DETAILED DESCRIPTION The primary purpose of this License Amendment Request (LAR) is to update the source term for HBRSEP2 to cover 18 to 24 month cycles. It also covers the removal of Part Length Shield Assemblies (PLSA) and the use of a fat pellet design. The development of fission product inventory is in accordance with Regulatory Guide 1.183 Section 3.1.

This LAR also proposes gap release fractions for high-burnup fuel rods (greater than 54 GWD/MTU) that exceed the 6.3 kW/ft LHGR limit in Footnote 11 of Table 3 in Regulatory Guide 1.183 (Non-LOCA Fraction of Fission Product Inventory in Gap), in order to support the cycle extension. Footnote 11 states:

As an alternative [to the non-LOCA gap fractions in Table 3 and the limits of Footnote 11], fission gas release calculations performed using NRC-approved methodologies may be considered on a case-by-case basis. To be acceptable, these calculations must use a projected power history that will bound the limiting projected plant-specific power history for the specific fuel load.

Based on the evaluation provided in Section 3.2, Duke Energy proposes to increase non-LOCA gap fractions for a maximum of 35 high-burnup fuel rods (greater than 54 GWD/MTU) in each fuel assembly that operates in the HBRSEP2 reactor. The increases are as follows:

The values in Regulatory Guide 1.183, Table 3 will be tripled for 85Kr, 134Cs, and 137Cs.

These increased gap fractions allow LHGRs up to 7.0 kW/ft for rod burnup above 54 GWD/MTU. Future fuel cycle designs for HBRSEP2 may include up to 35 fuel rods per fuel assembly operated at LHGRs up to the proposed limits.

The gap release analysis performed to support the higher LHGRs is described in detail in Section 3.2. The analysis calculated specific gap fractions in accordance with the method in the ANS 5.4 [2011] standard (Reference 4). Duke Energy has submitted a license amendment request related to the linear heat generation rate for Catawba, McGuire, and Oconee as well.

NRC approval was dated July 19, 2016. (Reference 14).

A new design basis rod ejection accident dose analysis is developed with an assumption of 10%

of rods in the core experiencing Departure from Nucleate Boiling (DNB). This analysis is developed in accordance with Appendix H of Regulatory Guide 1.183. The analysis is described in detail in Section 3.3.6.

The changes proposed in this LAR would be reflected in updates to the HBRSEP2 UFSAR.

Table 1 shows the pertinent accidents covered by the dose analysis, and their related UFSAR sections. The results of calculations of the dose consequences for these accidents are provided in Section 3.3.

Table 1. List of Impacted Accidents Accident UFSAR Sections Main Steam Line Break 15.1.5 Locked Rotor 15.3.2 Single Rod Withdrawal 15.4.3 Rod Ejection 15.4.8 Loss of Coolant Accident 15.6.5 Fuel Handling Accident 15.7.4

1.0 TECHNICAL EVALUATION

The updated source term analysis is detailed in Section 3.1. Gap release fractions for high-burnup rods (greater than 54 GWD/MTU) with an increased allowable LHGR have been calculated and are presented in Section 3.2. The updated HBRSEP2 dose consequence analyses are described in Section 3.3.

3.1 Source Term Analysis The updated isotopic inventory was determined in a manner consistent with Regulatory Guide 1.183, Position 3.1. This is consistent with what was performed to support the transition to using the Alternative Source Term, which was approved initially for selective implementation (Fuel Handling Accident), and later for full scope implementation and LOCA as described in References 10 through 12. Position 3.1 of Regulatory Guide 1.183 recommends using appropriate isotope generation and depletion computer codes. Position 3.1 of Regulatory Guide 1.183 calls for the use of "...an appropriate isotope generation and depletion computer code...",

and therefore, the source term analysis was performed using ORIGEN-S and ORIGEN-ARP (Reference 7).

A composite source term was created from both the 18-month fuel cycle source term and the 24-month fuel cycle source term. This was done by taking the maximum activity for either 18-month or 24-month cycles for all given isotopes. The purpose of this composite source term is to cover the transition from 18-month cycles to 24-month cycles and ensure a conservative source term input for use in the licensing basis accidents detailed in Section 3.3.

Table 2 provides the isotopes included in the HBRSEP2 source term, including the 60 isotopes (noble gas, halogen, and alkali metals) required by NUREG/CR-6604. The source term was

multiplied by 1.003 to reflect operation prior to shutdown at 100.3% of current rated power (100.3% of 2339 MWt, or 2346 MWt), which accounts for the uncertainty in measured core power (see Reference 13).

The assembly source term was derived from the core inventory based on the assumed fraction of fuel failures. A bounding radial peaking factor of 1.8 was then applied to simulate the effects of conservative application of local peaking factors, per the HBRSEP2 AST submittal, dated May 10, 2002.

The core inventory release fractions for the gap release and early in-vessel damage phases for the design basis LOCA utilized those release fractions provided in Regulatory Guide 1.183, Regulatory Position 3.2, Table 2 "PWR Core Inventory Fraction Released Into Containment." A request of this submittal (documented in Section 3.2) is to allow for 35 rods per assembly to exceed 6.3 kw/ft, provided the rods do not enter DNB; therefore this request only affects the Fuel Handling Accident. For non-LOCA events the fractions of the core inventory assumed to be in the gap are consistent with Regulatory Guide 1.183, Regulatory Position 3.2, Table 3, "Non-LOCA Fraction of Fission Product Inventory in Gap", except for Kr-85, Cs-134, and Cs-137, which are proposed to be tripled.

Table 2 shows the results of the maximization of the HBRSEP2 18-month and 24-month cycle source term analyses.

Table 2. HBRSEP2 Core Isotopic Inventory (Ci)

Isotope Activity (Ci) Isotope Activity (Ci) Isotope Activity (Ci)

Co58 5.990E+05 Ru103 1.019E+08 Cs136 3.712E+06 Co60 4.580E+05 Ru105 7.206E+07 Cs137 8.874E+06 Kr 85 8.246E+05 Ru106 3.835E+07 Ba139 1.148E+08 Kr 85m 1.653E+07 Rh105 6.632E+07 Ba140 1.112E+08 Kr 87 3.271E+07 Sb127 5.892E+06 La140 1.165E+08 Kr 88 4.377E+07 Sb129 1.824E+07 La141 1.040E+08 Rb 86 1.277E+05 Te127 5.800E+06 La142 1.004E+08 Sr 89 6.123E+07 Te127m 9.815E+05 Ce141 1.052E+08 Sr 90 6.475E+06 Te129 1.708E+07 Ce143 9.765E+07 Sr 91 7.690E+07 Te129m 3.276E+06 Ce144 8.487E+07 Sr 92 8.229E+07 Te131m 1.244E+07 Pr143 9.572E+07 Y 90 6.678E+06 Te132 9.045E+07 Nd147 4.176E+07 Y 91 8.006E+07 I131 6.339E+07 Np239 1.270E+09 Y 92 8.324E+07 I132 9.271E+07 Pu238 2.819E+05 Y 93 9.377E+07 I133 1.294E+08 Pu239 2.610E+04 Zr 95 1.075E+08 I134 1.453E+08 Pu240 3.681E+04 Zr 97 1.075E+08 I135 1.235E+08 Pu241 1.014E+07 Nb 95 1.084E+08 Xe133 1.295E+08 Am241 1.281E+04 Mo 99 1.176E+08 Xe135 4.161E+07 Cm242 3.383E+06 Tc 99m 1.042E+08 Cs134 1.294E+07 Cm244 4.427E+05

3.2 Gap Release Analysis The gap release analysis determines release fractions for a variety of volatile fission products in the gap between the pellet and cladding of a fuel rod. The computed release fractions correspond to a proposed increase in the Regulatory Guide 1.183 allowable fuel rod LHGR above 54 GWD/MTU burnup. The results of this analysis are used as isotopic inventory input to dose calculations for the Fuel Handling Accident.

HBRSEP2 has implemented the Alternative Source Term (AST) method in its current licensing basis, in accordance with Regulatory Guide 1.183 (References 10 through 12). Regulatory Guide 1.183 Table 3 provides gap release fractions for various volatile fission product isotopes and isotope groups, to be applied to non-LOCA accidents. This table limits the fuel rod LHGR to 6.3 kW/ft for rod burnups above 54 GWD/MTU, but a footnote to the table (Footnote 11) states that gap fractions calculated directly by the licensee may be considered on a case-by-case basis, if the calculations follow NRC-approved methodologies.

In recent years, experimental data have demonstrated that fuel pellets undergo significant thermal conductivity degradation (TCD) at high burnup, which increases interior fuel pellet temperatures. Reference 9 discusses this issue in more detail. Higher fuel temperatures will yield larger fission gas release fractions in the ANS 5.4 [1982] and [2011] models (References 3 and 4), particularly in the high-burnup range.

The ANS 5.4 [1982] standard has been revised, and the update (ANS 5.4 [2011]) acknowledges the conservatism of the previous version, based on additional experimental data after 1982.

The revised standard mandates the use of a NRC-approved fuel performance code that accounts for TCD, in determining temperature inputs for the gap fraction computations.

Because the ANS 5.4 [2011] standard is consistent with the basis for a proposed revision to Reg Guide 1.183 (see Reference 2), this gap release analysis employs the ANS 5.4 [2011]

method, using a fuel performance code applicable for use with HBRSEP2 fuel rods (COPERNIC). The gap release analysis accounts for TCD, and considers all pertinent long-lived and short-lived isotopes.

The method employed for this analysis is described in more detail in Section 3.2.1. Results from the specific gap fraction computations are documented in Section 3.2.3.

3.2.1 Method ANS 5.4 [2011] provides a method for determining the release fractions of short half-life isotopes, while deferring to specific NRC-approved fuel performance codes for the calculation of release fractions for long-lived isotopes. Additional details and background information related to this standard are provided in References 5 and 6.

The method in the ANS 5.4 [2011] standard is a Booth diffusion model of the fuel, which includes empirical fits to measurement data to yield release fractions as a function of fuel temperature and burnup.

3.2.1.1 Fuel Rod Type Considered The Areva HTP 15x15 fuel rod design is considered for the fission gas release calculations.

This is the design that is currently being irradiated in the Robinson reactor. As this fuel type is representative of a general 15x15 PWR design, the analysis of the fuel rod is judged to be applicable to other 15x15 designs that may be used in the Robinson reactor.

3.2.1.2 Rod Operational Power Histories The core design must maintain fuel rod power peaking below the peaking analyzed in the dose analyses. Table 3 shows the rod powers that are used in the gap release analysis for Robinson HTP fuel. These powers bound the current core design limits. The rod powers shown are binned into time step (burnup) increments less than or equal to 2 GWD/MTU, consistent with the restrictions of the ANS 5.4 [2011] method.

The nominal core average power (deposited within the fuel rod) is calculated below. The 0.974 value represents the fraction of total heat from fission that is deposited within the fuel rod.

Robinson:

2339000 x 1.003 x 0.974

= = 5.945 157 x 204 x 12 With this core average rod power, peaking factors can be determined from the rod powers in Table 3. Note that the rod LHGR beyond 54 GWD/MTU burnup is 7.0 kW/ft. See the computational results of the gap fraction calculations in Section 3.2.3.

Table 3. Projected Rod Powers in the Gap Release Analysis Robinson HTP fuel Rod Burnup Range Average Rod (GWD/MTU) Power (kW/ft) 0-2 9.804

>2 - 4 9.690

>4 - 6 9.643

>6 - 8 9.621

>8 - 10 9.747

>10 - 12 9.856

>12 - 14 9.797

>14 - 16 9.726

>16 - 18 9.811

>18 - 20 9.849

>20 - 22 9.838

>22 - 24 9.724

>24 - 26 9.654

>26 - 28 9.579

>28 - 30 9.427

>30 - 32 9.351

>32 - 34 9.279

>34 - 36 9.160

>36 - 38 7.950

>38 - 40 7.977

>40 - 42 7.960

>42 - 44 7.880

>44 - 46 7.797

>46 - 48 7.680

>48 - 50 7.602

>50 - 52 7.524

>52 - 54 7.377

>54 - 56 7.000

>56 - 58 7.000

>58 - 60 7.000

>60 - 62 7.000 3.2.1.3 Isotopes Considered for the Gap Release Calculations Of the radionuclide groups discussed in Regulatory Guide 1.183, the Noble Gases, Halogens, and Alkali Metals are pertinent for the Fuel Handling Accident. Table 4 shows the list of isotopes, along with their Regulatory Guide 1.183 isotope category, and the associated gap fraction valid for rod powers below 6.3 kW/ft when burnup exceeds 54 GWD/MTU.

Table 4. Isotopes Evaluated in the Gap Release Analysis Reg Guide Reg Guide 1.183 1.183,Table 3 Isotope Isotope Category Gap Fraction Kr-85 Kr-85 0.10 Long-lived (> 1-yr half-Cs-134 Alkali Metals 0.12 life) Isotopes Cs-137 Alkali Metals 0.12 I-130 Other Halogens 0.05 I-131 I-131 0.08 I-132 Other Halogens 0.05 I-133 Other Halogens 0.05 I-134 Other Halogens 0.05 I-135 Other Halogens 0.05 Br-83 Other Halogens 0.05 Br-85 Other Halogens 0.05 Br-87 Other Halogens 0.05 Kr-83m Other Noble Gases 0.05 Kr-85m Other Noble Gases 0.05 Short-lived (< 1-yr half- Kr-87 Other Noble Gases 0.05 life) Isotopes Kr-88 Other Noble Gases 0.05 Kr-89 Other Noble Gases 0.05 Xe-131m Other Noble Gases 0.05 Xe-133m Other Noble Gases 0.05 Xe-133 Other Noble Gases 0.05 Xe-135m Other Noble Gases 0.05 Xe-135 Other Noble Gases 0.05 Xe-137 Other Noble Gases 0.05 Xe-138 Other Noble Gases 0.05 Rb-86 Alkali Metals 0.12 Rb-88 Alkali Metals 0.12 Rb-89 Alkali Metals 0.12

Rb-90 Alkali Metals 0.12 Cs-136 Alkali Metals 0.12 Cs-138 Alkali Metals 0.12 Cs-139 Alkali Metals 0.12 3.2.1.4 Computation Process using ANS 5.4 [2011]

Gap fractions for each of the isotopes in Table 4 are determined, using either a direct result from the COPERNIC fuel performance code (for long-lived isotopes), or by computing gap releases for individual axial and radial fuel nodes (for short-lived isotopes). Subsections 3.2.1.4.1 through 3.2.1.4.3 discuss the specific procedures. Short-lived isotope calculations require input nodal fuel temperatures and burnups for each time step listed in Table 3. These nodal inputs are produced by COPERNIC.

3.2.1.4.1 Long-Lived Nuclides (T1/2 > 1 year)

The long-lived isotopes listed in Table 4 (Kr-85, Cs-134, and Cs-137) are treated as stable. The Kr-85 fission gas gap fraction is taken directly from the fuel performance code (COPERNIC),

calculated at a 95/95 bounding tolerance. The fuel performance code must account for TCD in its model.

Gap fractions for Cs-134 and Cs-137 are determined by multiplying the Kr-85 release fraction by 2 , in accordance with Section 5 of ANS 5.4 [2011].

3.2.1.4.2 Very Short-Lived Nuclides (T1/2 < 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months )

The fission gas gap fraction (called the release-to-birth [R/B] ratio in this standard) for fuel radial node i in axial node m, during an irradiation period at constant temperature and power, is calculated as:

= (1) where:

= 120 cm if , (2)

= 650 cm if , > (3)

, is the fuel temperature for radial node i in axial node m (K) is the temperature at which bubbles become interlinked on grain boundaries, per the burnup-dependent equations below:

9800

= + 273 if 18.2 GWD/MTU (4) 176 x

= 1434 12.85 x + 273 if > 18.2 GWD/MTU (5)

is the accumulated pellet average burnup (GWD/MTU) of axial node m is the precursor effect with values for pertinent isotope n in Table 5 is the decay constant for the isotope n of interest (sec-1)

S/V is the surface area to volume ratio

/ , . / , (6)

, = 7.6 x 10 + 1.41 x 10 + 2 x 10

= 4 x 10 (7) is the outer diameter of the fuel pellet (cm) is the inner diameter of the fuel pellet (cm) [non-zero for annular pellets]

is the local linear heat generation rate at axial node m (W/cm)

For equation (1), values for the precursor variable are provided for specific isotopes in ANS 5.4 [2011]. Pertinent precursor coefficients are shown in Table 5. The standard also notes that if a value is not listed, the precursor effect is small enough that can be assumed to be unity.

The above equations yield a gap fraction for an individual radial and axial fuel node. The overall gap fraction for the entire fuel rod is determined by weighting the nodal gap releases by the power levels of the individual nodes, along with nodal volumetric weighting if necessary. Any burnup dependence on short half-life isotopic inventories is ignored, as noted in item 4 of Section 3.2.2. This assumption is evaluated further in Section 3.2.3.

Table 5. Pertinent Values of from ANS 5.4 [2011]

Precursor Isotope coefficient I-132 137 I-133 1.21 I-134 4.4 Kr-85m 1.31 Kr-87 1.25 Kr-88 1.03 Kr-89 1.21 Xe-133 1.25 Xe-135m 23.5 Xe-135 1.85 Xe-137 1.07

3.2.1.4.3 Remaining Short-Lived Nuclides (T1/2 > 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months )

The fission gas gap fraction (release-to-birth [R/B] ratio) for fuel radial node i in axial node m, is calculated as:

= (8) where:

= (9)

In the above equation, is the fractal scaling factor used for these longer-lived radioactive nuclides. Fractal scaling factors for isotopes with half-lives under 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months are less than ~ 1.0, with the exception of I-132. Reference 5 recommends that equation (8) be used with I-132, even though its half-life is less than 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , to account for the large pre-cursor effect of Te-132, which has a much longer half-life (3.2 days).

The diffusion coefficient in equation (8) is multiplied by a factor of 2 for any cesiums.

3.2.1.5 Computer Codes The following computer programs were used for the calculations discussed in Section 3.2.3.

Each of these codes has been internally validated.

COPERNIC - this is a NRC-approved fuel performance code (Reference 8).

gapfrac - this is a Visual Basic for Applications (VBA) program that computes fission gas gap fractions in accordance with ANS 5.4 [2011], using the method described in Section 3.2.1.4.

3.2.2 Assumptions / Calculation Bases The following assumptions and bases are employed for the gap release analysis:

1) Nominal (best-estimate) fuel rod design/operational input was used for the COPERNIC model.

2) The rod power history selected for this analysis (see Table 3) bounds the limiting HBRSEP2 power history, in accordance with Footnote 11 to Table 3 of Regulatory Guide 1.183.

3) The Regulatory Guide 1.183 Fuel Rod LHGR limit above 54 GWD/MTU burnup (6.3 kW/ft) is associated with the heat produced in the fuel (~ 0.974 fraction of total power produced), and does not include energy deposited directly to the coolant.

4) It is sufficient to characterize the inventories of short half-life isotopes (e.g. I-131) as dependent only on instantaneous power level. Any burnup-dependent effects are judged to have a negligible effect on calculated release fractions.

5) For the HBRSEP2 reactor, 100.3% of nominal rated reactor power was used as the baseline operating power in COPERNIC. This accounts for the uncertainty in measured core power (Reference 13). The 100.3% power corresponds to 2346 MWt.

6) For the gap fraction calculations, all fuel rod evaluations were performed using a sufficient number of equally-spaced axial fuel segments and equal-volume radial rings in the fuel pellet. The ANS 5.4 [2011] standard requires at least 7 equal-volume radial nodes, and 10 or more axial nodes for the gap fraction computations.

7) Fuel assembly axial power data from a recent core design were used to determine appropriate axial power shapes for the COPERNIC fuel performance code.

8) Steady state reactor power operation was assumed for applicability to fuel handling accidents. No major transients are considered that could release significant quantities of volatile fission products to the fuel rod gap.

9) The gap fraction evaluation was performed only for HBRSEP2 UO2 fuel with no integral gadolinia poisons. Previous analyses (Reference 14) have shown that fuel rods with gadolinia poison yield lower fission gas release than rods of the same U-235 enrichment that do not contain gadolinia.

10) In accordance with ANS 5.4 [2011], gap fractions calculated using equations (1) and (8) in Section 3.2.1.4 are multiplied by a factor of 5, to account for uncertainties in release predictions.

3.2.3 Fission Gas Release Analysis -- Calculations / Results Section 3.2.1 described the method that is used to compute gap fractions for a Robinson HTP fuel rod with the power history profile shown in Table 3. The computer programs used for the calculations are discussed in Section 3.2.1.5.

The first step in the analysis was to build input decks for the COPERNIC code, so that appropriate nodal fuel temperatures could be obtained for input to the ANS 5.4 [2011] gap release equations.

Input information for COPERNIC includes:

Fuel rod dimensions and mechanical design data

Fuel rod backfill pressures

Number of axial nodes modeled

Axial power shape information

Number of burnup time steps

Rod power history

Enrichment and axial blanket details

Reactor core operational data

Axial power shapes were obtained from a recent Robinson core design, as a function of rod burnup. Using these shapes and the other input information detailed above, a COPERNIC case was executed for a Robinson HTP fuel rod with a 5.00 wt % U-235 central fuel enrichment and natural uranium axial blankets. This axial enrichment zoning generally yields the maximum fission gas release results.

Next, the gap fractions for the short-lived Table 4 isotopes were calculated, using COPERNIC-computed fuel temperatures and the Table 3 power history. To perform the ANS 5.4 [2011] gap fraction calculations, the gapfrac Visual Basic for Applications (VBA) program was written. This program applies the methods outlined in Section 3.2.1 to determine isotope gap release fractions for the entire fuel rod irradiation history. The gapfrac code was also used for the supporting calculations in the Reference 14 submittal.

The ANS 5.4 [2011] fission gas release results from the gapfrac computations are shown in Figure 1, for selected short-lived isotopes. Note the dips in the gap fractions for these short-lived isotopes as power levels are reduced, and the isotopes quickly establish lower equilibrium concentrations in the gap.

Table 6 lists the gap fractions calculated for each isotope from Table 4. The directly-computed maximum COPERNIC fission gas release yields the gap release value for Kr-85. As noted in subsection 3.2.1.4.1, in accordance with ANS 5.4 [2011], gap fractions for Cs-134 and Cs-137 are determined by multiplying the Kr-85 release fraction by 2 .

From the above discussion, as well as the Table 6 results, it is clear that, if it is desired to exceed a 6.3 kW/ft LHGR above 54 GWD/MTU, increased gap fractions for the long-lived isotopes must be accounted for in dose analyses. The results of this analysis show that with the chosen power history in Table 3, calculated gap fractions remain below 2 times the Regulatory Guide 1.183 Table 3 values for Kr-85, Cs-134, and Cs-137. For additional conservatism, a bounding multiplier of 3 will be applied to the Regulatory Guide 1.183 Table 3 values for these isotopes.

For the short-lived isotopes, Table 6 and Figure 1 show that the maximum computed gap fractions remain well under the existing Regulatory Guide 1.183 Table 3 values.

0.014 0.012 0.010 Calculated Gap Fraction 0.008 I-131 I-132 0.006 Xe-133 Cs-136 0.004 0.002 0.000 0 10000 20000 30000 40000 50000 60000 70000 Rod Burnup (MWD/MTU)

Figure 1. Calculated Gap Fractions for Selected Short-Lived Isotopes - 5.00 wt % U-235 Robinson HTP Fuel

Table 6. Results from Gap Release Calculations Robinson HTP Reg Guide fuel calculated Isotope 1.183 Table 3 maximum gap Isotope Category Value fraction Ratio Long-lived (> 1-yr half-life) Isotopes (from COPERNIC results)

Kr-85 Kr-85 0.10 0.168 1.68 Cs-134 Alkali Metals 0.12 0.238 1.98 Cs-137 Alkali Metals 0.12 0.238 1.98 Short-lived (< 1-yr half-life) Isotopes (from gapfrac results)

I-130 Other Halogens 0.05 0.0043 0.09 I-131 I-131 0.08 0.0086 0.11 I-132 Other Halogens 0.05 0.0096 0.19 I-133 Other Halogens 0.05 0.0051 0.10 I-134 Other Halogens 0.05 0.0029 0.06 I-135 Other Halogens 0.05 0.0037 0.07 Br-83 Other Halogens 0.05 0.0023 0.05 Br-85 Other Halogens 0.05 0.0003 0.01 Br-87 Other Halogens 0.05 0.0002 0.00 Kr-83m Other Nobles 0.05 0.0020 0.04 Kr-85m Other Nobles 0.05 0.0036 0.07 Kr-87 Other Nobles 0.05 0.0019 0.04 Kr-88 Other Nobles 0.05 0.0025 0.05 Kr-89 Other Nobles 0.05 0.0004 0.01 Xe-131m Other Nobles 0.05 0.0094 0.19 Xe-133m Other Nobles 0.05 0.0062 0.12 Xe-133 Other Nobles 0.05 0.0081 0.16 Xe-135m Other Nobles 0.05 0.0036 0.07 Xe-135 Other Nobles 0.05 0.0046 0.09 Xe-137 Other Nobles 0.05 0.0004 0.01 Xe-138 Other Nobles 0.05 0.0007 0.01 Rb-86 Alkali Metals 0.12 0.0105 0.09 Rb-88 Alkali Metals 0.12 0.0008 0.01 Rb-89 Alkali Metals 0.12 0.0007 0.01 Rb-90 Alkali Metals 0.12 0.0003 0.00 Cs-136 Alkali Metals 0.12 0.0137 0.11 Cs-138 Alkali Metals 0.12 0.0011 0.01 Cs-139 Alkali Metals 0.12 0.0006 0.00

3.2.4 Conclusions With a bounding operational power history applied to the HBRSEP2 fuel rods for burnups up to 54.0 GWD/MTU, a maximum linear heat generation rate of 7.0 kW/ft has been evaluated for the remainder of the allowable rod burnup (54.0 to 62.0 GWD/MTU).

The conservative rod power history in Table 3 has been analyzed using a NRC-approved fuel performance code (COPERNIC), to obtain a fine mesh of fuel temperatures that include the effects of thermal conductivity degradation at high burnup. With these fuel temperatures, fission gas release calculations have been performed in conformance with the method described in the ANS 5.4 [2011] standard.

The calculations discussed in Section 3.2.3 show that, for the isotopes considered in fuel handling accident dose analyses, the Regulatory Guide 1.183 Table 3 gap fractions must be increased for the long-lived Kr-85, Cs-134, and Cs-137 isotopes, as shown in Table 7.

Computed fission gas release gap fractions for all other isotope groups remain well below the values from Table 3 of Regulatory Guide 1.183.

Table 7. Bounding Gap Fractions for Application to HBRSEP2 Fuel Handling Accidents Gap Fraction from Isotope or Isotope Table 3 of Reg Guide Bounding Gap Ratio Group 1.183 (Rev. 0) Fraction I-131 0.08 0.08 1 Kr-85 0.10 0.30 3 Other Noble Gases 0.05 0.05 1 Other Halogens 0.05 0.05 1 Cs-134 (Alkali Metal) 0.12 0.36 3 Cs-137 (Alkali Metal) 0.12 0.36 3 Other Alkali Metals 0.12 0.12 1 3.3 Design Basis Accident Dose Consequences With the source term reanalyzed and the gap release analysis complete, the dose consequences for each accident listed in Table 1 were calculated. Fuel failure assumptions will be verified on a cycle specific basis via the reload process.

This section outlines the changes to the licensing basis accident dose consequences. Each section discusses these changes and reports updated dose consequences. Enclosure 2 contains tables with the inputs to the dose consequences analysis for each accident.

3.3.1 Loss of Coolant Accident There are two changes to the current licensing basis. First, the LOCA analysis uses the updated source term detailed in Table 2. Second, the current licensing basis analysis takes credit for activity removal via diffusiophoresis. However, with this submittal, credit for diffusiophoresis is no longer assumed.

As a result of these changes, the updated doses for HBRSEP2 are 22.5 Rem Total Effective Dose Equilvalent (TEDE) for the Exclusion Area Boundary (EAB), 1.53 Rem TEDE for the Low Population Zone (LPZ), and 4.40 Rem TEDE for the control room.

3.3.2 Fuel Handling Accident (FHA)

There are three changes to the current licensing basis. The first change is that the FHA source term was recalculated based on the new isotopic activities shown in Table

2. Second, a request is made to allow up to 35 rods per assembly to exceed the maximum linear heat generation rate of 6.3 kw/ft, detailed in Table 3 of Regulatory Guide 1.183, but remain below a new maximum LHGR of 7.0 kW/ft. This change yields increased release fractions for Kr-85, Cs-134, and Cs-137, which are tripled from Regulatory Guide 1.183. All other isotopes maintain the release fractions outlined in Table 3 of Regulatory Guide 1.183.

Third, the accident timing is being updated as well. It is assumed that the accident occurs 116 hours0.00134 days 0.0322 hours 1.917989e-4 weeks 4.4138e-5 months after shutdown of the plant for a refueling outage.

As a result of these changes, the updated doses for HBRSEP2 are 4.34 Rem TEDE for the EAB, 0.22 Rem TEDE for the LPZ, and 3.56 Rem TEDE for the control room.

3.3.3 Main Steam Line Break (MSLB)

There are two changes to the current licensing basis. The first change is that the MSLB source term was recalculated based on the new isotopic activities shown in Table 2. Note that these pins are not permitted to exceed the 6.3 kw/ft limit detailed in Regulatory Guide 1.183. Further, the transition from 18-month cycles to 24-month cycles will not impact the accident scenarios without fuel failure, namely the pre-accident and accident induced iodine spiking cases. Second, the Primary-to-Secondary Leakage Rate (PSLR) is assumed to be 450 gpd, consistent with the TS value of 150 gpd per steam generator.

As a result of these changes, the updated doses for HBRSEP2 are 1.72 Rem TEDE for the EAB, 0.38 Rem TEDE for the LPZ, and 1.53 Rem TEDE for the control room.

3.3.4 Locked Rotor Accident (LRA)

There are three changes to the current licensing basis. One change pertains to the updated source term detailed in Table 2. Second, it is now assumed that 12 assemblies experience DNB. Note that these pins are not permitted to exceed the 6.3 kw/ft limit detailed in Regulatory Guide 1.183. Third, the PSLR is assumed to be 450 gpd, consistent with the TS value of 150 gpd per steam generator.

As a result of these changes, the updated doses for HBRSEP2 are 2.09 Rem TEDE for the EAB, 0.16 Rem TEDE for the LPZ, and 0.68 Rem TEDE for the control room.

3.3.5 Single Rod Withdrawal Accident (SRWA)

There are four changes to the current licensing basis. One change pertains to the updated source term detailed in Table 2. Second, it is now assumed that 4 assemblies in the core experience DNB and no assembly experiences fuel melt. Note that these pins are not permitted to exceed the 6.3 kw/ft limit detailed in Regulatory Guide 1.183. Third, the PSLR is assumed to be 450 gpd, consistent with the TS value of 150 gpd per steam generator. Fourth, no credit is taken for control room ventilation system switchover to emergency mode.

As a result of these changes, the updated doses for HBRSEP2 are 0.69 Rem TEDE for the EAB, 0.05 Rem TEDE for the LPZ, and 2.57 Rem TEDE for the control room.

3.3.6 Rod Ejection Accident (REA)

The rod ejection accident is currently not part of the suite of accidents with a dose consequence analysis in place in the HBRSEP2 licensing basis. The change from the current licensing basis is that the rod ejection dose consequences analysis is now analyzed assuming fuel failure.

The source term assumed for the rod ejection accident is based on the isotopic radioactivity levels in Table 2. It is assumed that 10% of rods in the core experience DNB.

A radial peaking factor of 1.8 is assumed. Note that these pins are not permitted to exceed the 6.3 kw/ft limit detailed in Regulatory Guide 1.183 Footnote 11. The activity is assumed to be released instantly and homogenously into the primary coolant with release fractions consistent with Appendix H of Regulatory Guide 1.183. Note that Footnote 11 sets gap fractions for rod ejection only for noble gases and iodines; additional gap fractions were set to 10% for bromine and 12% for alkali metals. Activity is released to the atmosphere via steaming from the steam generator pressure operated relief valves (PORVs) until the decay heat generated in the reactor core can be removed by the Residual Heat Removal (RHR) system at 53.2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months into the event.

The containment release model does not credit removal of fission products via containment spray. Iodine composition fractions are set to 97% elemental and 3%

organic. Containment leakage rate assumptions are consistent with the Technical Specifications of HBRSEP2, which is 0.1% by weight of the containment air for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and then 0.05% by weight of the containment air thereafter.

The calculation of radiation doses for post rod ejection steam generator releases takes the same profile for primary to secondary transfer as the corresponding calculations for the locked rotor and single rod withdrawal accidents. The PSLR is held constant to 150 gpd for each steam generator. It is also assumed to have the same cooldown profile, as the AST analysis of the single rod withdrawal accident.

The control room ventilation assumptions for the containment release scenario follows the following sequence. Prior to switchover to the emergency pressurization mode, the activity is assumed to enter the control room based on the normal ventilation intake rate of 400 cfm and the control room inleakage rate of 300 cfm. The control room is placed in the emergency pressurization ventilation mode upon an SI signal. For analysis purposes, the switchover to the emergency pressurization mode is assumed to occur at 5 minutes 35 seconds for the containment release scenario. Hagan Room inleakage into the control room is reduced as a result of switchover to the emergency pressurization mode.

Therefore, after pressurization, the assumed control room unfiltered inleakage is reduced to 230 cfm.

The control room ventilation assumptions for the secondary side release scenario remain the same as the other licensing basis accidents. Prior to switchover to the emergency pressurization mode, the activity is assumed to enter the control room based on the normal ventilation intake rate of 400 cfm and the control room inleakage rate of 300 cfm.

The control room is manually placed in the emergency pressurization ventilation mode upon a control room area radiation monitor isolation signal. Switchover is assumed to occur at one hour. Hagan Room inleakage into the control room is reduced as a result of switchover to the emergency pressurization mode. Therefore, after one hour the assumed control room unfiltered inleakage is reduced to 230 cfm.

Following the switchover to the emergency pressurization mode, the control room air is recirculated through filters at a rate of 2600 cfm, and the intake flow is also directed through the filters. Filter removal in the emergency pressurization mode of the control room ventilation system is modeled in the same manner as the current licensing basis.

The limiting doses for the rod ejection accident at HBRSEP2 are from the containment release and are 4.07 Rem TEDE for the EAB, 0.69 Rem TEDE for the LPZ, and 4.83 Rem TEDE for the control room.

Table 8. Baseline Accident Dose Consequences Baseline Doses (Rem TEDE)

Accident EAB LPZ Control Room Loss of Coolant Accident 24.1 / 25 1.58 / 25 4.46 / 5 Fuel Handling Accident 5.96 / 6.3 0.30 / 6.3 4.46 / 5 Main Steam Line Break - Fuel Failure 2.93 / 25 0.42 / 25 1.61 / 5 Locked Rotor 2.24 / 2.5 0.21 / 2.5 0.86 / 5 Single Rod Withdrawal 1.76 / 2.5 0.24 / 2.5 0.75 / 5 Rod Ejection NA NA NA Table 9. Updated Accident Dose Consequences Updated Doses (Rem TEDE)

Accident EAB LPZ Control Room Loss of Coolant Accident 22.5 / 25 1.53 / 25 4.40 / 5 Fuel Handling Accident 4.34 / 6.3 0.22 / 6.3 3.56 / 5 Main Steam Line Break - Fuel Failure 1.72 / 25 0.38 / 25 1.53 / 5 Locked Rotor 2.09 / 2.5 0.16 / 2.5 0.68 / 5 Single Rod Withdrawal 0.69 / 2.5 0.05 / 2.5 2.57 / 5 Rod Ejection 4.07 / 6.3 0.69 / 6.3 4.83 / 5

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria The proposed changes have been evaluated to determine whether applicable regulations and requirements continue to be met.

The requirement of 10 CFR 50.67(b)(1), specifically: A licensee who seeks to revise its current accident source term in design basis radiological consequence analyses shall apply for a license amendment under § 50.90. The application shall contain an evaluation of the consequences of applicable design basis accidents previously analyzed in the safety analysis report, are met by the information contained in this license amendment request.

Duke Energy Progress, Inc. has determined that the proposed changes do not require any exemptions or relief from regulatory requirements, other than the changes to the current licensing basis discussed in this submittal.

The analysis described in this submittal performed in accordance with Regulatory Guide 1.183, demonstrates that HBRSEP, Unit No. 2 complies with the Total Effective Dose Equivalent (TEDE) dose limits delineated in 10 CFR 50.67(b)(2) at the exclusion area and low population zone boundaries, as well as the control room; and, therefore, requests an operating license amendment to implement the new AST licensing basis.

Per HBRSEP Unit No. 2 UFSAR Sections 3.1.1.1, the General Design Criteria (GDC) in existence at the time HBRSEP, Unit No. 2 was licensed (July 1970) for operation were contained in Proposed Appendix A to 10 CFR 50, General Design Criteria for Nuclear Power Plants, published in the Federal Register on July 11, 1967. (Appendix A to 10 CFR 50, effective in 1971 and subsequently amended, is somewhat different from the proposed 1967 criteria.)

HBRSEP, Unit No. 2 was evaluated with respect to the proposed 1967 GDC. The original FSAR contained a discussion of the criteria as well as a summary of the criteria by groups. UFSAR Section 3.1.2.11 identifies the following criterion for GDC-11, Control Room The facility shall be provided with a Control Room from which actions to maintain safe operational status of the plant can be controlled. Adequate radiation protection shall be provided to permit continuous occupancy of the Control Room under any credible post-accident condition or as an alternative, access to other areas of the facility as necessary to shut down and maintain safe control of the facility without excessive radiation exposures of personnel. (GDC 11)

The dose limits of the proposed revised AST continues to meet this criterion.

4.2 No Significant Hazards Consideration Determination Duke Energy Progress, Inc. is proposing changes to the licensing basis of Renewed Facility Operating License No. DPR-23 for H. B. Robinson Steam Electric Plant (HBRSEP), Unit No. 2. These changes update the licensing basis Alternate Source Term (AST). There are no Operating License or Technical Specifications changes required or proposed to implement this licensing basis change.

An evaluation of the proposed change has been performed in accordance with 10 CFR 50.91(a)(1) regarding no significant hazards considerations using the standards in 10 CFR 50.92(c). A discussion of these standards as they relate to this amendment request follows:

1. The Proposed Change Does Not Involve a Significant Increase in the Probability or Consequences of an Accident Previously Evaluated.

Revision of the AST does not affect the design or operation of HBRSEP, Unit No. 2. Rather, once the occurrence of an accident has been postulated, the new source term is an input to evaluate the consequences of the postulated accident. The revision of the AST has been evaluated. Based on the results of this analysis, it has been demonstrated that the dose consequences are within the regulatory guidance provided by the NRC. This guidance is presented in 10 CFR 50.67 and Regulatory Guide 1.183.

Therefore, this change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. The Proposed Change Does Not Create the Possibility of a New or Different Kind of Accident From Any Previously Evaluated.

The proposed change does not affect plant structures, systems, or components.

The proposed change is a revision evaluation and does not initiate design basis accidents.

Thus, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. The Proposed Change Does Not Involve a Significant Reduction in the Margin of Safety.

The proposed change is associated with a revision to the licensing basis for HBRSEP, Unit No. 2. The revised AST is in accordance with 10 CFR 50.67 and the associated Regulatory Guide 1.183. The analysis has been performed using conservative methodologies in accordance with regulatory guidance. The dose consequences are within the acceptance criteria found in the regulatory guidance associated with Alternative Source Terms.

The proposed change continues to ensure that doses at the exclusion area and low population zone boundaries, as well as the control room, are within the corresponding regulatory limits. Specifically, the margin of safety for the radiological consequences of these accidents is considered to be that provided by meeting the applicable regulatory limits.

Therefore, this change does not involve a significant reduction in a margin of safety.

Based on the above discussion, Duke Energy Progress, Inc. has determined that the requested change does not involve a significant hazards consideration.

5.0 ENVIRONMENTAL CONSIDERATION

10 CFR 51.22(c)(9) provides criteria for identification of licensing and regulatory actions for categorical exclusion for performing an environmental assessment. A proposed change for an operating license for a facility does not require an environmental assessment if operation of the facility in accordance with the proposed change would not (1) involve a significant hazards consideration; (2) result in a significant change in the types or significant increases in the amounts of any effluents that may be released offsite; (3) result in an increase in individual or cumulative occupational radiation exposure. Duke Energy Progress Inc., has reviewed this request and determined that the proposed change meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment is required in connection with the issuance of the amendment. The basis for this determination follows:

Duke Energy Progress Inc., is proposing a change to the licensing basis of Renewed Facility Operating License No. DPR-23 for H. B. Robinson Steam Electric Plant (HBRSEP), Unit No.

2. This change revises the licensing basis for HBRSEP, Unit No. 2, to implement the Alternative Source Term (AST) described in Regulatory Guide 1.183, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, for the evaluation of specific design basis accidents. There are no Operating License or Technical Specifications changes required or proposed to implement this licensing basis change.

Basis The proposed change meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9) for the following reasons:

1. As demonstrated in the No Significant Hazards Consideration Determination, the proposed change does not involve a significant hazards consideration.

2. As demonstrated in the No Significant Hazards Consideration Determination, the proposed change does not result in a significant increase in the consequences of an accident previously evaluated and does not result in the possibility of a new or different kind of accident. Therefore, the proposed change does not result in a significant change in the types or significant increases in the amounts of any effluents that may be released offsite.

3. The Alternative Source Term does not affect the design or operation of the facility.

Rather, once the occurrence of an accident has been postulated, the Alternative Source Term is an input to evaluate the consequences. Based on the results of this analysis, it has been demonstrated that the dose consequences are within the regulatory guidance provided by the NRC for use with the Alternative Source Term.

Therefore, the proposed change does not result in a significant increase in either individual or cumulative occupational radiation exposures.

6.0 REFERENCES

1. Regulatory Guide 1.183, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, Revision 0, U.S. Nuclear Regulatory Commission, July 2000.

2. Draft Regulatory Guide DG-1199, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors (Proposed Revision 1 of Regulatory Guide 1.183), U.S. Nuclear Regulatory Commission, October 2009.

3. ANSI/ANS-5.4-1982, Method for Calculating the Fractional Release of Volatile Fission Products from Oxide Fuel, American National Standard published by the American Nuclear Society, November 1982.

4. ANSI/ANS-5.4-2011, Method for Calculating the Fractional Release of Volatile Fission Products from Oxide Fuel, American National Standard published by the American Nuclear Society, May 2011.

5. PNNL-18212, Update of Gap Release Fractions for non-LOCA Events Utilizing the Revised ANS 5.4 Standard, Revision 1, Pacific Northwest National Laboratory (C. Beyer and P. Clifford), June 2011.

6. NUREG/CR-7003, Background and Derivation of ANS 5.4 [2011] Standard Fission Product Release Model, J. Turnbull and C. Beyer, prepared for the U.S.

Nuclear Regulatory Commission, January 2010.

7. ORNL/TM-2005/39 (Version 5.1) SCALE: A Modular Code System for Performing Standardized Computer Analyses for Licensing Evaluation, Oak Ridge National Laboratory, November 2006.

8. BAW-10231P-A, Revision 1, Copernic Fuel Rod Design Computer Code, Framatome ANP (now Areva), January 2004.

9. NRC Information Notice IN 2009-23, Nuclear Fuel Thermal Conductivity Degradation, October 2009.

10. H. B. Robinson Steam Electric Plant Unit 2 - Issuance of Amendment -

Technical Specification Change Regarding Selective Implementation of Alternative Radiological Source Term (TAC No. MB4632) - letter from R.

Subbaratnam (U.S. NRC) to J. Moyer (Carolina Power & Light), October 4, 2002.

11. H. B. Robinson Steam Electric Plant, Unit No 2 - Issuance of an Amendment on Full Implementation of the Alternative Source Term (TAC No. MB5105) -

letter from C. P. Patel (U.S. NRC) to J. Moyer (Carolina Power & Light),September 24, 2004.

12. H. B. Robinson Steam Electric Plant, Unit No 2 - Issuance of an Amendment on Full Implementation of the Alternate Source Term for the Loss-of-Coolant Accident (TAC No. MB5709) - letter from C. P. Patel (U.S. NRC) to T. D. Walt (Carolina Power & Light), July 11, 2006.

13. H. B. Robinson Steam Electric Plant, Unit No 2 - Issuance of Amendment Regarding a 1.7 Percent Power Uprate (TAC No. MB5106) - letter from R.

Subbaratnam (U.S. NRC) to J. Moyer (Carolina Power & Light), November 5, 2002.

14. Catawba Nuclear Station, Units 1 and 2; McGuire Nuclear Station, Units 1 and 2; Oconee Nuclear Station, Units 1, 2, and 3 - Issuance of Amendments Regarding Request to Use an Alternate Fission Gas Gap Release Fraction (CAC NOS. MF6480, MF6481, MF6482, MF6483, MF6484, MF6485, and MF6486)"

- letter from J. Hall (U.S. NRC) to R. Repko (Duke Energy), July 19, 2016.

United States Nuclear Regulatory Commission Enclosure II to Serial: RNP-RA/16-0057 8 Pages (Including this cover sheet)

This enclosure provides tables for each accident detailing the current license inputs and assumptions. It also contains the values used in the analyses that support this license amendment request.

Table 1. Loss of Coolant Accident - Dose Consequence Analysis Inputs and Assumptions

- - Note that all assumptions from Regulatory Guide 1.183 for the LOCA are applicable for this analysis, unless otherwise stated in this table.

Input or Assumption Description CLB Value Updated Value Reactor power with uncertainty (MW) 2346 No change UFSAR Updated to Table 2 Source Term (Ci)

Table 15.0.12-1 of Enclosure 1 Control room volume (ft3) 20,124 No change Control room outside air makeup (cfm) 400 No change Control room ventilation normal mode Iodine intake filter 0 No change efficiencies (%)

Control room ventilation emergency mode start time (sec) 35 No change Control room ventilation emergency mode recirculation (cfm) 2600 No change Control room ventilation emergency mode Iodine intake filter (95, 99, 95) No change efficiencies (elemental, particulate, organic %)

Control room ventilation normal mode unfiltered inleakage 170 No change (cfm)

Control room ventilation pressurized mode unfiltered inleakage 100 No change (cfm)

Containment volume (ft3) 1,958,526 No change Sprayed containment volume (ft3) 1,596,198.7 No change Unsprayed containment volume (ft3) 362,327.3 No change Containment cooling fans combined flow rate start time (sec) 76 No change Containment cooling fans combined flow rate (cfm) 65,000 No change Particulate spray removal coefficient (hr-1) 3.427 No change Elemental spray removal coefficient (hr-1) 20 No change Natural Deposition Removal Coefficient (hr-1) 0.1 No change Diffusiophoresis removal coefficient (hr-1) 0.2 Not credited Containment leak rate, first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months (percent weight / day) 0.1 No change Containment leak rate, after first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months (percent weight /

0.05 No change day)

All ESF leakage is assumed to be released from the Auxiliary Building and not the RWST. This is conservative as the NA No change Auxiliary Building has a higher X/Q Modeled ESF leakage as 2 times TRM limit (gph) 2 No change ESF leakage start time (minutes) 21 No change Flash fraction 0.1 No change Sump pH >7 No change Sump volume not relevant before ECCS leakage starts (0-21 NA No change minutes)

Sump volume from 21minutes to 40 minutes (ft3) 35,850 No change Sump volume from 40 minutes to 51.5 minutes (ft3) 40,889 No change Sump volume from 51.5 minutes to 30 days (ft3) 43,939 No change EAB dispersion factors (sec/m3) Enc. 2, Table 7 No change LPZ dispersion factors (sec/m3) Enc. 2, Table 7 No change CR dispersion factors, containment release, cont nearest point Enc. 2, Table 7 No change (sec/m3)

EAB dispersion factors, ESF release, RHR HX room (sec/m3) Enc. 2, Table 7 No change

Table 2. Fuel Handling Accident - Dose Consequence Analysis Inputs and Assumptions

- - Note that all assumptions from Regulatory Guide 1.183 for the FHA are applicable for this analysis, unless otherwise stated in this table.

Input or Assumption Description CLB Value Updated Value Reactor power with uncertainty (MW) 2346 No change UFSAR Updated to Table 2 Source Term (Ci)

Table 15.0.12-1 of Enclosure 1 Number of assemblies damaged 1 No change Peaking factor 1.8 No change Decay time (hr) 56 116 Fuel rod pressurization (psig) < 1200 No change Number of rods that can exceed 6.3 kW/ft 0 35 Kr-85 release fraction, if over 6.3 kW/ft Not permitted 0.30 Cs-134, Cs-137 release fractions if over 6.3 kW/ft Not permitted 0.36 Control room volume (ft3) 20,124 No change Control room outside air makeup (cfm) 400 No change Control room ventilation normal mode Iodine intake filter 0 No change efficiencies (%)

Control room ventilation emergency mode start time (hr) 1 No change Control room ventilation emergency mode recirculation (cfm) 2600 No change Control room ventilation emergency mode Iodine intake filter (95, 99, 95) No change efficiencies (elemental, particulate, organic %)

Control room ventilation normal mode unfiltered inleakage (cfm) 300 No change1 Control room ventilation pressurized mode unfiltered inleakage 230 No change1 (cfm)

Refueling cavity water iodine DF (FHA in containment) 200 No change Refueling cavity water noble gases DF (FHA in containment) 1 No change Spent fuel pool water iodine DF (FHA in fuel handling building) 138 No change Spent fuel pool water noble gases DF (FHA in fuel handling 1 No change building)

Depth of water above fuel inside containment (ft) 23 No change Depth of water above fuel inside fuel handling building (ft) 21 No change Release to the environment is over a 2-hour time period NA No change Fuel handling building filter efficiencies % (elemental, organic) (90, 70) No change EAB dispersion factors (sec/m3) Enc. 2, Table 7 No change LPZ dispersion factors (sec/m3) Enc. 2, Table 7 No change CR dispersion factors, FHA in containment, Enc. 2, Table 7 No change cont nearest point (sec/m3)

EAB dispersion factors, FHA in fuel handling building, Enc. 2, Table 7 No change plant stack (sec/m3) 1 The CLB analysis also included cases with 85 cfm and 500 cfm unfiltered inleakage, but only the 300 cfm case was retained for this submittal.

Table 3. Locked Rotor - Dose Consequence Analysis Inputs and Assumptions

- - Note that all assumptions from Regulatory Guide 1.183 for the LRA are applicable for this analysis, unless otherwise stated in this table.

Input or Assumption Description CLB Value Updated Value Reactor power with uncertainty (MW) 2346 No change UFSAR Updated to Table 2 Source Term (Ci)

Table 15.0.12-1 of Enclosure 1 Number of assemblies damaged (DNB) 17 12 Peaking factor 1.8 No change Number of rods that can exceed 6.3 kW/ft and experience DNB 0 No change Release duration (hr) 53.2 No change Reactor coolant specific activity (uCi/g) 0.5 No change Secondary side specific activity (uCi/g) 0.1 No change Control room volume (ft3) 20,124 No change Control room outside air makeup (cfm) 400 No change Control room ventilation normal mode Iodine intake filter 0 No change efficiencies (%)

Control room ventilation emergency mode start time (hr) 1 No change Control room ventilation emergency mode recirculation (cfm) 2600 No change Control room ventilation emergency mode Iodine intake filter (95, 99, 95) No change efficiencies (elemental, particulate, organic %)

Control room ventilation normal mode unfiltered inleakage (cfm) 300 No change1 Control room ventilation pressurized mode unfiltered inleakage 230 No change1 (cfm)

Reactor coolant volume (ft3) 8,254 No change Secondary side volume (ft3) 4,262 No change Primary to Secondary Leak Rate (gpd) 432 450 Steam release from steam generators, 0-2 hours (lbm) 301,967.3 No change Steam release from steam generators, 2-8 hours (lbm) 566,768.3 No change Steam release from steam generators, 8-24 hours (lbm) 1,124,995.8 No change Steam release from steam generators, 24-53.2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months (lbm) 1,637,910.1 No change EAB dispersion factors (sec/m3) Enc. 2, Table 7 No change LPZ dispersion factors (sec/m3) Enc. 2, Table 7 No change CR dispersion factors, MSSV/PORVs (sec/m3) Enc. 2, Table 7 No change 1

The CLB analysis also included cases with 85 cfm and 500 cfm unfiltered inleakage, but only the 300 cfm case was retained for this submittal.

Table 4. Single Rod Withdrawal - Dose Consequence Analysis Inputs and Assumptions

- - Note that all assumptions from Regulatory Guide 1.183 for the SRW are applicable for this analysis, unless otherwise stated in this table.

Input or Assumption Description CLB Value Updated Value Reactor power with uncertainty (MW) 2346 No change UFSAR Updated to Table 2 Source Term (Ci)

Table 15.0.12-1 of Enclosure 1 Number of assemblies damaged (DNB) 1 4 Number of assemblies with fuel melt 3 0 Peaking factor 1.8 No change Number of rods that can exceed 6.3 kW/ft and experience DNB 0 No change Release duration (hr) 53.2 No change Reactor coolant specific activity (uCi/g) 0.5 No change Secondary side specific activity (uCi/g) 0.1 No change Control room volume (ft3) 20,124 No change Control room outside air makeup (cfm) 400 No change Control room ventilation normal mode Iodine intake filter 0 No change efficiencies (%)

Control room ventilation emergency mode start time (hr) 1 No credit Control room ventilation emergency mode recirculation 2600 No credit (cfm)

Control room ventilation emergency mode Iodine intake (95, 99, 95) No credit filter efficiencies (elemental, particulate, organic %)

Control room ventilation normal mode unfiltered inleakage (cfm) 300 No change1 Control room ventilation pressurized unfiltered inleakage 230 No credit (cfm)

Reactor coolant volume (ft3) 8,254 No change Secondary side volume (ft3) 4,262 No change Primary to Secondary Leak Rate (gpd) 432 450 Steam release from steam generators, 0-2 hours (lbm) 301,967.3 No change Steam release from steam generators, 2-8 hours (lbm) 566,768.3 No change Steam release from steam generators, 8-24 hours (lbm) 1,124,995.8 No change Steam release from steam generators, 24-53.2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months (lbm) 1,637,910.1 No change EAB dispersion factors (sec/m3) Enc. 2, Table 7 No change LPZ dispersion factors (sec/m3) Enc. 2, Table 7 No change CR dispersion factors, MSSV/PORVs (sec/m3) Enc. 2, Table 7 No change 1

The CLB analysis also included cases with 85 cfm and 500 cfm unfiltered inleakage, but only the 300 cfm case was retained for this submittal.

Table 5. MSLB (with Fuel Failure) - Dose Consequence Analysis Inputs and Assumptions

- - Note that all assumptions from Regulatory Guide 1.183 for the MSLB are applicable for this analysis, unless otherwise stated in this table.

Updated Input or Assumption Description CLB Value Value Reactor power with uncertainty (MW) 2346 No change UFSAR Updated to Source Term (Ci) Table Table 2 of 15.0.12-1 Enclosure 1 Number of damaged assemblies 2 No change Peaking factor 1.8 No change Number of rods that can exceed 6.3 kW/ft and experience DNB 0 No change Steaming release duration (hr) 53.2 No change Time when primary system temperature is < 212 F (hr) 98.8 No change Reactor coolant specific activity (uCi/g) 0.5 No change Secondary side specific activity (uCi/g) 0.1 No change Control room volume (ft3) 20,124 No change Control room outside air makeup (cfm) 400 No change Control room ventilation normal mode Iodine intake filter efficiencies 0 No change

(%)

Control room ventilation emergency mode start time (sec) 50 No change Control room ventilation emergency mode recirculation (cfm) 2600 No change Control room ventilation emergency mode Iodine intake filter (95, 99, 95) No change efficiencies (elemental, particulate, organic %)

CR vent. pressurized mode unfiltered inleakage (50sec to 1hr) (cfm) 300 No change1 CR vent. pressurized mode unfiltered inleakage (1hr to duration) 230 No change1 (cfm)

Reactor coolant volume (ft3) 8,254 No change Secondary side volume (ft3) 4,262 No change Primary to Secondary Leak Rate (gpd) 432 450 Integrated Steam Release from Ruptured and Unaffected Steam Generators Steam release from ruptured steam generator, 0 hours0 days 0 hours 0 weeks 0 months (lbm) 161,194 No change Steam release from ruptured steam generator, 0-2 hours (lbm) 161,304.2 No change Steam release from ruptured steam generator, 0-8 hours (lbm) 161,634.7 No change Steam release from ruptured steam generator, 0-24 hours (lbm) 162,516.0 No change Steam release from ruptured steam generator, 0-53.2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months (lbm) 164,124.3 No change Steam release from ruptured steam generator, 0-98.8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months (lbm) 166,636.1 No change Steam release from unaffected steam generators, 0 hours0 days 0 hours 0 weeks 0 months (lbm) 0 No change Steam release from unaffected steam generators, 0-2 hours (lbm) 300,116.1 No change Steam release from unaffected steam generators, 0-8 hours (lbm) 861,350.9 No change Steam release from unaffected steam generators, 0-24 hours (lbm) 1,971,677.3 No change Steam release from unaffected steam generators, 0-53.2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months (lbm) 3,582,768.8 No change Steam release from unaffected steam generators, 0-98.8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months (lbm) 3,582,768.8 No change EAB dispersion factors (sec/m3) Enc. 2, Table 7 No change 3

LPZ dispersion factors (sec/m ) Enc. 2, Table 7 No change CR dispersion factors, PORVs (sec/m3) Enc. 2, Table 7 No change CR dispersion factors, MSL (sec/m3) Enc. 2, Table 7 No change 1

The CLB analysis also included cases with 85 cfm and 500 cfm unfiltered inleakage, but only the 300 cfm case was retained for this submittal.

Table 6. Rod Ejection - Dose Consequence Analysis Inputs and Assumptions

- - Note that all assumptions from Regulatory Guide 1.183 for the REA are applicable for this analysis, unless otherwise stated in this table. Also note that the current license basis does not include a dose analysis for the rod ejection accident; therefore there are no current license basis values to display.

CLB Updated Input or Assumption Description Value Value Reactor power with uncertainty (MW) NA 2346 See Table 2 of Source Term (Ci) NA Enclosure 1 Percentage of rods experiencing DNB NA 10%

Peaking factor NA 1.8 Number of rods that can exceed 6.3 kW/ft and experience DNB NA 0 Release duration (hr) NA 53.2 Reactor coolant specific activity (uCi/g) NA 0.5 Secondary side specific activity (uCi/g) NA 0.1 Control room volume (ft3) NA 20,124 Control room outside air makeup (cfm) NA 400 Control room ventilation normal mode Iodine intake filter efficiencies (%) NA 0 5 minutes, 35 Control room ventilation emergency mode start (containment release) NA seconds Control room ventilation emergency mode start (secondary release) NA 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Control room ventilation emergency mode recirculation (cfm) NA 2600 Control room ventilation normal mode Iodine intake filter efficiencies NA (95, 99, 95)

(elemental, particulate, organic %)

Control room ventilation normal mode unfiltered inleakage (cfm) NA 300 Control room ventilation pressurized mode unfiltered inleakage (cfm) NA 230 Containment volume (ft3) NA 1,958,526 Containment cooling fans combined flow rate start time (sec) NA 76 Containment cooling fans combined flow rate (cfm) NA 65,000 Natural Deposition Removal Coefficient (hr-1) NA 0.1 Containment leak rate, first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months (percent weight / day) NA 0.1 Containment leak rate, after first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months (percent weight / day) NA 0.05 Reactor coolant volume (ft3) NA 8,254 Secondary side volume (ft3) NA 4,262 Primary to Secondary Leak Rate (gpd) NA 450 Steam release from steam generators, 0-2 hours (lbm) NA 301,967.3 Steam release from steam generators, 2-8 hours (lbm) NA 566,768.3 Steam release from steam generators, 8-24 hours (lbm) NA 1,124,995.8 Steam release from steam generators, 24-53.2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months (lbm) NA 1,637,910.1 EAB dispersion factors (sec/m3) NA Enc. 2, Table 7 LPZ dispersion factors (sec/m3) NA Enc. 2, Table 7 CR dispersion factors, containment release, cont nearest point (sec/m3) NA Enc. 2, Table 7 CR dispersion factors, secondary release, MSSV/PORVs (sec/m3) NA Enc. 2, Table 7

Table 7. Dispersion Factors (sec/m3)

Location 0-2 Hours 2-8 Hours 8-24 Hours 24-96 Hours96-720 Hours EAB 1.77E-03 1.77E-03 1.77E-03 1.77E-03 1.77E-03 LPZ 8.92E-05 3.50E-05 2.19E-05 7.95E-06 1.85E-06 Containment 4.15E-03 2.74E-03 1.17E-03 8.18E-04 6.74E-04 RHR HX Room 7.13E-03 5.49E-03 2.29E-03 1.71E-03 1.37E-03 Inside FHB 1.24E-03 8.97E-04 3.62E-04 2.58E-04 2.14E-04 MSSV/PORVs 2.60E-03 1.65E-03 7.22E-04 4.97E-04 4.01E-04 MSL 2.48E-03 1.57E-03 7.05E-04 4.74E-04 3.93E-04

United States Nuclear Regulatory Commission Enclosure Ill to Serial: RNP-RA/16-0057 36 Pages (Including cover sheet)

Markups to the TRM, Technical Specification Bases and UFSAR

Decay Time 3.12 3.12 DECAY TIME TRMS 3.12 (CTS 3.8.1.h)

APPLICABILITY: MODE 6.

COMPENSATORY MEASURES CONDITION REQUIRED COMPENSATORY MEASURE COMPLETION TIME A. Requirements of A.1 Suspend movement of fuel within Immediately TRMS not met. the core.

TEST REQUIREMENTS TEST FREQUENCY None. NA HBRSEP Unit No 2 3.12-1 PLP-100 Rev. O

Decay Time B 3.12 B 3.12 DECAY TIME BASES The restriction of not moving fuel in the reactor for a pe od of "fflG.hou after shutdown reduces the consequences of a fuel handling accident b p o "di g r decay of short-lived fission products and the reduction of fission gas inventory 1n any potentiafl ii fuel. Fuel handling accidents in containment and the Spent Fuel Building av t> e evaluated by postulating that the failure of all fuel rods in one assembly o curs 100-hours after shutdown.

HBRSEP Unit No 2 B 3.12-1 PLP-1 00 Rev. 0

Containment Ventilation Isolation Instrumentation B 3.3.6 BASES APPLICABLE The containment ventilation isolation radiation monitors SAFElY ANALYSES ensure closing of the ventilation isolation valves. They are the primary means for automatically isolating containment in the event of a fuel handling accident during shutdown. Containment isolation in turn ensures meeting the containment leakage rate assumptions of the safety analyses, and ensures that the calculated accidental offsite radiological doses are below 10 CFR 50.67 limits. Due to radioactive decay, containment is only required to isolated

uel handling accidents involving handling recently irradiated fue I.e., e t has occupied part of a critical reactor core within the previ uS'"S&ilou ~ ~

The containment ventilation isolation instrume ion satisfies Criterion 3 of the NRC PoUcy Statement.

LCO The LCO requirements ensure that the Instrumentation necessary to initiate Containment Ventilation Isolation, listed in Table 3.3.6-1, is OPERABLE.

1. Manual lnitiatjon The LCO requires two channels OPERABLE. The operator can initiate containment ventilation isolation at any time by using either of two pushbuttons In the control room. Either pushbutton actuates both trains. This action will cause actuation of Phase A and Containment Ventilation Isolation automatic containment isolation valves. Containment Ventilation Isolation can also be initiated by the manual Containment Spray buttons.

The LCO for Manual Initiation ensures the proper amount of redundancy is maintained in the manual actuation circuitry to ensure the operator has manual initiation capability.

Each channel consists of one push button and the interconnecting wiring to the actuation logic cabinet.

2. Automatic Actuation Logic and Actyation Relays The LCO requires two trains of Automatic Actuation Logic and Actuation Relays to be OPERABLE. The (continued)

HBRSEP Unit No. 2 B 3.3-122 Revision No. 31

Containment Ventilation Isolation Instrumentation B 3.3.6 BASES LCO 2. Automatic Actuation Logic and Actuation Relays (continued)

Automatic Actuation Logic and Actuation Relays actuate containment ventilation isolation upon receipt of an actuation signal from the Containment Radiation or Manual Initiation Functions. Containment ventilation Isolation also initiates on an automatic safety injection (SI) signal when operating in MODES 1, 2, 3, and 4. The Bases for LCO 3.3.2, "Engineered Safety Features Actuation System (ESFAS) Instrumentation," discusses this mode of initiation.

3. Containment Radiation The LCO specifies two required channels of radiation monitors to ensure that the radiation monitoring instrumentation necessary to initiate Containment Ventilation Isolation remains OPERABLE.

For sampling systems, channel OPERABILITY involves more than OPERABILITY of the channel electronics. OPERABILITY may also require correct valve lineups, sample pump operation, and filter motor operation, as well as detector OPERABl LITY. if these supporting features are necessary for trip to occur under the conditions assumed by the safety analyses.

4. Safety Injection Refer to LCO 3.3 2, Functions 1.a-f, for all initiating Functions and requirements.

APPLICABILITY The Manual Initiation, Automatic Actuation Logic and Actuation Relays, and Containment Radiation Functions are required to be OPERABLE in MODES 1, 2. 3, and 4, or movement of recently irradiated fuel assemblies (i. , e as occupied part of a critical reactor core within the pre us-56-h s) within containment. The Safety Injection Functions ar re uired be during MODES 1, 2, 3, and 4. Under these conditions, the

exists for an accident that could release significant fission p ctuct radioactMty (continued)

HBRSEP Unit No. 2 B 3.3-123 Revision No. 22

Containment Penetrations B 3.9.3 BASES (continued)

APPLICABLE During movement of irradiated fuel assemblies within SAFETY ANALYSES containment, the most severe radiological consequences result from a fuel handling accident involving handling recently irradiated fuel. The fuel handling accident is a postulated event that involves damage to irradiated fuel (Ref. 1). Fuel handling accidents analyzed include dropping a single irrad iated fuel assembly and handling tool or a heavy object onto other irradiated fuel assemblies. The requirem CO 3.9.6, "Refueling Cavity Water Level," and irradiated fuel vement containment closure capability or a minimum decay ti of 56 h without containment closure capability ensure th t t se of fission product radioactivity, subsequent to a fuel handling acciden results in doses that are well within (S 25%)the dose limits specified in 10 FR 50 .67.

11 6 Containment penetrations satisfy Criterion 3 of the NRC P Statement.

LCO This LCO limits the consequences of a fuel handling accident involving handling recently irradiated fuel in containment by limiting the potential escape paths for fission product radioactivity released within containment. The LCO requires any penetration providing direct access from the containment atmosphere to the outside atmosphere to be closed except for the OPERABLE containment ventilation penetrations. For the OPERABLE containment ventilation penetrations, this LCO ensures that these penetrations are isolable by the Containment Ventilation Isolation System. The OPERABILITY requirements for this LCO ensure that the automatic containment ventilation valve closure times specified in the UFSAR can be achieved and, therefore, meet the assumptions used in the safety analysis to ensure that releases through the valves are terminated, such that radiological doses are within the acceptance limit.

APPLI CABILITY The containment penetration requirements are applicable during movement of recently irradiated fuel assemblies within containment because this is when there is a potential (continued)

HBRSEP Unit No. 2 83.9-10 Revision No. 31

Containment Penetrations B 3.9.3 BASES (continued)

APPLICABILITY for the limitting fuel handling accident. In MODES 1, 2, 3, (continued) and 4, containment penetration requirements are addressed by LCO 3.6.1. In MODES 5 and 6, when movement of irradiated fuel assemblies within containment is not being conducted, the potential for a fuel handling accident does not exist. Additionally, due to radioactive decay, a fuel hand li n ccident involving handling fuel that was not recently irradiated (i. . , e has not occupied part of a critical reactor core within the prev* us 5 h s) will result in doses that are well within the guideline value speci n 10 CFR 50.67 even without containment closure capability. under these conditions no requirements are placed on containment pe etration status.

ACTIONS 116 If the containment equipment hatch, air lock, or any containment penetration that provides direct access from the containment atmosphere to the outside atmosphere is not in the required status, including the Containment Ventilation Isolation System not capable of automatic actuation when the containment ventilation valves are open, the unit must be placed in a condition where the isolation function is not needed. This is accomplished by immediately suspending movement of recently irradiated fuel assemblies within containment. Performance of these actions shall not preclude completion of movement of a component to a safe position.

SURVEILLANCE SR 3.9.3.1 REQUIREMENTS This Surveillance demonstrates that each of the containment penetrations required to be in its closed position is in that position. The Surveillance on the open ventilation valves will demonstrate that the valves are not blocked from closing. Also the Surveillance will demonstrate that each valve operator has motive power, which will ensure that each valve is capable of being closed by an OPERABLE automatic containment ventilation isolation signal.

The Surveillance is performed every 7 days during movement of recently irradiated fuel assemblies within containment. This Surveillance ensures that a postulated fuel handling (continued)

HBRSEP Unit No. 2 B 3.9-11 Revision No. 22

Refueling Cavity Water Level B 3.9.6 B 3.9 REFUELING OPERATIONS B 3.9.6 Refueling Cavity Water Level BASES BACKGROUND The movement of irradiated fuel assemblies with in containment requires a minimum water level of 23 ft above the top of the reactor vessel flange. During refueling, this maintains sufficient water level in the containment, refueling canal, fuel transfer canal, refueling cavity, and spent fuel pool. Sufficient water is necessary to retain iod ine fission product activity in the water in the event of a fuel handling accident (Ref. 1). Sufficient iodine activity would be retained to limit offsite doses from the accident to within Regulatory Guide 1.183 and 10 CFR 50.67 limits (Refs. 2 and 3)

APPLICABLE During movement of irradiated fuel assemblies, the water SAFETY ANALYSES level in the refueling canal and the refueling cavity is an initial condition design parameter in the analysis of a fuel handling accident in containment (Ref. 1). A minimum water level of 23 ft allows a decontamination factor of 200 to be used in the accident analysis for iodine. Therefore, consistent with Regulatory Guide 1.183, Appendix 8.2, the overall effective iodine decontamination factor is 200 for the refueling cavity, with a resulting chemical species released from the water of 57% elemental and 43%

organic iodine (Ref. 1). 116 The fuel handling accident ana 1s inside containment is described in Reference 1. Wi h 1 m water level of 23 ft and a minimum decay tim of 56 h rs prior to fuel handling, the analysis and test programs d onstr that the iodine release due to a postulated fuel handli id nt is adequately captured by the water and offsite doses are maintained within allowable limits (Refs. 2 and 3).

Refueling cavity water level satisfies Criterion 2 of the NRC Policy Statement.

(continued)

HBRSEP Unit No. 2 B 3.9-21 Revision No. 22

Containment Purge Filter System B 3.9.7 B 3.9 REFUELING OPERATIONS B 3.9.7 Containment Purge Filter System BASES BACKGROUND The Containment Purge Filter System filters airborne radioactivity released to the containment following a fuel handling accident involving handling recently irradiated fuel in the containment. During refueling outages, the Containment Purge Filter System, in conjunction with other normally operating systems, also provides environmental control of temperature and humidity in the containment.

The Containment Purge Filter System is a single train system which consists of a prefilter, a high efficiency particulate air (HEPA) fi lter, an activated charcoal adsorber section for removal of gaseous activity (principally iodines), and two fans (only one of the fans is required, the second fan is a spare). Ductwork, valves or dampers, and instrumentation also form part of the system.

The Containment Purge Filter System is a manually intitiated system, which may also be operated during normal plant operations.

The Containment Purge Filter System is discussed in the UFSAR, Sections 6.5.1, 9.4.3, and 15.7.4 (Refs. 1, 2, and 3, respectively) because it may be used for normal, as well as post accident, atmospheric cleanup functions.

APPLICABLE The containment purge filter system is not used for SAFETY ANALYSES mitigation of the fuel handling accident as described In UFSAR Section 15.7.4. This system is required to be OPERABLE and in operation during the movement of recently irradiated fuel 1.e., t at has occupied part of a critical reactor core within the previ s 56 ho rs).

In the event of a fuel handling accident involving recently i diated el, the containment purge filter system, in conjunction with the o ent ventilation isolation requirements of LCO 3.3.6 and the containment closure requirements of LCO 3.9.3, would significantly impede the radioactive release.

(continued)

HBRSEP Unit No. 2 B 3.9-24 Revision No. 22

Containment Purge Filter System B 3.9.7 BASES APPLICABLE The Containment Purge Filter System satisfies Criterion 3 of the SAFETY ANALYSES NRC Policy Statement.

(continued)

LCO The Containment Purge Filter System is required to be OPERABLE and operating. When the Containment Purge Filter System is in operatilon, the exhaust flow from containment shall discharge through the HEPA and impregnated charcoal filters.

The Containment Purge Filter System is considered OPERABLE when:

a. One fan is OPERABLE;

b. HEPA filter and charcoal adsorber are not excessively restricting flow, and are capable of performing their filtration function; and

c. Ductwork, valves, and dampers are OPERABLE, and air flow can be maintained.

APPLICABILITY During movement of recently irradiated fuel in the containment, the 116 Containment Purge Filter System is required to be OPERABLE and operating to alleviate the consequences of a fuel handling accident involving handling recently irradiated fuel (i.e., fuel that has

~~~l\Qied part of a critical reactor core within the previous urs).

ACTIONS A-1 and A-2 When the Containment Purge Filter System is inoperable or not in operation during movement of recently irradiated fuel assemblies in containment, Required Action A.1 requires each penetration providing direct access from the containment atmosphere to the outside atmosphere to be immediately (continued)

HBRSEP Unit No. 2 B 3.9-25 Revision No. 22

HBR2 UPDATED FSAR 15.0.12 COMMON DOSE CONSEQUENCE INPUTS FOR ALTERNATIVE SOURCE TERM (AST) ANALYSES

!SCALE computer code package 15.0.12.1 Source Tenns

{t Core inventory isolopics were developed using a bounding approach. The 6RIGEN S computer code (Reference 15.0.12-8) was used to develop lsotopics for a variety of burnups, enrichments, and burnup rates (power levels). A plant specific set of high burnup equilibrium fuel cycles were postulated to cover wide variations in cycle energy, enrichment, and batch sizes. The resulting ORIGEN..S calculated lsotopics were Increased to account for the 2% measurement uncertainty above the then-current licensed power of 2300 MWt to bound operation at power levels (including allowances for measurement uncertainty} of up to 2346 MWt. These adjusted Isotopic inventories were then compared to a bounding DOE/RW 0184, R1 {July 1992) LWR lsotopics Characteristics database of other possible isotopic Inventories that were developed using generic Westinghouse-style 15x15 fuel descriptions. A bounding , conservative inventory was chosen from this full, composite set of isotopics. Therefore, the core inventory used in the AST dose analyses should bound any fuel cycles up to 5 w/o enrichment, 2346 Mwt core power (including adjustments for measurement uncertainty), and 18 month cycle length. The core Inventory used in the AST analyses that Involve fuel damage is provided in Table 15.0.12-1.

Certain analyses require the use of RCS isotopic concentrations at the Technical Specifications limits. As specified in Reference 15.0.12-3 for iodine spiking considerations, certain events have been analyzed at higher RCS radionuclide concentrations. Unless otherwise noted in the individual analysis discussions, these events start from the RCS Inventory In Table 15.0.12-2.

Similarly. unless otherwise noted in the individual analysls discussions, for those events which consider releases from the secondary system, the seoondary system radionucfide concentrations at the Technical Specifications limits are used, as shown in Table 15.0.12p3.

Regulatory Gulde (RG) 1.183 requires that certain fuel design criteria be met in order to use the RG specified release fractions {Section 3.2 of the RG, Footnotes 10 and 11 ). Specifically, peak fuel burnup should not exceed 62,000 MWO/MTU (Footnote 10) and the maximum linear heat generation rate should not exceed 6 3 kw/ft peak rod average power for bum ups exceeding 54 GWD/MTU (Footnote 11 ). Using the bounding fuel cycle specifted to develop the AST core inventory, the maximum discharge batch exposure was 60,000 MWDIMTU . Also, tliis somce te1111 basis foel cycle shows tl1at foel b8tches ~with average bt1mt1ps in excess of 54,000 MWd/MTU haoe lieat ge11e1atio11 1ates less Hia11 e.a kw/ft at 2366 MWt rated t11e1mal powe1.

Applying the 2% increase to boand the App9ndix I( Measaren1e11t Uncertainly Recovery (MUR)

JH3Wer ~JH:ete iAereeses this pareMeter to 6 °12 kw/ft . Therefore, both Regt1latory Gt1ide footnote restrletieAs are t:Aet 15.0.12.2 Other Common lnout§ Table 15.0.12-4 presents Control Room input parameters and Table 15.0.12-5 presents breathing rates and occupancy factors used in the dose analyses 15.0.12-1 Revision No 20

HBR2 UPDATED FSAR TABLE 15.0.12-1 CORE RADIONUCLIDE INVENTORY @ T = 0, 2346 Mwt

'\otope Curies lsotooe Curies Isotope Curie~

Co-~ 5.99E+05 Ru-103 9.87E+07 Cs-136 3.s?hoo Co-60 '\_ 4 .58E+05 Ru-105 6.83E+-07 Cs-137 Jla'1 E+06

/

Kr-85 ~.30E+05 Ru-106 3.73E+07 Ba-139 / 1.15E+08 Kr-85m 1 .~ +07 Rh-105 6.33E+07 Ba-14d 1.13E+08 Kr-87 3.03~7 Sb-127 5.38E+06 La..140 1.17E+08 Kr-88 4.20E+of\ Sb-129 2.03E+07 ~-141 1.02E+08 Rb-86 1.16E+05 1&.-121 5.31E+o i La-142 9.83E+07 Sr-89 5.90E+07 Te-f um s.ar_i+'os Ce-141 1.04E+08 Sr-90 6.16E+06 Te-12e'\. v.9"o e+o1 Ce-143 9.55E+07

'\

Sr-91 7.39E+07 Te-129m / ( 3.84E+06 Ge-144 8.18E+07 Sr-92 7.88E+07 Te-131.rf( ~+07 Pr-143 9.34E+07 Y-90 6.62E+06 Te-.f1'2 s.a1"eto1 Nd-147 4.17E+07 Y-91 7.69E+07 Jtf31 6.20E+~ NP-239 1.25E+09 Y-92 7.93E+O; / 1-132 9.02E+07 r'Ru-238 2.61E+06 Y-93 6.01&1o1 1-133 1.28E+08 P~9 2.44E+04 Zr-95 1..eS'e+os 1-134 1.41E+08 Pu-2~ 3.55E+04 Zr-97 A .OOE+08 1-135 1.21E+08 Pu-241 "\ ~ 9.89E+06 Nb-95 / 1.06E+08 Xe-133 1.28E+08 Am-241 ~8E+04 Mo-99' 1.16E+08 Xe-135 3.68E+07 Cm-242 3.2~+06 i;{.ggm 1.03E+08 Cs-134 1.25E+07 Cm-244 3.88E+~

!Replace with Insert A I "

15.0.12-2 Revision No. 20

Insert A:

Isotope Curies Isotope Curies Isotope Curies Co58 5.990E+05 Ru103 1.019E+08 Cs136 3.712E+06 Co60 4.580E+05 Ru105 7.206E+07 Cs137 8.874E+06 Kr 85 8.246E+05 Ru106 3.835E+07 Ba139 1.148E+08 Kr 85m 1.653E+07 Rh105 6.632E+07 Ba140 1.112E+08 Kr 87 3.271E+07 Sb127 5.892E+06 La140 1.165E+08 Kr 88 4.377E+07 Sb129 1.824E+07 La141 1.040E+08 Rb 86 1.277E+05 Te127 5.800E+06 La142 1.004E+08 Sr 89 6.123E+07 Te127m 9.815E+05 Ce141 1.052E+08 Sr 90 6.475E+06 Te129 1.708E+07 Ce143 9.765E+07 Sr 91 7.690E+07 Te129m 3.276E+06 Ce144 8.487E+07 Sr 92 8.229E+07 Te131m 1.244E+07 Pr143 9.572E+07 Y 90 6.678E+06 Te132 9.045E+07 Nd147 4.176E+07 Y 91 8.006E+07 I131 6.339E+07 Np239 1.270E+09 Y 92 8.324E+07 I132 9.271E+07 Pu238 2.819E+05 Y 93 9.377E+07 I133 1.294E+08 Pu239 2.610E+04 Zr 95 1.075E+08 I134 1.453E+08 Pu240 3.681E+04 Zr 97 1.075E+08 I135 1.235E+08 Pu241 1.014E+07 Nb 95 1.084E+08 Xe133 1.295E+08 Am241 1.281E+04 Mo 99 1.176E+08 Xe135 4.161E+07 Cm242 3.383E+06 Tc 99m 1.042E+08 Cs134 1.294E+07 Cm244 4.427E+05

HBR2 UPDATED FSAR TABLE 15.0.12-2 RCS EQUILIBRIUM ACTIVITY LIMITED TO e.z5 ~Cb'6RAM DOSE EQUIVALENT 1-1 S1 PRIOR TO THE ACCIOENT, 2946 MVo't Change this to read:

SGTR Activity (uCi/g)

(based on 0.25 uCi/g) lsotone µCl/GRAM Kr-85 5.41E-01 1.06E+OO Kr-85m 1.30E-01 2.54E-01 Kr-87 8.91E-02 1.74E-01 Add these Add another column here, 6.27E-01 values to the with a heading that says: Kr.-88 3.20E-01 Non-SGTR Activity (uCi/g) new column (based on 0.5 uCi/g)

Rb-86 N~lfalble !Negligible 1-131 1.93E-01 3.78E-01 t-132 7.12E-02 1.39E-01 1-133 3.11E-01 6.09E-01 1-134 4.37E-02 8.56E-02 3.27E-01 1-135 1.67E-01 Xe-133 2.14E+01 4 .18E+01 Xe-135 5.88E-01 1.15E+OO Cs-134 2.0SE-02 4 .03E-02 Cs-136 2.96E-03 5.79E-03 2.19E-01 Cs-137 1.12E-01 15.0.12-3 Revision No. 20

HBR2 UPDATED FSAR TABLE 15.0.12*3 SECONDARY COOLANT SYSTEM EQUILIBRIUM ACTIVITY Replace with these Isotope SGTR NON-SGTR updated values EVENTS

µCl/GRAM 1-131 7.72E-02 7.55E-02 1-132 2.85E-02 2.79E-02 1-133 1.25E-01 1.22E-01 1.71E-02 1-134 1.75E-02 6.54E-02 1-135 6.69E-02 Cs-134 2.06Ee-03 4.03E-03 Cs-136 2.96E-04 5.79E-04 Cs-137 2.19E-02 1.12E-02 Note: This table presents the secondary coolant system equilibrium activity that was assumed for the Steam Generator Tube Rupture event and for other events resulting in the release of secondary side activity. The iodine nuclide activity is based on the Technical Specifications limit for Dose Equivalent lodlne-131 of 0.1 µCi/gm for the secondary side Foi the SSTR, the cesit:rm nt:1clide aeti1.1ity is based on 19% of the RCS eeeit1m eetivit,*

eaffes~eRdiAg to e TeehRieel SpeelfieatieAs lifflit ef 0.26 t:iCblgm Qese ec:t1:1i11aleAt I 131 .

For the non*SGTR e11ents, the cesit1m nt1elide eetivit)' ie based o" 10% of the RGS eeeitim activity eerreepe,,ding to e Teehnieel Speeifieetie,,s limit ef 9.59 ttGtlgm Dase EEt1:1ivelent 1=131 .

15.0.12--4 Revision No. 20

HBR2 UPDATED FSAR (2) For the case of the accident Induced Iodine spike, the postulated MSLB event induces an iodine spike. RCS activity is conservatively assumed to be initially at twice the activity presented in Table 15.0.12-2. Iodine is released from the fuel into the RCS at a rate of 500 times the iodine equilibrium release rate for a period of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . For fuel cycles where fuel breach may occur, a conservative, bounding analysis was performed to assess the Impact of postulating the breach of 2 fuel assemblies durfng a MSLB accident.

Other assumptions used in the radiological consequence analysis:

Coincident loss of offsite power.

157 fuel assemblies in the core.

Fuel gap inventory fractions.

Krypton 85: 10%

Other Noble Gases: 5%

Iodine 131: 8%

Other Halogens: 5%

Alkali Metals (Cs, Rb): 12%

Volume of the flu id of the RCS is 8254 ft3 (minimum volume, used to determine RCS concentration) and 9623 ft3 (maximum volume, used to determine iodine equilibrium appearance rate) at 575.9°F and 2235 psig.

RCS activity conservatively remains constant throughout the Pre-Accident Iodine Spike MSLB and the 2 failed fuel assembly events (no dilution of the RCS activity from the safety injection system is considered).

RCS mass remains constant throughout the MSLB event (no change In the RCS mass as a result of the MSLB or from the safety Injection system).

Data used to calculate the iodine equilibrium appearance rate:

Maximum Nominal Letdown Flow: 120gpm@130 °F, 2235 psig Uncertainty Applied to Letdown Flow: 10%

Maximum Identified RCS Leakage: 10 gpm Maximum Unidentified RCS Leakage: 1 gpm

Maximum radial peaking factor Is 1.8.

Tbe pi:lmai.y to-secoRdary leak rate in tRe ateam geAeratefS i9 based on the leak-rate-llmiting condition for operation specified in the Teehnieel Speeifieetions ef 75 gpd Increased by a factor of 2 (159 gpd, ~1thid1 is 9.1G4 gpm). The leakage Is e1313eFti0ned between the steam generators in st1ch a m21nne1 that the calculated dose is maximized.

The operational prin1ary*to-seeondary le21kage is consentath1ely asst1med to be 0.11 The primary-to-secondary leak rate in the steam generators is based on the maximum accident induced steam generator tube leakage criterion in the Technical Specifications of 150 gpd (0.104 gpm) per steam generator for a total of 450 gpd (0.3125 gpm). The analysis conservatively assigns leakage to the faulted steam generator of 0.115 gpm and 0.1975 gpm to the unaffected generators.

15.1.5-5b Revision No. 25

HBR2 UPDATED FSAR 15.1.5.5 Conclusions In a conservative estimation of the consequences, the shutdown margin is lost and the core returns to power. When the auxillary feedwater flow is terminated, heatup of the primary with resulting negative moderator and doppler feedback effects will augment the negative reactivity inserted from the boron to terminate the power excursion. Evaluation of the peak tuel Linear Heat Generation Rate shows that fuel centerline melting does not occur and core subchannel calculations show that DNB does not occur. The limiting cases are the Hot Zero Power case for Fuel Centerline Melting and Hot Full Power for MONBR.

For the MSLB with a pre--accident iodine spike, the 2-hour dose at the EAB is 0.26 rem TEDE.

The dose at the LPZ is 0.03 rem TEOE. The Control Room doses at inleakages of 300 and 500 cfrn are 0.14 and 0.21 rem TEDE, respectively.

For the MSLB with an accident induced iodine spike, the 2-hour dose at the EAB is 0.75 rem TEDE. The dose at the LPZ is 0.10 rem TEDE. The Control Room doses at in/eakages of ~

and 500 cfm are 0.45 and 0.73 rem TEDE. respectively. I0.38 I IL" l2EJ For the MSLB with a breach of two fu~our dose at the EAB is~

TEDE. The dose at the LPZ is 0:42' rem TEDE. The Control Room deees at iRleakages ef 390 rem and 500 efm ere 1.61 ene 2.71 rem TEOE, respeeth*ely.

The offsite dose acceptance criterion established by Refer ce 15.0.12-3 for the pre-accident iodine spike or fuel damage is that doses should be le han the 10 CFR 50.67 gu!deline of 25 rem TEDE. The offsite dose acceptance criterion blished by Reference 15.0.12-3 for the accident induced Iodine spike is that doses sh a be less than 10% of the 10 CFR 50.67 guideline, or Jess than 2.5 rem TEDE. Th ontrol Room dose acceptance criterion established by 10 CFR 50.67 for the MSLB is 5 re EDE.

Therefore, the offsite and Con Room TEDE doses due to a MSLB event meet the dose acceptance criteria.

!dose at an inleakage of 300 cfm is 1.53 Rem TEDE.

15.1.5-Sa Revision No. 24

HBR2 UPDATED FSAR 15.3.2.5 Radioloaical Conseguences The NRC has approved implementation of the Alternative Source Term methodology (Reference 15.0.12*3) for analysis of the radiological consequences of this event (Reference 15.3.2-1).

An instantaneous seizure of a reactor coolant pump is assumed to occur, rapidly reducing flow through the affected reactor coolant loop. Due to the pressure differential between the primary and secondary systems and assumed steam generator tube leakage, fission products are discharged from the primary into the secondary system. A portion of this radioactivity is released to the outside atmosphere from the secondary coolant system through either the steam generator atmospheric relief valves or safety valves. In addition, radioactivity is contained in the primary and secondary coolant before the accident and some of this activity is released to the atmosphere as a result of steaming from the steam generators following the accident. ~

This ev n result in fuel damage. In order to bound the maximum number of fuel as Iles expected to experience fuel clad damage, this anatysis conservatively assumes that

~semblies experience clad damage, but no fuel melting is assumed to occur.

Other assumptions used in the radiological consequence analysis:

Loss of offsite power at the time of reactor trip This drives the release from the secondary coolant system through the SG relief valves, since condenser cooling is lost

Maximum radial peaking factor is 1.8 r-j2l

157fuelassef!lblie~ ~

FOf thefrda~ fuel assemblies, the activity released from the fuel clad failure is based on the following gap inventory fractions*

Krypton 85: 10%

Other Noble Gases: 5%

Iodine 131 : 8%

Other Halogens: 5%

Alkali Metals {Cs, Rb): 12%

All gap activity in the damaged fuel rods is instantaneously released

The chemical form of the radioiodine released from the damaged fuel is 95% cesium iodide (Csl), 4.85% elemental iodine. and O 15% organic iodide

The minimum volume (hot) of the reactor coolant system is 8,254 ft3 , based on a temperature of 575.9°F and a pressure of 2235 psig.

The primary~to-secondary leakage to the steam generators mixes instantaneously and homogeneously with the secondary water without flashing .

15 3.2-3 Revision No. 20

HBR2 UPDATED FSAR

RCS equilibrium activity concentration is conservatively assumed to be twice the values in Table 15.0.12-2. Since fuel damage is assumed, no iodine spiking is assumed for the equilibrium RCS activity.

The primary-to-secondary leak rate is limited te 9.3 g~m tetal tAret:igA tRe 3 &teeFA geRerateFe whieh be1:1Ads TS limit ef 75 spd per 6/G.

The minimum volume of the secondary side coolant is 88,641 lbs per steam generator.

The integrated mass of steam released from the steam generators as a function of time Is 301,967.3 lbm (0- 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months ), 868,735.6 lbm (0 - 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ), 1,993,731 .4 lbm (0 - 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ), and 3,631,641.5 lbm (0- 53.2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months ).

The halogen and alkali metal partition coefficient for the steam generators is 100.

Iodine releases from the steam generators to the environment are 97% elemental and 3% organic.

All noble gas radionuclides released from the primary system are released to the environment without holdup, reduction, or mitigation.

The time required for one train of the RHR System to establish adequate shutdown cooling to terminate releases from the steam generators is 53.2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

15. 2.6 Cont;el1i1sjon ~ ~12.09 I For he locke~dent, the 2-hour dose at the EAB is r24-Tem TEDE. The dose at the LP is 0.21 rem TEDE. The Control Room doses at inle!lkages of 300 end 500 erm ere 9.8&

lished by Reference 15.0.12-3 for this accident Is tha doses should be less than 10% o 10CFR 50.67 guidelines, or less than 2.5 rem TEOE.

Th Control Room dose acceptanc criterion established by 10CFR 50.67 for this accident is 5 re TEOE.

Th refore, the offsite and Co I Room TEOE doses due to a locked rotor event meet the dose ac ptance criteria.

!dose at an inleakage of 300 cfm is 0 .68 Rem TEDE .

set to 150 gpd per steam generator for a total of 450 gpd.

15.3.2-3a Revision No. 25

The initial Input for the case analyzed is the same as that previously identified to provide the limiting transient response for the uncontrolled RCCA bank withdrawal at power. For the withdrawal of a single full-length RCCA, a maximum radial power peaking augmentation factor of 1.100 was used.

15.4.3.1.4 Radiological Consequences The NRC has approved implementatfon of the Alternative Source Term methodology (Reference 15.0.12-3) for analysis of the radiological consequences of this event (Reference 15.4.3-2).

The single RCCA rod withdrawal event causes an insertion of positive reactivity whrch results in a power excursion transient and may cause fuel damage. Due to the pressure differential between the primary and secondary systems and assumed steam generator tube leakage, fission products are discharged from the primary into the secondary system. A portion of this radioactivity is released to the outside atmosphere through either the steam generator atmospheric relief valves or safety valves. In addition, radioactivity is con1ained in the primary and secondary coolant before the accident, and some of this activity is released to the atmosphere as a result of steaming from the steam generators following the accident ,C_a_p-it-a-li-

.... ze- "t-h-is-,,.....1 This event can result in fuel damage. hr 01de1 to bow Id~~

assemblies ex~eetee ta experience ftlel elad damage, th1Sanatysis conservatively assumes that four a~semblies _experience clad damag~d that three of those fot1r assemblies else experience melttAfJ. !Change the comma to a period I Other assumptions used In the radiologlcat consequence analysis:

Loss of offsite power at the time of reactor trip This drives the release from the secondary coolant system through the SG relief valves, since condenser cooling is lost.

Maximum radial peaking factor is 1.8.

157 fuel assemblies in the core.

For the 4 damaged fuel assemblies, the activity released flout the fael elad breech end fael n1eltin9 Is based on the following fractions:

Krypton 85: 10% breech end 199% melt Other Noble Gases: 5% bteach end 109% melt Iodine 131

8% breach and *9% melt Other Halogens: 5% breech and 49% melt Alkali Metals (Cs, Rb): 12% breech a"d 39% melt All gap activity in the damaged fuel rods is instantaneously released

The chemical form of the radioiodine released from the damaged fuel is 95% cesium iodide (Csl), 4.85% elemental iodine, and 0.15% organic iodide.

The minimum volume (hot) of the reactor coolant system is 8,254 ft 3, based on a temperature of 575.9°F and a pressure of 2235 psig.

15.4.3-3 Revision No. 26

set to 150 gpd per steam generator for a total of 450 gpd .

The primary-tcrsecondary leakage to the steam generator mixes Instantaneously and homogeneously with the secondary water without flashin .

RCS equilibrium activity concentration is conservatively ssumed to be twice the values In Table 15.0.12-2. Since fuel damage is assumed, no dine spiking is assumed for the equilibrium RCS activity.

The primary-tcrsecondary leak rate is limltea te Q.3 gpm letal Uuewg~ tt:ie 3 s&ea~

ge"eratOf! whieh be1:2"ds TS limit ef 75 gpd per SIG.

The minimum votume ~f the secondary side coolant Is 88,641 lbs per steam generator.

The integrated mass of steam released from the steam generators as a function of time is 301,967.3 lbm (0 - 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months ), 868,735.6 lbm (0 - 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months ), 1,993,731.4 lbm (0- 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ). and 3,631 ,641 .5 lbm (0- 53.2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months ).

The halogen and alkali metal partition coefficient for the steam generators is 100.

Iodine releases from the steam generators to the environment are 97% elemental and 3%organic.

All noble gas radionuclides released from the primary system are released to the environment without holdup, reduction, or mitigation.

The time required for one train of the RHR System to establish adequate shutdown cooling to terminate releases from the steam generators is 53.2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

15.4.3.1.5 Conclusions I extreme The minim m DNB ratio calculated for the event is 0.885, which is less than the safety limit. The ra ial power peaking calculated for the single RCCA withdrawal Is localized in the neighborh d of the withdrawn RCCA. Only one of the 157 fuel assemblies in the core is calculated experience bolling transition. The peak pellet LHGR was calculated to be under the threshold Ii it.

CCA withdrawal event Is classified as a Condition Ill event. Less than 10 percent of the core e eriences boiling transition. Reactor vessel pressurization is well below 110 percent of the design imit. It is not anticipated that core cooling would be significantly hindered by less than 10 percent fuel failures. No more limiting fault is engendered by the occurrence of the event The result of th analysis is thus in conformance with the acceptance criteria for a Condition Ill event and is, the efore, acceptable. I0.05 I ~ !0.69 I For the si le RCCA ~ident, the 2-hour dose at the EAB is 1.16 Fem TEDE. The dose at th LPZ Is 9.2~-rem TEDE. The Control Room dases at iAloakages gf 300 aRd 600 Gfm The offsit dose acceptance criterion established by Referen e 15.0.12-3 for this accident is that dose should be less than 10% of the 10 CFR 50.67 gul elines, or less than 2.5 rem TEDE.

The Contr I Room dose acceptance criterion established by O CFR 50.67 for this accident is 5 remTED .

Therefore the offsite and Control Room TEDE doses duet a single RCCA withdrawal event meet the ose acceptance criteria.

dose at an inleakage of 300 cfm is 2.57 Rem TEDE.

Insert new bullet:

No credit is taken for control room ventilation switchover to emergency mode .

15.4.3-3a Revision No. 26

HBR2 UPDATED FSAR The key analysis conditions for the limiting case are summarized below:

Initial power 102% of 2300 MWt Ejected RCCA worth Bounding (maximum) value

[136.8 pcm)

Moderator temp. coefficient +5.Q pcm/°F Doppler coefficient -0.976 pcrnrF Delayed neutron fraction, ~ Bounding (minimum) value

[0.006252)

Pellet-to-clad HTC Bounding (maximum) core-average value 1387 BTU hr-ft2-0 f 15.4 .8.4 Analvsjs of Resy!ts The sequence of events for the analysis is given fn Table 15.4.8-1. The transient tripped the reactor on the high-flux reactor trip. The key system response parameters are shown in Figures 15.4.8-1through15.4.8-4.

The pellet energy deposition was conservatively evaluated for BOC and EOC, at HFP and HZP, using the SPC Generic Rod Ejection methodology. The results of this analysis show that the peak deposited energy is 169.6 ca!lg, which is less than the 280 cal/g limit as stated in Regulatory Guide 1.77.

15.4.8.6 CeRGIY&io~

The 1esults of tl1e a11alysis de111011sbate that the euent acceptance criteria ere met. The piedicted MDNSR is 1.1ag, Tl:li& i& greatertl:laR the 1141 DNS limit Tt:1e predicied peak enecgy depositio~ is less thaR U:le 280 eallg limit. n1er-efeFe, Ae ftlel feih:1res ere predicted to occur, end the-a i1 "~ aig"lfiHAI Fa8ielegiaal releeee a~e te t~ia e11eAt.

!Add Insert BI 15.4.8-3 Revision No. 26

Insert B:

15.4.8 .5 Radiological Consequences The NRC has approved implementation of the Alternative Source Methodology (Reference 15.0.12-3) for analysis of the radiological consequences of the event (Reference 15.4.8-2). The analysis follows Appendix H of Regulatory Guide 1.183.

This event can result in fuel damage. This analysis conservatively assumes that 10% of the rods in the core experience clad damage, but no fuel melting is assumed to occur.

Other assumptions used in the radiological consequence analysis:

Loss of offsite power at the time of reactor trip. This drives the release from the secondary coolant system through the SG relief valves, since condesner cooling is lost.

Maximum radial peaking factor is 1.8.

157 assemblies in the core.

For the 10% of the rods in the core that are damaged, the activity released from the fuel clad failure is based on the following gap release fractions :

Krypton 85: 10%

Bromine: 10%

Other Noble Gases: 10%

Iodine 131 : 10%

Other Halogens: 10%

Alkalie Metals (Cs, Rb): 12%

All gap activity in the damaged fuel rods is instantaneously released .

The minimum volume (hot) of the reactor coolant system is 8 ,254 ft3, based on a temperature of 575.9°F and a pressure of 2235 psig .

The primary-to-secondary leakage to the steam generators mixes instantaneously and homogeneously with the secondary water without flashing.

RCS equilibrium activity concentration is conservatively assumed to be twice the values in Table 15.0.12-2. Since fuel damage is assumed, no iodine spiking is assumed for the equilibrium RCS activity.

The primary-to-secondary leak rate is set to 150 gpd per steam generator for a total of 450 gpd.

The minimum volume of the secondary side coolant is 88,641 lbs per steam geenrator.

The integrated mass of steam released from the steam generators as a function of time is 301,967.3 lbm (0- 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months ), 868 ,735.6 lbm (0-8 hours), 1,993,731.4 lbm (0 -24 hours), and 3,631,641.5 lbm (0- 53.2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months ).

The halogen and alkali metal partition coefficient for the steam generators is 100.

Iodine releases to the environment are 97% elemental and 3% organic.

All noble gas radionuclides released from the primary system are released to the environment without holdup, reduction, or mitigation.

The time required for one train of the RHR System to establish adequate shutdown cooling to terminate releases from the steam generators is 53.2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

Containment leakage rate is 0.1% by weight of the containment air for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and then 0.05% by weight of the containment air thereafter.

Switchover to emergency control room pressurization is assumed to occur at 5 minutes 35 seconds for the containment release and one hour for the secondary release.

15.4.8 .6 Conclusion For the rod ejection accident, the 2-hour dose at the EAB is 4.07 rem TEDE. The dose at the LPZ is 0.69 rem TEDE. The Control Room dose at an inleakage of 300 cfm is 4.83 rem TEDE.

The offsite doe acceptance criterion established by Reference 15.0.12-3 for this accident is that doses should be less than 25% of the 10CFR50.67 guidelines, or less than 6.3 rem TEDE. The Control Room dose acceptance criterion established by 10CFR50.67 for this accident is 5 rem TEDE.

Therefore, the offsite and Control Room TEDE doses due to a rod ejection event meet the dose acceptance criteria .

HBR2 UPDATED FSAR REFERENCES. SECTION 15.4 ANF-89-151 (A), "ANF-Relap Methodololgy for Pressurized Water Reactors:

Analysis of Non*LOCA Chapter 15 Events," Advanced Nuclear Fuels Corporation, Richland , WA, April 1992.

15 4.1-2 NSAL-00-016, *Rod Withdrawal from Subcritical Protection in Lower Modes,"

Westinghouse, 2000.

15.4.3-1 EMF-96-029 (P)(A), Volumes 1 and 2, and Attachment, "Reactor Analysis System for PWRs Volume 1 - Methodology Discription, Volume 2 -

Benchmarking Results" January 1997.

15.4.3-2 NRC Letter dated September 24, 2004, "H. B. Robinson Steam Electric Plant, Unit No. 2 - Issuance of an Amendment on Full Implementation of the Alternative Source Term (TAC No. MB5105):

15.4.7-1 EMF-96-029 (P)(A), Volumes 1 and 2, and Attachment, "Reactor Analysis System for PWRs Volume 1 - Methodology Discription, Volume 2 -

Benchmarking Results" January 1997 15.4.8-1 XN-NF-78-44(NP)(A), "A Generic Analysis of the Control Rod Ejection Transient for PWR's," Exxon Nuclear Company, Richland, WA, October 1983.

INSERT:

15.4.8-2 Reference for the pending RNP 24-month cycle dose consequences SER .

15.4R-1 Revision No. 26

HBR2 UPDATED FSAR The free air volume of the unsprayed volume for spray train 'A' is:

1,958,526 ft3

0.171 = 334,907 .9 ft3 The free air volume of the sprayed volume for spray train 'B' is:

1,958,526 ft3

0.815=1,596,198.7 ft3 The free air volume of the unsprayed volume for spray train 'B' ts:

1,958, 526 ft3

0 .185 =362,327.3 ft3

5. Operation of the spray trains is as follows:

3 minutes spray initiates 77 minutes spray terminates 87 minutes spray initiates 167 minutes spray terminates 12 .] I

6. The elemental iodine spray removal coefficient is 20 hf1 for both spray trains. he I

j3.427 for the "B" spray train.

rn maximum allowable decontamination factor for the elemental iodine spray rem val coefficient is 200, which is achieved at 2.01 hours1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months for the sprar train and-2:eS-hours JC"' 3.484

7. ~ The particulate iodine spray removal coefficient is~ hr"1 for the GAn spray train and S:62rhr"1 for the "s* spray train. These spmy remtn*el coefficients incli:ide e remoual 1

coefficient of 0.2 hf for the effect of diffasiopl 101 esis. The maximum allowable decontamination factor for the particulate iodine spray removal coefficient Is 50, at which 12.76 I time the removal coe~ient Is reduced by a factor of 10, which is achieved at 2.66-hours for the "A" spray train nd 2-*hours for the Men spray train . ~ /

~

8. Two safety-related Containment cooling fans, at 65,000 cfm each, begin operation at 76 seconds. The mixing rate between the sprayed and unsprayed regions was assumed to be 65,000 cfm after 76 seconds.

9. Natural deposition Is credited In the containment sprayed volume (wllen sprays are not operational) and in the containment unsprayed region {including when sprays operational). The natural deposition removal coefficient is 0.1 hr'1 *

10. The maximum allowable primary containment leak rate per Technical Specifications Is 0.1 % by weight of the containment air for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months at P 1 of 42 psig. Per Reference 15.0.12-3, the leak rate is reduced after the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to 50% of the Technical Specification leak rate.

11 . The sump pH is maintained > 7.0 .

15.6 .5-6 Revision No. 21

HBR2 UPDATED FSAR 15.6.5.5.6 Other assumptions

1. The offsite and control room breathing rates are:

Offslte CEAB and LPZ> CR Time Rate (m3/sec) Time Rate (m;J/sec) 0-8 hr 3.5E-04 0-30davs 3.5E-04 8-24 hr 1.8E-o4 1-30 davs 2.3E-04

2. The control room occupancy factor is 1.0 for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , 0.6 for 1 through 4 days, and 0.4 for 4 days through 30 days.

3. The meteorological dispersion factors (XJQ) in seclm3 are:

Cont. RHR Nearest HX Point Room Time Period EAB LPZ CR CR 0-2 hours 1.77E-03 8.92E-05 4.15E-03 7.13E-03 2- 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 1.77E-03 3.SOE-05 2.74E-03 5.49E-03 8- 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 1.77E-03 2.1 9E-05 1.17E-03 2.29E-03 1-4days 1.77E-03 7.95E-06 8.18E-04 1.71E-03 4- 30 days 1.77E-03 1.85E-06 6.74E-04 1.37E-03 Results 122.5 j~ l4.40 I The ma um 2-hour dose at the EAB is 24.7 rem TEDE. Th~ose at the LPZ is+.62-rem TEDE. The Control Room dose ls-it:51-rem TEDE.

Contributing sources to the control room dose include:

Containment Leakage ESF Leakage Containment Shine External Radioactive Cloud

,. . 2. .3. . ., ,. . ,9E~+. . .,. O. .,. .,O / '

1.94E+OO 3.86E-02 3.67E-02 15.6.5-8 Revision No. 25

HBR2 UPDATED FSAR Table 15.6.5-4 Core Acflyity lso~~

Co-58 Co-60 Curles r-6.99E+05 4.58E+05 Isotopes Ru-103 Ru-105 Curies 9.87E+07 6.83E+07 Isotope s

Cs-136 c7

~2E+06 Cs-137 / " 8.87E+06

/

Kr-85 7.30~05 Ru-106 3.73E+07 Ba-138' 1.15E+08 Kr-85m 1.51E+Q:l Rh-105 6.33E+07 BaA'40 1.13E+08 Kr-87 3.03E+07 '>Sb-127 5.38E+06 ~-140 1.17E+08 Kr-88 4.20E+07 Sbc:129 2.03E+O'V La-141 1.02E+08 Rb-86 1.16E+05 Te-1'R7 5.31Etd6 La-142 9.83E+07 Sr-89 5.90E+07 Te-12Tht 8.81E+05 Ce-141 1.04E+08 Sr-90 6.16E+06 Te-129 l\..'kOOE+07 Ce-143 9.55E+07 Sr-91 7.39E+07 Te-129m V:NWE+06 Ce-144 8.18E+07 Sr-92 7.88E+07 Te-13tm 1.23£+07 Pr-143 9.34E+07 Y-90 6.62E+06 Te-Uf2 8.91 E~7 Nd-147 4.17E+07 Y-91 7.69E+07 1~ 1 6.20E+07' Ne>-239 1.25E+09 Y-92 7.93E+07 11-132 9.02E+07 'Ru-238 2.81E+06 Y-93 6.07E+07 1-133 1.26E+08 Pu~39 2.44E+04 Zr-95 1.05.i'+08 1-134 1.41E+08 Pu-249... 3.55E+04 Zr-97 1..00E+08 1-135 1.21E+08 Pu-241 " .... 9.89E+06 Nb-95 /l .06E+08 Xe-133 1.28E+08 Am-241 "l..18E+04 Mo-99/ 1.16E+08 Xe-135 3.68E+07 Cm-242 3.2'SE+06 Tc;9!Ml 1.03E+08 Cs-134 1.25E+07 Cm-244 3 . 88~5

/ ) j\

!Replace with Insert C I 15.6 5-13 Revision No. 21

Insert C:

Isotope Curies Isotope Curies Isotope Curies Co58 5.990E+05 Ru103 1.019E+08 Cs136 3.712E+06 Co60 4.580E+05 Ru105 7.206E+07 Cs137 8.874E+06 Kr 85 8.246E+05 Ru106 3.835E+07 Ba139 1.148E+08 Kr 85m 1.653E+07 Rh105 6.632E+07 Ba140 1.112E+08 Kr 87 3.271E+07 Sb127 5.892E+06 La140 1.165E+08 Kr 88 4.377E+07 Sb129 1.824E+07 La141 1.040E+08 Rb 86 1.277E+05 Te127 5.800E+06 La142 1.004E+08 Sr 89 6.123E+07 Te127m 9.815E+05 Ce141 1.052E+08 Sr 90 6.475E+06 Te129 1.708E+07 Ce143 9.765E+07 Sr 91 7.690E+07 Te129m 3.276E+06 Ce144 8.487E+07 Sr 92 8.229E+07 Te131m 1.244E+07 Pr143 9.572E+07 Y 90 6.678E+06 Te132 9.045E+07 Nd147 4.176E+07 Y 91 8.006E+07 I131 6.339E+07 Np239 1.270E+09 Y 92 8.324E+07 I132 9.271E+07 Pu238 2.819E+05 Y 93 9.377E+07 I133 1.294E+08 Pu239 2.610E+04 Zr 95 1.075E+08 I134 1.453E+08 Pu240 3.681E+04 Zr 97 1.075E+08 I135 1.235E+08 Pu241 1.014E+07 Nb 95 1.084E+08 Xe133 1.295E+08 Am241 1.281E+04 Mo 99 1.176E+08 Xe135 4.161E+07 Cm242 3.383E+06 Tc 99m 1.042E+08 Cs134 1.294E+07 Cm244 4.427E+05

HBR2 UPDATED FSAR overall DF of 200 that Is specified in Reference 15.7.4-3. The elemental DF was re-evaluated, in accordance with the relationships defined in Reference 15.7.4-5, for 21 feet of coverage.

Elemental OF was conservatively reduced from the generic, 23 foot coverage value of DF=285 to a 21 foot coverage value of DF*174. Organic DF (the other component of the overall OF) was specified in Reference 15.7.4-3 to be DF=1, which means that this chemical species is not reduced at all by the water coverage. Since this component is already at Its most conservative value, it is not affected by the change in water coverage from the generic 23 feet to 21 feet.

Recombining the elemental and organic components using the overall OF equation from Reference 15.7.4-5, the generic overall DF from Reference 15.7.4-3 ls reduced from DF=200 to DF=138.

The activity could be released either in the containment or in the Auxiliary (Fuel Storage)

Building. Ventilation systems in both areas are in operation under administrative control during refueling. In evaluating doses inside the structures, the assumption is made that the release is drawn directly into the ventilation system before substantial mlxing occurs. Radioactivity monitors would immediately indicate and alarm the increased activity level. Activity in the containment would automatically close the purge ducts, although the offsite dose was conservatively evaluated assuming that the entire radionuclide inventory would be released before containment isolation. In evaluating the dose to refueling personnel inside the containment, or inside the Fuel Storage Building, the lack of substantial mixing and existence of alarms that would cause a prompt evacuation, lead to the conclusion that the total personnel dose would be small. Following evacuation, re-entry to the buildings would be delayed or othe1Wlse planned using indicated radiation levels, and would factor in any airborne radioactivity cleanup that may have occurred.

The Alternative Source Term methodology of Reference 15.7.4-3 Is used to calculate offsite exposures using the RADTRAD Ve1sion 9.62 computer code (Reference 15.7.4-7). The analysis of the accident occurring in the containment does not credit containment ventilation filtration systems, in order to conservatively bound the consequences of the event which might occur with containment openings, such as the equipment hatch, personnel airlock, or other penetrations, not sealed. In accordance with Reference 15.7.4-3, all activity released to the containment atmosphere is released to the environment over a two hour period. The analysis of the accident occurring in the fuel handling building does apply credit for the fuel handling building ventilation and filtration systems. All activity which is released to the fuel handling building atmosphere is released over a two hour period through the ventilation and filtration systems to the environment. Conservatively. both the containment and fuel handling building releases are modeled as ground level releases for offsite dose analysis purposes.

HBR 2 specific Atmospheric Dispersion Factors (known as X/Q factors) for offsite dose consequence evaluations were developed from meteorological data gathered at HBR 2 during the 9 year time period of 1988 through 1996. The data was evaluated using the PAVAN computer code (described in Reference 15.7.4-9), which implements Regulatory Guide 1.145 methods (Reference 15.7.4-10). For on-site receptor locations (including such locations at the Control Room), the meteorological data from this same time period was evaluated using the ARCON96 computer code (described in Reference 15.7.4-6). Minor changes to ARCON96 default values were made to implement draft NRC guidance contained in Reference 15.7.4-8.

15.7.4-4 Revision No. 18

jSCALE computer code package HBR 2 I UPDATED FSAR HBR-2 specific iatopic source terms were developed using a bounding approach. The ORIGEN 6 eemJt1:1ter eede was used to develop isotopics for a variety of bumups, enrichments, and burnup rates (power levels). Sensitivity studies were run with various combinations of bumups and enrichments to identify a bounding single assembly isotopic source tenn. The assembly source term was multiplied by 1.02 to reflect operation prior to shutdown at 102% of rated power (102% of 2300 MWlh). A bounding local peaking factor of 1.8 was then applied to simulate the effect of this accident for the assembly containing the peak fission product Inventory.

The dose to the operator In the Control Room was evaluated for the implementation of the Reference 15.7.4-3 methodology. Table 15.7.4-3 provides the Control Room modeling parameters used as input for this evaluation. Control Room unfiltered air inleakage for these dose analyses was conservatively evaluated for a total of 300 cfm. Automatic switchover, backed up by confirmatory operator action, is credited at one hour to switch the Control Room ventilation from Nonnal alignment to Emergency Pressurization mode.

15.7.4.3 Radloloalcal Conseguences 15.7.4.3.1 Postulated fuel handling accident In the fuel handling building Using the assumptions listed in Table 15.7.4-1, and the Reference 15.7.4-3 analysis methods, the offsite and HBR 2 Control Room TEDE doses due to the FHA in the Fuel Handling Building are shown in Table 15.7.4-4 to meet the specified acceptance criteria 15.7.4.3.2 Postulated fuel handling accident inside containment Using the assumptions listed in Table 15.7.4-2, and the Reference 15.7.4-3 analysis methods, the offsite and HBR 2 Control Room TEDE doses due to the FHA In the Containment are shown in Table 15.7.4-4 to meet the specified acceptance criteria.

15.7.4-4a Revision No. 26

HBR2 UPDATED FSAR TABLE 15.7A-1 FUEL HANDLING ACCIDENT IN FUEL HANDLING BUILDING Add new assumption after 3 ASSUMPTIONS Number of pins that can exceed

1. Accident occurs at least-664iours after reactor shutdown 6.3 kW/ft over 54 GWD/MTU:

35

2. Plant thermal power (prior to shutdown) Is 2346 Mwth

3. All rods in one assembly rupture, releasing their gap activity

4. Bumup in affected assembly is bounded up to 60,000 MWD/MTU

5. Enrichment in affected assembly is bounded up to a nominal 4.95 weight percent U-235.

Higher enrichments up to 5.0 weight percent are bounded by sensitivity study results Add new assumption after 7

6. Assembly Peaking Factor: 1.8 Fraction of assembly activity in gap for rods over

7. Fraction of assembly activity in gap: 54 GWD/MTU and 6.3 kW/ft:

1-131 0.08 Cs-134 0.36 Kr-85 0.10 Cs-137 0.36 Other Noble Gases 0.05 Kr-85 0.30 Other Halogens 0.05 Alkali Metals 0.12

8. Iodine form split: In Clad Gap Above Pool Aerosol (Csl) 95% NIA Elemental 4.85% 57%

Organic 0.15% 43%

9. Pool OF (for 21 foot coverage):

Elemental Iodine 174 Organic Iodine 1 Effective (Iodine) 138 Noble Gases 1

10. 157 assemblies in the core

11. All activity released from the pool is exhausted as a ground level release over two hours to the environment through the FHB air handling and filtration system.

12. FHB air handling system filter efficiencies for iodine removal:

Elemental Iodine 90%

Organic Iodine 70%

15.7.4-5 Revision No. 18

HBR2 UPDATED FSAR TABLE 15.7.4-1 <Continued)

FUEL HANDLING ACCIDENT IN FUEL HANDLING BUILDING ASSUMPTIONS 3

13. Breathing Rate: {m /sec)

!jme Pedod EAB/LPZC1> Control Room 0-8 hr 3.SE-04 3.SE-04 8-24 hr 1.8E-04 3.5E-04 1-30 days 2.3E-04 3.SE-04

14. Atmospheric Dispersion Factors (X/Q): (sec/m3 )

Time Period .EAfi. ~ Control Room 0 - 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 1.77E-03 8. 92E-05 1.24E-03 15.

(1) - EAB =Exclusion Area Boundary

=

LPZ Low Population Zone 15.7.4-Sa Revision No 18

HBR2 UPDATED FSAR TABLE 15.7.4-2 FUEL HANDLING ACCIDENT INSIDE CONT A!NMENT ASSUMPTIONS Add new assumption after 3

1. Accident occurs at least-5&-hours after reactor shutdown Number of pins that can exceed 6.3 kW/ft over 54 GWD/MTU:

2. Plant thermal power (prior to shutdown) is 2346 M'Mh 35

3. AU rods in one assembly rupture, releasing their gap activ

4. Burnup in affected assembly is bounded up to 60,000 MWDIMTU

5. Enrichment in affected assembly is bounded up to a nominal 4 .95 weight percent U-235. Higher enrichments up to 5.0 weight percent are bounded by sensitivity study results

6. Assembly Peaking Factor: 1.8 Add new assumption after 7 Fraction of assembly activity in gap for rods over

7. Fraction of assembly activity in gap: 54 GWD/MTU and 6.3 kW/ft:

1-131 0.08 Cs-134 0.36 K~85 ~10 Cs-137 0.36 Other Noble Gases 0.05 Kr-85 0.30 Other Halogens 0.05 Alkali Metals 0.12

8. Iodine form split: In Clad Gap Above Pool Aerosol (Csl) 95% NIA Elemental 4.85% 57%

Organic 0.15% 43%

9. Cavity OF (for 23 foot coverage):

Elemental Iodine 500 Organic Iodine 1 Effective (Iodine) 200 Noble Gases 1

10. 157 assemblies in the core

11. All activity released from the cavity is exhausted as a ground level release over two hours to the environment

12. Containment air handling system and containment closure are not credited

13. Breathing Rate (m3/sec}.

Time Period EABILPz<1> Control Room 0-8 hr 3.5E-04 3.5E-04 8-24 hr 1.8E-04 3.SE-04 1-30 days 2.3E-04 3.5E-04 15.7.4-6 Revision No. 18

HBR2 UPDATED FSAR TABLE 15.7.4-2 lQontinuedl FUEL HANDLING ACCIPENT INSIDE CONTAINMENT ASSUMPTIONS

14. Atmospheric Dispersion Factors (X/Q): (sec/m3)

Tjme Period ~ LPZ Control Room 0 - 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 1.77E-03 8.92E-05 4.1 SE-03

15. Full Core Isotopic Activity: Curies at T=O Nuclide 6 .339E+07 1-131 1-132 9 .271 E+07 1-133 1.294E+08 1-134 1.453E+08 1-135 1.235E+08 Kr-85 8.246E+05 Kr-85m Kr-87 1.653E+07 Kr-88 3.271E+07 Xe-133 4 .377E+07 Xe-135 1.295E+08 4 .161E+07 (1) - EAB = Exclusion Area Boundary LPZ = Low Population Zone 15.7.4-Sa Revision No. 18

HBR2 UPDATED FSAR TABLE 1§.7.4-4 FUEL HANDLING ACCIDENT DOSE ANAL YS!S RESULTS EAB<11 LPz<2l CR 0ose<3l Crem TEDEl (rem TEDE> (rem TEDEl 14.34 I 10.22 I 13.ss 1 FHA Inside Containment ~ r:iMfl -&.30.. -4: FHA Inside FHB Regulatory Limit

-5:-1G ~ "6:29-10.011 ~ 10.33 6.3 6.3 5 I

1. Worst 2-hour integrated dose at Exclusion Area Boundary.

2. 30-day integrated dose at Low Population Zone.

3. Assumes a conservative unfiltered inleakage of 300 cfm for the first hour and 230 cfm thereafter.

15.7.4-8 Revision No. 18

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