RA-08-045, Response to Draft Request for Additional Information - Amergen Application to Revise Technical Specifications Regarding Secondary Containment Operability Requirements During Refueling

Response to Draft Request for Additional Information - Amergen Application to Revise Technical Specifications Regarding Secondary Containment Operability Requirements During Refueling

ML081330320
Person / Time
Site:	Oyster Creek
Issue date:	05/05/2008
From:	Cowan P AmerGen Energy Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
RA-08-045
Download: ML081330320 (13)

Text

Amer(5e SM AmerGen Energy Company, LLC www.exeloncorp.com An Exelon Company 2o0 Exelon Way Kennett Square, PA 19348 10 CFR 50.90 RA-08-045 May 5, 2008 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001 Oyster Creek Nuclear Generating Station Facility Operating License No. DPR-16 NRC Docket No. 50-219

Subject:

Response to Draft Request for Additional Information - AmerGen Application to Revise Technical Specifications Regarding Secondary Containment Operability Requirements During Refueling

Reference:

AmerGen Letter to USNRC, 'Technical Specification Change Request 338 -

Secondary Containment Operability Requirements During Refueling," dated November 2, 2007 This letter provides additional information in response to an NRC request for additional information (RAI) received via NRC facsimile, dated March 25, 2008, regarding Oyster Creek Nuclear Generating Station (Oyster Creek) Technical Specification (TS) Change Request 338 -

"Secondary Containment Operability Requirements During Refueling," submitted to the NRC for review on November 2, 2007 (Reference). The additional information is provided in the Enclosure.

AmerGen Energy Company, LLC (AmerGen) has reviewed the information supporting a finding of no significant hazards consideration that was previously provided to the NRC in the referenced document. The additional information provided in this submittal does not affect the bases for concluding that the proposed license amendment does not involve a significant hazards consideration.

No new regulatory commitments are established by this submittal. If any additional information is needed, please contact Mr. David Robillard at (610) 765-5952.

Response to Request for Additional Information - TSCR - 338 - Secondary Containment Operability Requirements During Refueling May 5, 2008 Page 2 I declare under penalty of perjury that the foregoing is true and correct. Executed on the 5th day of May, 2008.

Sincerely, Pamela B. C6wan Director - Licensing and Regulatory Affairs AmerGen Energy Company, LLC

Enclosure:

Response to Draft Request for Additional Information cc:

S. J. Collins, USNRC Administrator, Region I M. S. Ferdas, USNRC Senior Resident Inspector, Oyster Creek G. E. Miller, USNRC Project Manager, Oyster Creek P. Baldauf, Assistant Director, New Jersey Bureau of Nuclear Engineering

ENCLOSURE OYSTER CREEK NUCLEAR GENERATING STATION RESPONSE TO DRAFT REQUEST FOR ADDITIONAL INFORMATION TECHNICAL SPECIFICATION CHANGE REQUEST No. 338 "SECONDARY CONTAINMENT OPERABILITY REQUIREMENTS DURING REFUELING"

ENCLOSURE OYSTER CREEK NUCLEAR GENERATING STATION RESPONSE TO DRAFT REQUEST FOR ADDITIONAL INFORMATION TECHNICAL SPECIFICATION CHANGE REQUEST No. 338 "SECONDARY CONTAINMENT OPERABILITY REQUIREMENTS DURING REFUELING" By letter dated November 2, 2007, AmerGen Energy Company, LLC (AmerGen) submitted an amendment request for the Oyster Creek Nuclear Generating Station (Oyster Creek). The proposed amendment would revise the Oyster Creek Technical Specifications regarding secondary containment operability requirements during refueling.

The Nuclear Regulatory Commission staff has reviewed the information provided in support of the proposed amendment and finds that the following information is required to complete its review:

INTEGRITY OF FACILITY DESIGN BASIS

1. NRC Question For the Oyster Creek revised Fuel Handling Design Basis Accident (DBA) source term analysis done in accordance with 10 CFR 50.67, Alternative Source Term (AST) methodology, provide the current licensing basis (CLB) parameters along with the revised values where applicable, as listed in Table 4-3 of the November 2, 2007 submittal. In addition, provide the basis for any changes to the CLB parameters as a result of the proposed AST application for the fuel handling accident (FHA). The NRC staff requests that the licensee include CLB information and the revised parameters in a table whether or not the individual parameter changed for this amendment request (See regulatory positions 1.3.2 and 1.3.4 of Regulatory Guide (RG) 1.183 and NRC Regulatory Issue Summary (RIS) 2006-04).

Response

The AST parameters of Table 4-3 of the November 2, 2007 submittal are shown below with additional columns for the current licensing basis values and the basis for the changes from the current licensing basis to AST values provided in the November 2, 2007 submittal.

Enclosure Response to Request for Additional Information Page 2 of 10 Parameter or AST Value' Pre-AST and Current Justification and Method Licensing Basis Comments Reactor Power 1969 MWth 1930 MWth nominal This value is unchanged and includes 2% margin for 1969 MWth instrument uncertainty relative Appendix K to the rated thermal power of 1930 MWth.

Fuel Assembly 9 x 9 with 140 rods 8 x 8 with 124 rods The 9x9 array is the bounding Configuration and damaged damaged assumption for current Oyster properties Creek licensing basis FHA.

(Taken from GESTAR) There are 22 of the 8 x 8 fuel assemblies remaining in the reactor core (to be replaced during the next refuel outage).

Radial Peaking Factor 1.7 1.5 Conservative bounding assumption. RG 1.183 justifies this change.

Allowable Fuel Bumup and RG 1.183, Table 3.

N/A Some peak rod-average non-LOCA gap fractions Fuel bundle peak burnup burnups can exceed 54 will not exceed 62 GWD/MT. However, the 6.3 GWD/MTU. For fuel kW/ft linear heat generation rate exceeding 54 GWD/MTU, limit will not be exceeded. This the maximum linear heat data is reviewed for each cycle.

generation rate will not exceed 6.3 kW/ft.

FHA Radionuclide From the 60 isotopes TID-14844 Spent fuel source terms are Inventory forming the RADTRAD based on the same bounding library used in the Loss of reactor core source terms as Coolant Accident (LOCA) was used for the approved AST analysis, with decay to 24 LOCA analysis. The hours.- Gap activities are Radionuclide inventory from the per RG 1.183.

previous analysis was based on TID 14844 assumptions.

Underwater Noble Gases: 1 Noble Gases: 1 For conservatism, the effective Decontamination Factor minimum depth of 23 feet is Particulate (cesiums and Particulates: infinity assumed to. be the water rubidiums): infinity coverage over the reactor core.

This is the worst-case location Iodine: 200, corresponding Iodine: 100 for a fuel drop FHA to take to a 23-ft water depth for an place (significantly more assembly drop into the damage is produced than a reactor vessel shorter drop in the fuel pool).

Iodine chemical distribution The chemical form of TID-14844 From RG 1.183. Since the pH radioiodine released from of the pool water is not the fuel to the spent fuel maintained above 7, iodine is pool is assumed to be 95%

assumed to be 97% elemental Csl, 4.85% elemental and 3% organic in the air space iodine, and 0.15% organic above the pool. The design iodide. 95% CsI, basis FHA calculation uses only instantaneously dissociates gaseous iodine (no particulate).

in the pool water. \\All iodine is re-evolved as elemental iodine.

Activity Transport to the Activity reaching the refuel Elevated release via filtered All (99.9999%) activity is Environment floor airspace will essentially SGTS to Main Stack exhausted to the environment be all exhausted within 2 within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> per RG 1.183.

hours by using an artificially This also provides an allowance high exhaust rate.

for uneven mixing in the refuel floor airspace.

Release Pathways Activity reaching the refuel Filtration by SGTS with For conservatism, no credit is floor airspace will essentially elevated release through taken for filtration by the SGTS, be all exhausted within 2 the Main Stack.

or the elevated release resulting hours by using an artificially from exhaust through the Oyster high exhaust rate. The Creek Main Stack.

I release pathways are

Enclosure Response to Request for Additional Information Page 3 of 10 Parameter or.;ý.

AST Value; Pre-AST and Current Justification and Licensing Basis Comments Method....

Lie sn described in Table 4-1.

Dose Conversion Factors EPA Federal Guidance ICRP60 Dose conversion As recommended in RG 1.183 Reports 11 and 12 Factors, RG 1.109 and implemented in RADTRAD.

Offsite Dose Limit 6.3 rem TEDE 25 rem whole body After 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> per 10 CFR50.67 300 rem thyroid and RG 1.183 Control Room Dose Limit 5 rem TEDE for the duration 5 Rem Whole Body, or its Per 10 CFR50 App. A, GDC 19 of the accident equivalent to any, part, and 10 CFR50.67 CR Volume Volume 27,500 ft3 Volume 27,500 ft' For the AST values, as previously demonstrated for the CR Intake Rate 14,000 cfm 14,000 cfm approved LOCA analysis, the CR dose is maximized when an intake rate of 4,000 cfm is achieved (parametric evaluation previously performed). The maximum intake of 14,000 cfm in the purge mode is used for conservatism. This also eliminates the need to address the unfiltered inleakage rate, since additional intake produces no additional dose.

Refuel Floor Ventilation For this analysis, an artificial Exhaust flow via SGTS This new AST assumption Rate and Volume volume of 100 ft3 with an evacuates 99.9999% of all artificially high exhaust rate activity within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

is assumed for simplicity.

CR Potential Release Points Limiting X/Qs (0 - 2 hr)

Limiting X/Qs with Murphy The most limiting 0-2 hr Campe method (FHA) are potential release point was Drywell Access Facility 1.61 E-03 sec/m3 inappropriate for AST.

determined to be the MAC (West Wall)

Facility Personnel Airlock Drywell Access facility 1.93E-03 sec/m3 Pre-AST values used Murphy-(South Wall)

Campe methodology. New AST values utilize ARCON96 Commodities Penetration 1.77E-03 sec/m3 assumptions per RG 1.194.

on the RB South Wall Commodities Penetration 5.21 E-03 sec/m3 on the RB North Wall MAC Facility Personnel 6.75E-03 sec/m3 Airlock MAC Facility Entrance 6.62E-03 sec/m3 RB Roof Hatch 1.82E-03 sec/m3 Stack Tunnel Door 8.55E-04 sec/m 3

East Airlock Door 1.40E-03 sec/m 3 Reactor Building Wall 2.15E-03 sec/m3 (Diffuse Area)

Trunion Room Door to 3.73E-03 sec/m3 Turbine Building EAB Release Point Basis Normal RB exhaust stack Main Stack AST FHA releases are and Distance to EAB and 414 m (considered as considered as ground level applicable to all release 414 m releases. Previously approved locations) for TS Amendment No. 262, Limiting Dispersion dated April 26, 2007 (Reference Factors (0 - 2 hr) 1.41 E-03 sec/i 3 7.3).

Enclosure Response to Request for Additional Information Page 4 of 10 Parameter or AST Value Pre-AST and Current.

Justification and Method Licensing Basis Comments LPZ Release Point Basis Normal RB exhaust stack Main Stack AST FHA releases are and Distance to LPZ and 3218 m (considered as considered as ground level applicable to all release 3218 m releases. Previously approved locations) for TS Amendment No. 262, Limiting Dispersion Normal RB exhaust stack dated April 26, 2007 (Reference Factors (0 - 2 hr) 1.35E-04 sec/i 3 and 3218 m (considered as 7.3).

applicable to all release locations) 1.35E-04 sec/m3 Note: References cited in the Table are from the November 2, 2007 submittal.

ACCIDENT SOURCE TERM

2. NRC Question In the FHA AST assumptions AmerGen stated that the General Electric (GE)-1 1 fuel burnup is 27.6 giga-watt-days (GWD)/Metric Ton Uranium (MTU) so that the 6.3 kilo-watt (kW) / foot (ft) restriction outlined in footnote 11 of RG 1.183 does not apply. Please clarify that the GE 11 9X9 fuel, in the course of its projected power history for any specific fuel load, will not exceed the value of 54 GWD/MTU (See Regulatory Position 3.2 and footnote 11 of RG 1.183).

Response

Since peak rod-average burnups can exceed 54 GWD/MTU, AmerGen has incorporated alternative source term limits in its procedures and processes. Procedure NF-AB-1 10-2210, Core Loading Pattern Development, requires that the core loading be developed in accordance with NRC Regulatory Guide 1.183. There is a procedural step to ensure that 1) peak rod-average burnup of the fuel is less than 62 GWD/MTU, 2) that the maximum linear heat generation rate does not exceed 6.3 kilo-watt/ft for any bundle exceeding a peak rod burnup of 54 GWD/MTU, 3) that the cycle length not exceed 711 EFPD and 4) that the maximum radial peaking factor does not exceed 1.7.

3. NRC Question Given the information above to be correct, does Oyster Creek have other fuel types that would exceed 54 GWD/MTU and be more limiting for a FHA at Oyster Creek as far as the maximum fuel inventory at the end of fuel life (See Regulatory Position 3.2 and footnote 11 of RG 1.183).

Response

AmerGen has processes in place that require checks to be performed on peak rod-average exposures for all fuel types. The Oyster Creek core design engineer utilizes the 3D simulator code, PANACEA, to confirm compliance with NRC Regulatory Guide 1.183. For any fuel rod that exceeds 54 GWD/MTU rod average exposure, the kW/ft value is checked to confirm that the linear hear generation rate is less than 6.3 kW/ft. For Oyster Creek's current operating cycle (Cycle 21), maximum rod average exposure at end-of-cycle is 55.1

Enclosure Response to Request for Additional Information Page 5 of 10 GWd/MTU. The maximum linear heat generation rate for fuel rods projected to exceed 54 GWd/MTU is 5.1 kW/ft.

4. NRC Question Please clarify the term "Reactor Well" discussed in your submittal in Section 4.3 under the heading "Decontamination Factor."

Response

The term reactor well refers to the reactor cavity, which includes the reactor vessel and the volume of water that connects the reactor vessel to the spent fuel pool during refueling operations.

5. NRC Question In Section 4.3 of the submittal, AmerGen determined that "A drop over the reactor well is more limiting than accidents in the spent fuel pool (SFP)." Provide the reference or detail that shows the drop in the well or as described in the Oyster Creek final safety analysis report (FSAR), "a drop onto the reactor core from the maximum height allowed by the refueling equipment," as the limiting accident (See Regulatory Position 2.4 of RG 1.183).

Response

The Oyster Creek UFSAR Section 15.7.4.4 describes the number of fuel rodsthat are damaged during a design basis fuel handling accident. Dropping a fuel assembly onto the reactor core from the maximum height allowed by the refueling equipment (less than 30 feet) results in an impact velocity of 40 ft/sec. The kinetic energy acquired by the falling assembly is less than 17,000 ft-lb and is dissipated in one or more impacts. The first impact is expected to dissipate most (80%) of the energy and cause the largest number of cladding failures. This first impact is calculated to cause 105 fuel rods to fail. The second impact is calculated to cause 19 rods to fail. No rods are expected to fail during the third impact.

For a fuel assembly drop onto the spent fuel storage racks within the spent fuel pool, the number of cladding failures due to the initial impact is expected to be much less than the 105 calculated for the 30-foot drop. This is due to the fact that the kinetic energy due to a drop of 4 feet (a conservative height based on plant procedure restrictions of 2 feet) is less than that for a drop of about 30 feet.

Using the standard kinematics equation, s = v2/2a Where s = distance v = velocity a = acceleration due to gravity (32 ft/sec 2 in air, for conservatism),

Enclosure Response to Request for Additional Information Page 6 of 10 Solving for v; v = sqrt (2as) = sqrt (2

• 32 ft/sec 2

• 4ft) = 16 ft/sec This is the velocity achieved for a 4-foot drop (16 ft/sec). Since the velocity at the bottom of the fall in the spent fuel pool is 40% of that for a drop in the vessel, the number of rods damaged in the first impact would be 105

• 0.4 = 42. Since the second and third impacts are when the assembly falls from the vertical (to the horizontal position), this damage amount (19 fuel rods) remains unchanged. Therefore, the total fuel rods damaged would be 42 + 19 = 61. This amounts to 49.2% of the fuel rods damaged during a drop in the fuel pool.

6. NRC Question Appendix B of RG 1.183 allows an overall decontamination factor of 200 if the depth of water above damaged fuel is 23 feet or greater. If the depth of the water is less than 23 feet, the decontamination factor will have to be determined on a case-by-case method (Ref.

B-i) of RG 1.183. Provide either a detailed analysis that proves the statement in Section 4.3 that a shorter drop in the fuel pool would result in less radiological release to the containment or conservatively adjust the decontamination factor used for the FHA to account for water depth less than 23 feet (See Regulatory Position 2.4 of RG 1.183 and Regulatory Position 2 of Appendix B of RG 1.183).

Response

When a fuel handling accident occurs in the spent fuel pool, it is assumed that the dropped bundle will rest on the tops of the bail handles of the fuel in the storage racks. This puts the highest point of the (now horizontal) dropped assembly at 22.27 feet below the surface of the water when the fuel pool is at the minimum allowed water level.

Using the equations from the Reference B-1 of Regulatory guide 1.183, Burley's "Evaluation of Fission Product Release and Transport", the DF for 22.5 feet water coverage is determined to be 183.2 compared to the value of 200 determined for 23 feet. This is an 8.4% decrease in overall assumed DF. This slight decrease in DF is compared to the approximately 50% increase in damaged fuel rods assumed for the drop over the reactor vessel compared to the lesser fuel damage that would occur for a drop over the spent fuel pool. For conservatism, the accident with the most fuel damage produced was assumed (i.e., the drop over the reactor well, approximately 30 feet) while simultaneously limiting the decontamination factor assumed for a 23-foot drop to 200, and is therefore bounding for either possible accident.

ANALYSIS ASSUMPTIONS AND METHODOLOGY

7. NRC Question In Table 4-3 of the submittal, the Exclusion Area Boundary (EAB) and the Low Population Zone (LPZ) limiting dispersion factors are listed as "Normal RB exhaust stack" with limiting atmospheric dispersion factor (X/Q values) of 1.41 E-03 second per cubic meter (sec/m3 ) for the 414 m EAB and 1.35E-4 sec/m 3 for the 3218 m LPZ. The reference to stack release and the dispersion values are in conflict with a worst-case ground level release as well as the

Enclosure Response to Request for Additional Information Page 7 of 10 values listed in Attachment 1 (Calculation C-1302-822-E310-082) of the submittal Table 6.1.3 which shows ground level release EAB X/Q value of 1.10E-3 sec/m 3 and LPZ X/Q value of 5.60E-5 sec/m 3. Please clarify the above apparent discrepancies in your submittal.

Response

The EAB and LPZ values printed in Table 6.1.3 of Calculation C-1 302-822-E310-082 are incorrect and not used in the analysis. The values of 1.41 E-03 sec/m 3 for the 414 m EAB and 1.35E-4 sec/m 3 for the 3218 m LPZ are used correctly in the RADTRAD runs (see Page B-4 of the calculation) to calculate the overall dose. This editorial discrepancy did not impact any of the conclusions reached in the referenced submittal, and has been entered into the corrective action program.

METEOROLOGY ASSUMPTIONS

8. NRC Question In Table 4-3, on page 15 of Enclosure 1 to the submittal dated November 02, 2007, the licensee states:

"The most limiting 0-2 hr potential release point was determined to be the MAC facility entrance."

However, the table later lists the most limiting 0-2 hr (i.e., highest X/Q value) potential release point for the FHA was the Monitor and Control (MAC) Facility Personnel Airlock source location with a value of 6.75E-03 sec/m 3. The 0-2 hr X/Q value for the MAC Facility Entrance source location is 6.62E-03 sec/m 3. Please clarify this statement and confirm which 0-2 hr value was used in calculating the control room dose estimate for the Oyster Creek Nuclear Generating Station.

Response

The intent was to indicate the limiting X/Q in Table 4-3 as the MAC Facility Personnel Airlock. However, the first such indication incorrectly stated that it was for the MAC facility entrance instead of the MAC Facility Personnel Airlock (as stated on the line where the X/Q value is listed). The correct X/Q values were used in the RADTRAD runs in the calculation (see Page B-4 of the calculation). This editorial discrepancy did not impact any of the conclusions reached in the referenced submittal, and has been entered into the corrective action program.

9. NRC Question Two control room Heating, Ventilating, and Air-Conditioning System intakes (Intake A and Intake B) were used as the respective receptor points for each source location for the onsite X/Q analyses. ARCON96 was used to perform these calculations of estimated atmospheric dispersion values. The guidance for ARCON96, RG 1.194, notes that certain considerations should be evaluated in identifying the control room outside air intakes for which X/Q values should be considered.

Enclosure Response to Request for Additional Information Page 8 of 10 Please clarify if the two intakes to the control room used to model the onsite X/Q values are:

(1) in the same wind direction window or not and (2) share an imbalance of the intake flow rate to the control room. Based on these clarifications, please identify which ARCON96 X/Q equation was used for each source/receptor pair analyzed in this submittal and confirm that this was the appropriate methodology for the relative calculation. For example, if the Oyster Creek control room has dual intakes (e.g., Intake A and Intake B) with equal intake flow rates and share a wind direction window, then equation (5a) is deemed appropriate for this particular onsite X/Q analysis.

Response

(1) Each CR intake was analyzed individually for each source-receptor pair. The intake with the highest X/Q for each pair was used in the dose analysis. Therefore, there is no need to address wind direction window issues. Only one intake is used at a time. The accident analysis assumes it is the one with the highest X/Q.

(2) The ARCON96 point source calculation, which maximizes the X/Q, was used. There is no intake imbalance since we assume that the intake with the highest individual X/Q is used in the calculation.

10. NRC Question ARCON96 was used to compute the onsite X/Q values, which is supported by the guidance of RG 1.194. On page 14 of this regulatory guide (i.e., 1.194-14), it states that:

"If the distance to the receptor is less than about 10 meters, the ARCON96 code and the procedures in Regulatory Position 4 should not be used to assess X/Q values."

Considering this statement, please justify the use of the ARCON96 computer code to estimate the X/Q value for the Reactor Building Wall source location to the Control Room Intake A receptor point. The horizontal distance for this particular source/receptor pair is noted as 7.9 meters (25.8 feet), which is considerably less than 10 meters (as specified appropriate for use of the ARCON96 computer code in RG 1.194). Please indicate what impact this may have on the resulting X/Q values if used inappropriately.

Response

The subject diffuse area source-to-receptor distance of 7.9 meters that was assumed in Calculation C-1302-822-E310-081 is not~the closest distance from Intake A to the actual physical Reactor Building wall, but instead to the much nearer vertical planar surface projected perpendicular to the line of sight from Intake A to the center of the Reactor Building. This plane was positioned to intersect the Reactor Building at the point that minimizes the plane's horizontal distance from Intake A (as shown in Attachment G, page 5 of 6 of the Calculation). This distance is less than 10 meters, the approximate source-to-receptor distance less than which Regulatory Guide 1.194 indicates that ARCON96 may not be appropriate. However, if the X/Q for this scenario were to be recalculated by ARCON96 with a horizontal distance of 12.1 m (i.e., the distance from the closest point on the actual Reactor Building surface to Intake A, as depicted in the sketch below) in accordance with the diffuse area source method described in Section 3.2.4.5 of R.G 1.194, the resulting X/Q values would be smaller.

Enclosure Response to Request for Additional Information Page 9 of 10 Oyster Creek RAI #10 Assumed Diffuse Area Building Surface Release Point Scenario: Reactor Building Wall to Intake A Projected Plan

//

Plant North 14.7 True North

/-----1 Direction from Intake to Assumed Building Surface Release Point:

900 from Plant North 90'

- 14.70 = 75.3' from True North 137 ft ake the Note: Drawing not to scale.

1 Per RG 1.194, Section 3.2.4.5, the release point is defined as the closest point on the building surface to the intake.

Enclosure Response to Request for Additional Information Page 10 of 10

11. NRC Question Numerous assumptions were made relative to the building wake area and distance calculations for source/receptor pairs used in the assessment of control room X/Q values at the Oyster Creek Nuclear Generating Station. In order to justify these assumptions, a more graphically detailed schematic is needed than the plant layout provided in Attachment A to the amendment request.

Please provide a legible and suitably scaled schematic of the Oyster Creek Nuclear Generating Station site area showing true north. The drawing should indicate site boundaries and the location and orientation of principal plant structures within the site area representative of those used in the FHA most limiting X/Q analyses as presented in Table 4-3 of this submittal. Highlight postulated release and receptor points or locations and include relative straight-line or taut string distances, as deemed appropriate, on the drawing.

Response

Attachment E of the previously provided calculation C-1302-822-E310-081 contains all pertinent details with sufficient clarity with respect to all release points and receptors needed for ARCON96 input.

ADDITIONAL INFORMATION

12. NRC Question Provide the Oyster Creek proposed FSAR Markup, as suggested by Regulatory Position 1.5 of RG 1.183, or revised updated final safety analysis report pages outlining the Oyster Creek revised AST licensing basis for the FHA. Regulatory Position 1.6 of RG 1.183 outlines the FSAR update requirements including a reference to 10 CFR 50.71.

Response

NRC RG 1.183 recommends submitting the affected FSAR pages annotated with changes that reflect the revised analysis or submitting the actual calculation. To meet this recommendation, AmerGen provided the actual calculation documentation with the original license amendment request submittal (Reference). The affected FSAR pages are typically revised as part of the AmerGen process for implementing the NRC approved TS amendment, and as such, have not yet been developed.