Text

License Amendment Request to Technical Specification 3.9.3, Containment Penetrations, Personnel Air Locks Doors Open During Fuel Movement
	ML071210081
Person / Time
Site:	Farley
Issue date:	04/27/2007
From:	Stinson L; Southern Nuclear Operating Co
To:	; Document Control Desk, NRC/NRR/ADRO
References
	NL-07-0067
	Download: ML071210081 (164)
	v • d • e

L. M. Stinson (Mike)

Southern Nuclear Vice President Operating Company, Inc.

Fleet Operations Support 40 Inverness Center Parkway Post Office Box 1295 ri Birmingham, Alabama 35201 Tel 205.992.5181 Fax 205.992.0341 S

H SOUTHERN COMPANY April 27, 2007 Energy to Serve Your World' Docket Nos.: 50-348 NL-07-0067 50-364 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001 Joseph M. Farley Nuclear Plant, Units 1 and 2 License Amendment Request to Technical Specification 3.9.3, Containment Penetrations, Personnel Air Locks Doors Open Durinq Fuel Movement Pursuant to 10 CFR 50.90, Southern Nuclear Operating Company (SNC) hereby requests the following amendment:

Revise Joseph M. Farley Nuclear Plant (FNP) Units I and 2 Technical Specifications (TS) for Limiting Condition for Operation (LCO) 3.9.3, "Containment Penetrations," to allow the containment personnel air locks that provide direct access from the containment atmosphere to the auxiliary building to be open during refueling activities if appropriate administrative controls are established. The proposed changes are based on NRC-approved Technical Specifications Task Force (TSTF) Traveler TSTF-68, Revision 2. provides the basis for the proposed change. Enclosure 2 contains the TS and TS Bases markup pages. Enclosure 3 contains the TS and TS Bases clean typed pages. Enclosure 4 provides a list of Regulatory Commitments. provides an excerpt from the SNC Atmospheric Dispersion Factors Calculation. Enclosure 6 provides an excerpt from the SNC Design Bases Fuel Handling Accident Calculation.

SNC requests approval of the proposed license amendments by June 30, 2008.

This request date is to support the planned Unit 2 Refueling Outage 19 currently scheduled to start in October 2008. The proposed changes would be implemented within 90 days of issuance of the amendment.

Mr. L. M. Stinson states he is a Vice President of Southern Nuclear Operating Company, is authorized to execute this oath on behalf of Southern Nuclear Operating Company and to the best of his knowledge and belief, the facts set forth in this letter are true.

(Affirmation and signature are provided on the following page.)

U. S. Nuclear Regulatory Commission NL-07-0067 Page 2 If you have any questions, please advise.

Respectfully submitted, SOUTHERN NUCLEAR OPERATING COMPANY L. M. Stinson Sworn to and subscribed before me this.J_7 day of 14,0

, 2007.

Notary Public My commission expires: JM /,

6 Q'e5/O LMS/CHM

Enclosures:

1. Basis for Proposed Change

2. Technical Specifications and Bases Markups Pages

3. Technical Specifications and Bases Clean Typed Pages

4. List of Regulatory Commitments

5. Excerpt from SNC Atmospheric Dispersion Factors Calculation

6. Excerpt from SNC Design Bases Fuel Handling Accident Calculation cc:

Southern Nuclear Operating Company Mr. J. T. Gasser, Executive Vice President Mr. J. R. Johnson, Vice President - Farley Mr. D. H. Jones, Vice President - Engineering RTYPE: CFA04.054; LC# 14518 U. S. Nuclear Regulatory Commission Dr. W. D. Travers, Regional Administrator Ms. K. R. Cotton, NRR Project Manager - Farley Mr. E. L. Crowe, Senior Resident Inspector - Farley Alabama Department of Public Health Dr. D. E. Williamson, State Health Officer

Joseph M. Farley Nuclear Plant, Units 1 and 2 License Amendment Request to Technical Specification 3.9.3, Containment Penetrations, Personnel Air Locks Doors Open During Fuel Movement Basis for Proposed Change

Joseph M. Farley Nuclear Plant, Units 1 and 2 License Amendment Request to Technical Specification 3.9.3, Containment Penetrations, Personnel Air Locks Doors Open During Fuel Movement Basis for Proposed Change Table of Contents 1.0 Summary Description 2.0 Detailed Description 2.1 General Layout of FNP Auxiliary Building and Containment 2.2 Main Control Room 2.3 Containment 2.4 Proposed Change 3.0 Technical Evaluation 3.1 FHA Radiological Consequence Analysis 3.2 Emergency Worker Exposure 3.3 Technical Specification Changes 3.4 Technical Evaluation - Summary and Conclusions 4.0 Regulatory Evaluation 4.1 Significant Hazards Consideration 4.2 Applicable Regulatory Requirements / Criteria 4.3 Precedent 4.4 Conclusions 5.0 Environmental Consideration 6.0 References Table 1 Parameters and Assumptions Used in FHA Analysis Table 2 Atmospheric Dispersion Factors Used in FHA Analysis Table 3 Dose Results from Analysis of FHA Accident in Containment with Equipment Hatch and Personnel Air locks Open Figure 1 Auxiliary Building Elevation 155 ft. Layout Figure 2 Auxiliary Building Roof Ventilation Systems Air Intakes

CTMT Personnel Air locks Open - TSTF 68 NL-07-0067, Enclosure 1 Page 2 of 24 1.0 Summary Description This evaluation supports a request to revise Operating License (OL) NPF-2 and NPF-8 for Joseph M. Farley Nuclear Plant (FNP), Units 1 & 2 to allow reactor containment building personnel air locks, which provide direct access from the containment atmosphere to the auxiliary building, to be open during refueling activities if appropriate administrative controls are established. The air locks in question are the personnel air lock and the emergency personnel air lock. Currently, Technical Specifications (TS) 3.9.3 requires that one door in each air lock be closed during core alterations or movement of irradiated fuel inside containment.

The proposed change would revise TS Limiting Condition for Operation (LCO) 3.9.3, "Containment Penetrations," and is based on NRC-approved TSTF-68, Revision 2, "Containment Personnel Air Lock Doors Open During Fuel Movement" (reference 1). The NRC staff approved this TSTF on August 16, 1999. A revised Fuel Handling Accident (FHA) calculation for FNP shows acceptable dose consequences.

Revising TS 3.9.3 to permit personnel air lock doors to be open during core alterations or fuel movement has the following benefits:

" Reduced occupational exposure.

" Faster personnel evacuation from containment in the event of a FHA.

" Increased reliability of air lock doors due to reduced wear.

" Easier delivery of equipment to critical path activities inside containment.

" More flexibility in scheduling activities not on the critical path.

" Easier access to and from containment for equipment, personnel, laundry, and trash.

" Cleaner working environment.

In summary, the dose consequences of a FHA inside containment with the containment air locks open for the duration of the accident release are within the radiological dose guidelines of 10 CFR 100 and GDC 19. SNC requests that the NRC approve the proposed amendment based on the operational benefits, additional administrative controls, and acceptable dose consequences.

2.0 Detailed Description 2.1 General Layout of FNP Auxiliary Building and Containment FNP consists of two Westinghouse three loop pressurized water reactors located side by side with the auxiliary building located between the two containment buildings. FNP has a common control room located on the 155 ft. elevation of the auxiliary building. The ground elevation at FNP is 155 ft. Figure 1 provides a general layout of the 155 ft. elevation and relative location of the control room.

CTMT Personnel Air locks Open - TSTF 68 NL-07-0067, Enclosure 1 Page 3 of 24 2.2 Main Control Room The main control room is located on the 155 ft. elevation of the auxiliary building and provides a protected environment from which operators can control the unit(s) following an uncontrolled release of radioactivity, chemicals, or toxic gas. This environment is protected by the integrity of the Control Room Envelope (CRE) and the operation of the Control Room Emergency Filtration/Pressurization System (CREFS). The Unit 1 and 2 control room is a common room served by a shared CREFS.

The control room boundary is the combination of walls, floor, roof, ducting, valves or dampers, ESF HVAC equipment housings, doors, penetrations and equipment that physically form the CRE. The CRE is the area within the confines of the control room boundary that contains the spaces that control room operators inhabit to control the plant. This space is protected for normal operation, natural events, and accident conditions.

The CREFS components are arranged in redundant, safety related ventilation trains. The location of components within the CRE and ducting of the CRE ensure an adequate supply of filtered air to all areas requiring access. The CREFS provides airborne radiological protection for the control room operators, as demonstrated by the control room accident dose analyses for the most limiting design basis loss of coolant accident (LOCA), fission product release presented in the FSAR, Chapter 15.

Maintaining the integrity of the CRE limits the quantity of contaminants allowed into the CRE so that the radiological dose criteria of GDC 19 are met. The analysis of toxic gas releases demonstrates that the toxicity limits are not exceeded in the control room following a toxic chemical release. Figure 2 provides a general layout of the ventilation systems intake points located on the roof of the auxiliary building.

2.3 Containment The containment functions to contain fission product radioactivity that may be released from a reactor fuel assembly following a fuel handling accident, such that offsite radiation exposures are maintained within the requirements of 10 CFR 100, Reactor Site Criteria. Additionally, the containment structure provides radiation shielding from the fission products that may be present in the containment atmosphere following accident conditions. The requirements on containment penetration closure provided in Technical Specification (TS) 3.9.3 ensure that a release of fission product radioactivity within containment will be restricted from escaping to the environment.

The containment is a barrier to the release of fission products that breach the fuel cladding and reactor coolant pressure boundary during a core-damaging accident. The containment barrier, including air locks, is designed to limit the release of fission products such that offsite radiation exposure is well below the limits of 10 CFR 100.

CTMT Personnel Air locks Open - TSTF 68 NL-07-0067, Enclosure 1 Page 4 of 24 Each containment is equipped with two personnel air locks. Both air locks are located on the 155 ft. elevation and open into the auxiliary building (see figure 1). The air lock design utilizes two interlocked personnel access doors, which are part of the containment pressure boundary.

Each door is made leak tight by means of a seal against two surfaces, with provisions for testing seal integrity. Each exterior door has been provided with dual seals; thus, pressurizing between the two seals can test each individual exterior door seal.

The personnel air locks provide a means for personnel access during MODES 1, 2, 3, and 4. The doors are normally interlocked to prevent simultaneous opening when containment OPERABILITY is required.

During periods of shutdown when containment closure is not required, the door interlock mechanism may be disabled, allowing both doors of an air lock to remain open for extended periods when frequent containment entry is necessary. TS 3.9.3 currently requires at least one door of each personnel air lock to be closed during CORE ALTERATIONS or movement of irradiated fuel assemblies within containment.

2.4 Proposed Change FNP TS 3.9.3,"Refueling Operations - Containment Penetrations,"

currently requires one door in each air lock to be closed during operations involving core alterations or fuel movement inside containment. The proposed change would allow both doors on each air lock to be open during operations involving core alterations or fuel movement inside containment if appropriate administrative controls are established and maintained. Specifically, the proposed change revises TS 3.9.3(b) to read "One door in each air lock is capable of being closed."

During refueling, a large number of personnel can be in containment at any given time. Should containment evacuation be required, such as during a FHA, it would take a number of cycles of the air lock(s) to evacuate personnel from the containment. This could cause the workers waiting to exit to be unnecessarily exposed to any released activity or other hazardous material. Under the proposed change, allowing both air lock doors to be open, but with one door capable of being closed, containment could be evacuated more rapidly and efficiently. After evacuation is complete, one of the personnel air lock doors would then be closed. This would reduce the exposure to workers in the event of an accident while maintaining exposure to the public within guidelines.

Additionally, it is expected that maintenance of the personnel air locks will also be reduced by limiting the number of open/close operations during a refueling as well as expediting the movement of personnel and equipment in and out of containment.

The proposed change is based on TSTF-68 which was approved based on (1) acceptable radiological consequences from a FHA, and (2) the implementation of administrative procedures to ensure that in the event of

CTMT Personnel Air locks Open - TSTF 68 NL-07-0067, Enclosure 1 Page 5 of 24 a refueling accident that the open air locks can and will be closed following containment evacuation. The SNC dose calculations document for the FHA assumes the air lock doors are not closed. Plant procedures will direct that the equipment hatch and at least one door in each air lock be promptly closed following containment evacuation.

SNC is processing a supporting TS Bases change requiring that in the event of a FHA an open personnel air lock can and will be closed promptly following containment evacuation. No other controls are considered necessary to implement this amendment. This closure can be completed within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months of the fuel handling accident.

In summary, these proposed amendments would allow the personnel air locks to remain open under appropriate administrative controls during CORE ALTERATIONS and movement of irradiated fuel in containment.

Implementation of the proposed amendments also requires a change to TS Bases 3.9.3 in accordance with TSTF 68, Rev. 2.

3.0 Technical Evaluation To support this proposed TS amendment, SNC performed a reanalysis of the FHA described in the FNP FSAR Chapter 15. The reanalysis was required since the proposed TS amendment would result in the containment personnel air locks being open during a postulated FHA. The open air locks would change the containment release characteristics and could alter the offsite and control room projected dose.

TS amendment number 165 / 157 was issued September 30, 2004 (reference

2) that allowed the containment equipment hatch to remain open during fuel movement. Therefore, the new FHA analysis has been performed with the containment equipment hatch and personnel air locks open during the postulated FHA inside containment. Parameters and assumptions used in the FHA in containment are provided in Table 1.

3.1 FHA Radiological Consequence Analysis This accident analysis postulates that the spent fuel assembly with the highest inventory of fission products of the 157 assemblies in the core is dropped during refueling. All of the fuel rods in the assembly are conservatively assumed to rupture, releasing the radionuclides within the fuel rods to the reactor cavity water. Volatile constituents of the core fission product inventory migrate from the fuel pellets to the gap between the pellets and the fuel rod clad. The fission product inventory in the fuel rod gap of the damaged fuel rods is assumed to be instantaneously released because of the accident. Fission products released from the damaged fuel are decontaminated by passage through the overlaying water in the reactor cavity depending on their physical and chemical form.

SNC assumes no decontamination for noble gases, an overall effective decontamination factor of 200 for radioiodines, and retention of all particulate fission products. SNC also assumed that essentially 100

CTMT Personnel Air locks Open - TSTF 68 NL-07-0067, Enclosure 1 Page 6 of 24 percent of the fission products released from the reactor cavity are released to the environment in 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months without any credit for filtration.

10 CFR 50, Appendix A, General Design Criterion (GDC) 19, "Control Room," requires that licensees maintain the control room in a safe condition under accident conditions. Under these conditions, the licensee must provide adequate radiation protection to permit access and occupancy of the control room. 10 CFR 100.11, "Determination of exclusion area, low population zone [LPZ] and population center distance," on the other hand, establishes the dose guidelines (but used as limits) for the exclusion area and for the public.

In order to show that the radiation doses to people onsite and offsite will meet the above regulatory requirements, SNC performed evaluations of accident radiation doses for the FHA inside containment. Regulatory guidance for these evaluations is provided in the form of regulatory guides and standard review plans. The regulatory requirements are contained in 10 CFR 50, Appendix A, GDC 19 and 10 CFR 100.11, as supplemented by Regulatory Position 4.4 and 4.5 of RG 1.195. The guidance of RG 1.194, "Atmospheric Relative Concentrations for Control Room Radiological Habitability Assessments at Nuclear Power Plants" was also utilized.

The assumptions that were used in this analysis are presented in Table 1.

The Atmospheric Dispersion Factors (X/Q) used in the FHA analysis are provided in Table 2. The exclusion area boundary (EAB), low population zone (LPZ), and control room doses results for the FHA analysis are provided in Table 3. An excerpt from the SNC FHA dose calculation is provided in enclosure 6 SNC used analysis methods and assumptions consistent with the conservative regulatory requirements and guidance identified above.

SNC's estimates of the control room doses continue to comply with these criteria (5 rem whole body or equivalent). SNC's estimates of the EAB and LPZ doses continue to comply with these criteria (6.3 rem whole body and 75 rem thyroid).

3.1.1 Atmospheric Dispersion Estimates 3.1.1.1 Calculation Background To support the review and issuance of TS amendment 165 / 157 (reference 2), the SNC calculation for Control Room and Technical Support Center (TSC) air intake X/Q was submitted to the NRC.

The safety evaluation report in reference 2 provides a description of the review performed and the acceptability of the resulting X/Qs for use in FHA radiological consequence analysis. No changes have been made to this calculation since it was used in support of reference 2. Excerpts from the SNC Atmospheric Dispersion Factors Calculation are provided in enclosure 5.

CTMT Personnel Air locks Open - TSTF 68 NL-07-0067, Enclosure 1 Page 7 of 24 3.1.1.2 Control Room Atmospheric Dispersion Factors SNC calculated X/Q values using the ARCON96 atmospheric dispersion computer code (NUREGICR-6331, Revision 1, "Atmospheric Relative Concentrations in Building Wakes"). SNC determined the control room X/Q values for releases from each of the two (Units 1 and 2) containment equipment hatches as well as from the four personnel air locks (via plant vent stack or containment wall) to the control room air intakes, and the Technical Support Center (TSC) air intake.

A set of X/Q values was produced using both the 4 year and 4.5 year meteorological data. SNC reviewed both sets of X/Q values to determine the most conservative value for each release / intake point.

Enclosed are excerpts from the SNC calculations for determining the X/Q values for the control room and TSC which include the ARCON96 output files. This is provided in enclosure 5.

The ARCON96 computer runs are consistent with FNP site configuration drawings. Specific areas of note are as follows:

Figure 1 provides a general layout of the 155 ft. elevation for the containment equipment hatch, personnel air locks and the control room.

Figure 2 provides a general layout of auxiliary building rooms showing the relation between the ventilation system intakes and the containment hatches and air locks.

SNC modeled the releases as a ground-level source. SNC took into consideration the difference in elevation between the release height and the intake height.

SNC only considered the portion of the reactor building that is higher than the containment hatch doors or control room intake in the building wake analysis for equipment hatch and containment building release respectively. This approach is conservative because smaller building cross-sectional areas produce higher X/Q values. No wake mixing was considered in the vent stack release analysis.

The release from the containment personnel air locks into the auxiliary building is picked up by the auxiliary building rad-side ventilation system and vented out of the auxiliary building through the plant vent stack. As a conservative measure, the release point is considered to be the containment wall since it is closer to the control room intakes and results in higher X/Q.

CTMT Personnel Air locks Open - TSTF 68 NL-07-0067, Enclosure 1 Page 8 of 24 The resulting air intake X/Q values were reviewed and the following values were used in the FHA analysis: (refer to Table 2 of this enclosure and section 8.1 of SNC atmospheric dispersion factors calculation provided in enclosure 5.)

(1) Routine outside air makeup through the control room normal air intake prior to control room isolation; The normal control room air intake is closer to the release pathways as compared to the control room emergency air intakes, but bounded by the Unit 2 to TSC intake location, thus the routine outside air makeup through the normal air intake can be represented by the U2 to TSC air intake X/Q values for the FHA. This assumption results in conservatively short distances that bound the actual plume travel distances between the release point and the control room normal air intake. The X/Q values for the TSC intake from all release points were reviewed. The largest and most conservative X/Q value was chosen. The value was for "U2 Reactor" to the TSC, 5.06 x 10-3.

(2) Unfiltered inleakage when the control room is isolated but not pressurized; When the control room is isolated, the control room normal intake and the TSC intake are closed. The X/Q values for the Unit 1 and Unit 2 control room emergency intakes from all release points were reviewed. The largest and most conservative X/Q value was chosen. The value was for "U2 Reactor" to the Ul control room, 1.66 x 10-3.

(3) Emergency filtered pressurization air makeup through the control room emergency air intake when the control room is isolated and pressurized; When the control room is isolated and pressurized, the only intake point is the control room emergency intake. The X/Q values for the Unit 1 and Unit 2 control room emergency intakes from all release points were reviewed.

The largest and most conservative X/Q value was chosen.

The value was for "U2 Reactor" to the Ul control room, 1.66 x 103.

(4) For an assumed release from the equipment hatch, to the control room emergency intake; When the control room is isolated and pressurized, the only intake point is the control room emergency intake. The X/Q values for the Unit 1 and Unit 2 control room emergency intakes from Unit 1 and Unit 2 equipment hatch

CTMT Personnel Air locks Open - TSTF 68 NL-07-0067, Enclosure 1 Page 9 of 24 release points were reviewed. The largest and most conservative X/Q value was chosen. The value was for "U1 Hatch Door" to the Unit 1 control room, 8.79 x 10-4.

3.1.1.3 Offsite Atmospheric Dispersion Factors SNC evaluated offsite doses using offsite (EAB and LPZ) X/Q values presented in the FNP FSAR Tables 2.3-12 and 15B-2.

These values are presented in Table 2. They represent sector independent (overall site) five percentile X/Q values derived from hourly records of onsite data from the period April 1971 through March 1972. Details on the calculation can be found in the FNP FSAR Section 2.3.4. This is consistent with values used in TS amendment 165/157 (reference 2).

3.1.2 Control Room Unfiltered Inleakage The Control Room Emergency Filtration/Pressurization System (CREFS) components are arranged in redundant, safety related ventilation trains. The location of components within the Control Room Envelope (CRE) and ducting of the CRE ensure an adequate supply of filtered air to all areas requiring access.

For the FHA analysis, conservative values for intake of unfiltered air in to the control room are assumed. A list of parameters and assumptions are provided in Table 1.

The following is a brief description of the unfiltered inleakage assumed in the FHA:

At the start of the FHA, normal unfiltered intake of 2,340 ft3/min is assumed.

The control room radiation monitors will activate and isolate the control room ventilation system within 30 sec.

Following isolation, un-pressurized, unfiltered infiltration of 600 ft3/min is assumed.

Within 10 minutes, the control room operators will manually start CREFS.

Following the start of CREFS, Pressurized unfiltered infiltration of 450 ft3/min is assumed.

These assumed inleakage values bound the tested values identified as part of the GL 2003-01 response and documented in the NRC closure letter (reference 4). This provides a conservative basis for unfiltered inleakage in the FHA analysis.

CTMT Personnel Air locks Open - TSTF 68 NL-07-0067, Enclosure 1 Page 10 of 24 The NRC staff issued GL 2003-01, "Control Room Habitability." SNC responded to this GL by letter dated August 25, 2004 (reference 5). In its response, SNC reported that inleakage testing using the ASTM E741 tracer gas methodology determined a control room unfiltered inleakage rate of 25 cfm during the pressurization mode, 33 cfm during the isolation mode and 87 cfm during the normal alignment. All of these modes are used for the FHA analysis. The NRC issued a letter on September 20, 2006 (reference 4) confirming the receipt of the information provided in the August 25, 2004 letter from SNC and that SNC's response to GL 2003-01 was considered complete.

3.1.3 Control Room Dose and Offsite Doses The exclusion area boundary (EAB), Low Population Zone (LPZ), and control room dose calculation consequences for the FHA are provided in Table 3.

The analysis results demonstrate that the offsite and control room doses due to a FHA in the containment building with the containment equipment hatch and the containment personnel air locks open are within the acceptance criteria given in RG 1.195, GDC 19 and 10 CFR 100.

3.2 Emergency Worker Exposure The proposed revision to TS 3.9.3 would allow for the containment equipment hatch and the personnel air locks to be open during core alterations and/or during movement of irradiated fuel assemblies within containment. The SNC evaluation of the overall impact that the proposed modification would have on its ability to control radiation exposure to emergency workers in the event of a FHA inside containment was provided in support of TS amendment 165/157 (reference 2) and remains applicable to the currently proposed TS change. Additionally, procedures are in place to ensure that exposures are controlled consistent with Environmental Protection Agency Emergency Worker and Lifesaving Activity Protective Action Guides as referenced in 10 CFR 50.47(b)(1 1).

Maintenance personnel have procedural guidance for normal closure of the containment equipment hatch and personnel airlocks. These activities are performed routinely during refueling outages. Also, training and actual performance or simulation of the hatch closure is required of Mechanical Maintenance Journeymen.

A pre-job brief will be performed with the designated trained hatch / air locks closure crew prior to starting core alterations. The pre-job brief would discuss: the requirements of the Radiation Worker Permit or Health Physics Plan that would be used for the closure of the containment equipment hatch and personnel air locks, expected radiological conditions, protective clothing requirements and actions to take if the plant emergency alarm is used.

CTMT Personnel Air locks Open - TSTF 68 NL-07-0067, Enclosure 1 Page 11 of 24 The plant procedures require personnel to wear dosimetry when entering a Radiologically Controlled Area in support of the containment equipment hatch closure. Protective clothing would be required to be worn in contaminated areas and authorization would be given to allow personnel to wear protective clothing over their personal clothing. Eye protection would also be required.

Through General Employee Training, maintenance personnel train in the use of respiratory devices or other methods to limit the intake of radioactive material. While respirators are available, they would not be required for entries to provide emergency closure of the equipment hatch and personnel air locks. Although not anticipated, if a person was potentially exposed to airborne radioactive iodine, the issuance of potassium iodide (KI) as a thyroid blocking agent would be considered.

Existing plant procedures are in place to provide guidance for use of respiratory equipment or utilization of KI should the need arise. Because of the existing training, no additional training would be required to support closure activities.

If any monitor alarms as a result of a FHA, Health Physics (HP) would perform the following:

(1) safely evacuate personnel; (2) contact the control room and HP Supervisor for additional actions; (3) secure access to the area by non-essential personnel; (4) conduct additional sampling as directed; and (5) provide dedicated HP support to the designated trained closure crew to include escort to the work area, setting dose rates for the workers and escort out of containment.

The radiological conditions at the containment equipment hatch and personnel air locks would be assessed directly by HP personnel providing support to the designated trained closure crew using an instrument that can detect high levels of beta radiation and through the following air sampling activities:

(1) Continuous Air Monitors would be in service inside containment, in the area of the containment equipment hatch outside of containment and in the area of the personnel air locks inside of the auxiliary building any time the hatch and personnel air locks are open and fuel movement is in progress. Short interruptions of monitoring are expected to support response check of the instruments, filter change out or replacement of a malfunctioning instrument with one on standby,

CTMT Personnel Air locks Open - TSTF 68 NL-07-0067, Enclosure 1 Page 12 of 24 (2) Low Volume Air samplers would be running continuously inside containment, in the area of the containment equipment hatch outside of containment and in the area of the personnel air locks inside of the auxiliary building any time the equipment hatch and personnel air locks are open and fuel movement is in progress.

Air sampling would be for particulate and iodine activity.

(3) If the situation warrants, noble gas samples would be taken.

In the event that contamination of personnel occurs, whole body counts would be utilized to assess the radiological exposure. If additional assessments are required due to known or suspected intakes of radioactive material, follow-up bioassay sampling and analysis may be conducted.

SNC performed an evaluation of potential doses to the worker's thyroid, whole body, and skin. The equipment hatch and the personnel air locks will be closed within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months of the event. The closure crew will require less than 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months inside containment. The dose to a crew member inside the containment for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months would be approximately 48.2 rem thyroid, 1.1 rem skin, and < 0.1 rem whole body. Crew members transiting to the containment equipment hatch from inside the containment would include their transit time in the 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months total stay time. Crew members transiting to the hatch from outside the containment would be exposed to activity exhausted from the containment (which would reduce the exposure inside containment). Assuming complete exhaust during the transit, the resultant doses would be about 1.3 rem thyroid, and < 0.1 rem whole body and skin. Radiation doses of these magnitudes are within regulatory exposure limits and do not represent an impact to worker health.

No other significant contributions to worker dose are expected since the containment purge filter is outside the containment and the pre-access filter, typically not used during refueling, is on the opposite side of the containment and is partially shielded by the steam generator and compartment walls.

Based on the above, SNC will be able to control radiation doses to its emergency workers within the regulatory limits.

3.3 Technical Specification Changes SNC's current LCO 3.9.3.b requires one door in each air lock to remain closed during core alterations and movement of irradiated fuel assemblies within containment by stating "One door in each air lock closed." To allow for the air locks to remain open during these conditions, SNC proposes modifying the LCO to state, "One door in each air lock is capable of being closed;"

CTMT Personnel Air locks Open - TSTF 68 NL-07-0067, Enclosure 1 Page 13 of 24 3.4 Technical Evaluation - Summary and Conclusions As described above, SNC reviewed the assumptions, inputs, and methods used to assess the impacts of the proposed change to the FNP TSs.

Based on its review, SNC finds that the analysis methods and assumptions are consistent with the conservative regulatory requirements and guidance identified in Section 4.0 below. SNC finds that with reasonable assurance, the estimates of the control room doses would continue to comply with these criteria (5 rem whole body or equivalent).

SNC also finds, with reasonable assurance, the estimates of the EAB and LPZ doses would continue to comply with these criteria (6.3 rem whole body and 75 rem thyroid). Therefore, the proposed license amendment is acceptable with regard to the radiological consequences of the postulated FHA. Furthermore, SNC also finds that there is reasonable assurance that it will be able to control radiation doses to its emergency workers within the regulatory limits.

4.0 Regulatory Evaluation 4.1 Significant Hazards Consideration Southern Nuclear Operating Company (SNC) has evaluated whether or not a significant hazards consideration is involved with the proposed changes by focusing on the three standards set forth in 10 CFR 50.92(c) as discussed below:

1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The proposed change would allow the personnel air lock doors, and emergency air lock doors to remain open during fuel movement and core alterations. These doors are normally closed during this time period in order to prevent the release of radioactive material in the event of a fuel handling accident (FHA) inside containment. These doors are not initiators of any accident. The probability of a FHA is unaffected by the operational status of these doors.

The new FHA analysis with open containment personnel air locks demonstrates that maximum offsite dose is within the acceptance limits specified in RG 1.195. The FHA analysis results in maximum off site doses of 68.5 rem to the thyroid and 0.2 rem to the whole body.

The calculated control room dose is also within the acceptance criteria specified in GDC 19. The analysis results in thyroid and whole body doses to the control room operator of 39.6 rem and < 0.1 rem, respectively.

CTMT Personnel Air locks Open - TSTF 68 NL-07-0067, Enclosure 1 Page 14 of 24 Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The proposed change does not involve the addition or modification of any plant equipment. Also, the proposed change will not alter the design, configuration, or method of operation of the plant beyond the standard functional capabilities of the equipment. The proposed change involves a TS change that will allow the air lock doors to be open during core alterations and fuel movement inside containment.

Open doors and penetrations do not create the possibility of a new accident. Administrative controls will be implemented to ensure the capability to close the containment in the event of a FHA.

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

3. Does the proposed amendment involve a significant reduction in a margin of safety?

Response: No.

The proposed change has the potential to increase the post-FHA dose at the Site Boundary, Low Population Zone and in the control room.

However, a revised FHA analysis demonstrates that the dose consequences at both locations remains within regulatory acceptance limits and the margin of safety as defined by 10 CFR 100 and GDC 19 has not been significantly reduced. To ensure a bounding calculation, the revised FHA was performed with conservative assumptions. For example, it assumes the unfiltered release to the outside atmosphere of all airborne activity reaching the containment. Additional margin will be established through administrative procedures to require that the equipment hatch and at least one door in each air lock be closed following an evacuation of containment.

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

Based on the above, SNC concludes that the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

CTMT Personnel Air locks Open - TSTF 68 NL-07-0067, Enclosure 1 Page 15 of 24 4.2 Applicable Regulatory Requirements / Criteria The following lists the regulatory requirements and plant-specific design bases related to the proposed change to TS 3.9.3.

Regulatory Requirements a The regulatory basis for TS 3.9.3, "Containment Penetrations," is to ensure that the primary containment is capable of containing fission product radioactivity that may be released from the reactor core following a FHA inside containment. This ensures that offsite radiation exposures are maintained within the requirements of 10 CFR 100.

0 10 CFR Part 50, Appendix A, General Design Criterion (GDC) 16, "Design," requires that reactor containment and associated systems shall be provided to establish an essentially leak-tight barrier against the uncontrolled release of radioactivity to the environment and to assure that the containment design conditions important to safety are not exceeded for as long as the postulated accident conditions require.

N GDC 19, "Control Room," requires that adequate radiation protection shall be provided to permit access and occupancy under accident conditions without personnel receiving radiation exposure in excess of 5 REM whole body, or its equivalent to any part of the body for the duration of the accident.

a GDC 61, "Fuel Storage and Handling and Radioactivity Control,"

requires that the fuel storage and handling, radioactive waste, and other systems which may contain radioactivity shall be designed to assure adequate safety under normal and postulated accident conditions.

0 The parameters of concern and the acceptance criteria applied are based on the requirements of 10 CFR 100 and GDC 19 with respect to the calculated radiological consequences of a FHA and GDC 61 with respect to appropriate containment, confinement, and filtering systems.

Regulatory Guidance UFSAR Section 15.4.5 - The FNP Units 1 & 2 design basis FHA is defined as the dropping of a spent fuel assembly onto the spent fuel pool fuel storage area or inside containment. Both analyses assume the rupture of the cladding of all the fuel rods in the assembly. Section 15.4.5 of the UFSAR discusses the consequences of a postulated FHA inside containment.

Regulatory Guide 1.194, June 2003, "Atmospheric Relative Concentrations for Control Room Radiological Habitability

CTMT Personnel Air locks Open - TSTF 68 NL-07-0067, Enclosure 1 Page 16 of 24 Assessments at Nuclear Power Plants," The design guidance and assumptions of Regulatory Guide 1.194 are used with plant meteorological data for years 2000 through 2003 to calculate control room and technical support center X/Q for releases from the containment, containment equipment hatch, and plant vent. The plant meteorological data have been supplemented by a data set constructed to generate a 4.5-year data set with an overall frequency from the SSE, S, and SSW directions similar to that in the 1971 through 1975 data set presented in UFSAR section 2.3.

Regulatory Guide 1.195, May 2003, "Methods and Assumptions for Evaluating Radiological Consequences of Design Basis Accidents at Light-Water Nuclear Power Reactors," describes the methodology used by FNP Units 1 & 2 in the evaluation of a fuel handling accident in the containment except: Dose conversion factors described in subsections 4.1.2 and 4.1.4 are not used. Instead, dose conversion factors from ICRP 30, "Limits for Intakes of Radionuclides by Workers," are used as described in UFSAR Table 151B-1.

Regulatory Guide 1.196, May 2003, "Control Room Habitability at Light-Water Nuclear Power Reactors," provided detail guidance regarding control room habitability and the documentation of the control room habitability licensing bases as well as a program for testing the systems. FNP complies with this RG with the following exceptions:

1. FNP implemented a Technical Specification (TS) change that is an acceptable alternative to the TS change identified in RG 1.196 which includes control room envelope (CRE) integrity testing and periodic assessments. These TS changes implement a Control Room Integrity Program.

2. FNP continues to use RG 1.52, Revision 0, for the design. FNP has updated to Revision 3, dated June 2001, for testing only.

3. FNP uses RG 1.140 for information only for non-safety-related air filtration systems.

4. FNP uses ASHRAE Guideline 1-1996 for reference only for maintenance programs for systems that handle hazardous chemicals and smoke challenges.

5. FNP uses RG 1.197, dated May 2003, for testing of the CRE.

FNP has completed initial testing and will perform future testing of the Control Room Habitation System (CRHS) in accordance with RG 1.197, with the exceptions as defined in NEI 99-03, Rev. 1, Appendix EE, "ASTM E741 Exceptions."

6. Surveys of hazardous chemicals will continue to be performed on a frequency of every 6 years. Offsite will be conducted in the year

CTMT Personnel Air locks Open - TSTF 68 NL-07-0067, Enclosure 1 Page 17 of 24 the tracer gas testing is scheduled, and the onsite will be offset by 3 years.

FNP Units 1 & 2 use NUREG-0800, U.S. NRC Standard Review Plan, Section 15.4.5, to evaluate system design features and plant procedures provided for the mitigation of the radiological consequences of postulated fuel handling accidents.

Regulatory criteria and guidance are contained in Regulatory Guide 1.195 and Section 15.7.4 of NUREG-0800. The calculated doses are within the Standard Review Plan criteria of 6.3 REM to the whole body and 75 REM to the thyroid.

Section 15.4.5 of the FNP Units 1 & 2 UFSAR describes system design features and plant procedures for mitigating the radiological consequences of postulated FHAs. Offsite calculation assumes no credit for iodine removal by the atmosphere filtration system filters. All radioactivity released to the containment is assumed to be released to the environment at ground level over a two-hour period.

4.3 Precedent TSTF-68 was based primarily on the license amendment granted to Arkansas Nuclear One, Unit 2 on September 28, 1995 (reference 6). This change incorporates into the Improved TS NUREGs the option to allow both containment personnel air lock doors to remain open during fuel movement. This option has been granted to many plants starting with the August 31, 1994 approval of an amendment to the Calvert Cliffs Nuclear Power Plant Technical Specifications.

The proposed change is based on TSTF-68, Rev. 2 which was approved by the NRC based on (1) acceptable radiological consequences from a FHA, and (2) the implementation of administrative procedures to ensure that in the event of a refueling accident that the open air locks can and will be closed following containment evacuation. These conditions have been met by SNC.

4.4 Conclusions The technical analysis performed by SNC demonstrates that the dose consequences at the exclusion area and low population zone boundaries are within the limits of 10 CFR 100. Therefore, the proposed License amendment is in compliance with the General Design Criteria (16, 19, and 61), Regulatory Guide 1.195, and Section 15.7.4 of the SRP (NUREG-0800).

CTMT Personnel Air locks Open - TSTF 68 NL-07-0067, Enclosure 1 Page 18 of 24 5.0 Environmental Consideration SNC has determined that the proposed amendment would change requirements with respect to the installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. SNC has evaluated the proposed change and has determined that the change does not involve, (i) a significant hazards consideration, (ii) a significant change in the types of or significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. As discussed above, the proposed changes do not involve a significant hazards consideration and the analysis demonstrates that the consequences from a FHA are within the 10 CFR 100 limits. The implementation of administrative controls precludes a significant increase in occupational radiation exposure.

Accordingly, the proposed change meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51, specifically 10 CFR 51.22(c)(9). Therefore, pursuant 10 CFR 51.22(b), an environmental assessment of the proposed change is not required.

6.0 References

1. Technical Specification Task Force, TSTF-68, Revision 2, "Containment Personnel Air lock Doors Open During Fuel Movement" August 16, 1999. (ML040440044)

2. NRC Letter Dated September 30, 2004, Joseph M. Farley Nuclear Plant, Units 1 and 2 RE: Issuance of Amendments [165/157] (TAC Nos.

MC0625 and MC0626) (ML042820368)

3. NRC Letter Dated September 30, 2004, Joseph M. Farley Nuclear Plant, Units 1 and 2 RE: Issuance of Amendments [166/158] (TAC Nos.

MC4186 and MC4187) (ML042780424)

4. NRC Letter Dated September 20, 2006, NRC Receipt of Joseph M.

Farley Nuclear Plant, Units 1 and 2, Response to Generic Letter 2003-01 "Control Room Habitability" (TAC Nos. MB9802 and MB9803)

(ML062510470)

5. SNC Letter Dated August 25, 2004, "Joseph M. Farley Nuclear Plant, Response to Generic Letter 2003 Control Room Habitability" (ML042400122)

6. NRC Letter Dated September 28, 1995, Arkansas Nuclear One, Unit 2, Issuance of Amendment No. 166 to Facility Operating License No. NPF-6 (TAC NO. M92150)

CTMT Personnel Air locks Open - TSTF 68 NL-07-0067, Enclosure 1 Pagje 19 of 24 Table 1 (sheet 1 of 2)

Parameters and Assumptions Used in FHA Analysis Parameter or Assumption Core thermal power Time between plant shutdown and accident Minimum water depth between tops of Damaged fuel rods and water surface Damage to fuel assembly Fuel assembly activity Activity release from assembly Radial peaking factor Decontamination factor in water Elemental iodine (99.75%)

Organic iodine (0.25%)

Noble gases Containment Exhaust flow rate Containment Exhaust isolation time Iodine filtration system (Mwt)

(hr)

(ft)

Value Used 2,831 100 23 All rods ruptured Highest powered fuel assembly in core region discharged Gap activity in ruptured rods per RG 1.195, Table 2 1.7 400 1

1 53,500 NA Containment purge system (not credited)

(cfm)

Filter efficiency (all species)

Dose Conversion Factors NA ICRP 30

CTMT Personnel Air locks Open - TSTF 68 NL-07-0067, Enclosure 1 Page 20 of 24 Table 1 (sheet 2 of 2)

Parameters and Assumptions Used in FHA Analysis Parameter or Assumption Value Used CR(1) Normal HVAC unfiltered intake (ft3/min) 2,340(2)

CR Un-pressurized unfiltered infiltration (ft 3/min) 600 CR Filtered pressurization rate (ft 3/min) 450 CR Pressurized unfiltered infiltration (ft 3/m in) 450 CR Filtered recirculation rate (ft3/min) 2,700 CR Unfiltered ingress/egress rate (ft3/min) 10 CR Filter efficiencies (all forms of iodine)

(%)

Pressurization air 98.5(3)

Recirculation air 94.5(3)

CR Isolation time (s) 30 CR Pressurization Sys Manual Start Time (min) 10 CR Volume (ft3) 114,000 CR Operator breathing rate (m3/s) 3.47 x 10.4 Percent of time operator is in control room

(%)

following loss-of-coolant accident 0-1 day 100 1-4 days 60 4-30 days 40 Notes:

(1)

Control Room is represented by "CR".

(2)

The amounts of unfiltered inleakage into the control room used in this new FHA analysis are consistent with the values submitted to the NRC in support of the September 30, 2004 amendment (reference 2). The only control room inleakage value that was adjusted was the control room normal HVAC unfiltered intake. It was changed from 3,000 ft3/min to 2,340 ft3/min. This was to reflect the value found in TS 5.5.18, "Control Room Integrity Program (CRIP)." TS 5.5.18 was introduced into the TS with amendment 166/158 issued on September 30, 2004 (reference 3).

(3) Filter efficiencies have been reduced by 0.5 percent for all forms of iodine to account for bypass leakage.

CTMT Personnel Air locks Open - TSTF 68 NL-07-0067, Enclosure 1 Page 21 of 24 Table 2 Atmospheric Dispersion Factors Used in FHA Analysis Equipment Personnel Hatch Air locks Boundary Time Interval X/Q Value (sec/m3)

X/Q Value (sec/m 3) 0 - 30 sec 8.7 9 x 10.4 5.06 x 10-3 Control Room 30 sec-2 hrs 8.79 x 10-4 1.66 x 10-3 Site Boundary 0 - 2 hrs 7.6 x 10-4 7.6 x 10-4 Low Population 0 - 2 hrs 2.8 x 10-4 2.8 x 10-4 Zone

CTMT Personnel Air locks Open - TSTF 68 NL-07-0067. Enclosure 1 Paae 22 of 24 Table 3 Dose Results from Analysis of FHA Accident in Containment with Equipment Hatch and Personnel Air locks Open Dose (rem)

Thyroid Analysisl RG Whole Body Analysisi RG Skin AnalysisI RG 2 hr Exclusion Area Boundary (EAB) 68.5 75 0.2 6.3 0.5 50 10 CFR 100 30 day Low Population Zone Boundary (LPZ) 25.3 75 0.1 6.3 0.2 50 10 CFR 100 30 day Control Room 39.6 50

<0.1 5

1.0 50 GDC 19

CTMT Personnel Air locks Open - TSTF 68 NL-07-0067, Enclosure 1 Page 23 of 24 Figure 1 Auxiliary Building Elevation 155 ft. Layout North

CTMT Personnel Air locks Open - TSTF 68 NL-07-0067, Enclosure 1 Page 24 of 24 Figure 2 Auxiliary Building Roof Ventilation Systems Air Intakes North UI Elec Chase Intake U1 Mechanical Equipmenl Room Access Door U1 Non-radArea. Intake N

4 U tU!

Control Room.Intake Control U2 Elec Chase Intake t

0U2 Mechanical Equipment Room Access Door SO

.U2 Non-rad Area intake.

U2 Control Room.Intake0 Unit 2 Normal Intake NraIntake ntakeEm.

Personnel nt~keAir Lock 155' 12 Rad Side.Area Intake Equipment Hatch 155' U2 Plant Vent Stack\\\\

I Personnel Air

\\2Pln Ven Stac

~I i

Em. Personnel Air Lock 155' U1 Rad.Side Area I 1

0 Equipment Hatch 155' P~ersonnel 1u Plant Vent Stack AirLock 155' I

LVULIK I U IXI

-i U-I Iuxiiary uunaing' Roof If

Joseph M. Farley Nuclear Plant, Units 1 and 2 License Amendment Request to Technical Specification 3.9.3, Containment Penetrations, Personnel Air Locks Doors Open During Fuel Movement Technical Specifications and Bases Markups Pages

NL-07-0067, Enclosure 2, Page 1 of 9 Containment Penetrations 3.9.3 3.9 REFUELING OPERATIONS 3.9.3 Containment Penetrations LCO 3.9.3 The containment penetrations shall be in the following status:

a.

The equipment hatch is capable of being closed and held in place by four bolts; c

e

b.

One door in each air lock closed; and

c.

Each penetration providing direct access from the containment atmosphere to the outside atmosphere either:

1.

closed by a manual or automatic isolation valve, blind flange, or equivalent, or

2.

capable of being closed by an OPERABLE Containment Purge and Exhaust Isolation System.

I APPLICABILITY:

During CORE ALTERATIONS, During movement of irradiated fuel assemblies within containment.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

One or more containment A.1 Suspend CORE Immediately penetrations not in ALTERATIONS.

required status.

AND A.2 Suspend movement of Immediately irradiated fuel assemblies within containment.

Farley Units 1 and 2 3.9.3-1 Amendment No. 165 (Unit 1)

Amendment No. 157 (Unit 2)

I No Changes.

Info Only I NL-07-0067, Enclosure 2, Page 2 of 9 Containment Penetrations 3.9.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.3.1 Verify each required containment penetration is in the 7 days required status.

SR 3.9.3.2 Verify each required containment purge and exhaust 18 months valve actuates to the isolation position on an actual or simulated actuation signal.

SR 3.9.3.3

NOTE 7 days Only required for an open equipment hatch.

Verify the capability to install the equipment hatch.

Farley Units 1 and 2 3.9.3-2 Amendment No. 165 (Unit 1)

Amendment No. 157 (Unit 2)

NL-07-0067, Enclosure 2, Page 3 of 9 Containment Penetrations B 3.9.3 B 3.9 REFUELING OPERATIONS B 3.9.3 Containment Penetrations BASES BACKC limited to dose consi within res limi

ROUND During CORE ALTERATIONS or movement of irradiated fuel assemblies within containment, a release of fission product radioactivity within containment will be recdtdrzmzopge4h#

wwhen the LCO requirements are met. In MODES 1, 2, 3, and 4, this is accomplished by maintaining containment OPERABLE maintain as described in LCO 3.6.1, "Containment." In MODE 6, the potential euencess for containment pressurization as a result of an accident is not likely; queatos therefore, requirements to isolate the containment from the outside julatowy atmosphere can be less stringent. The LCO requirements are ts referred to as "refueling integrity" rather than "containment OPERABILITY." Refueling integrity means that all potential escape paths are closed or capable of being closed. Since there is no potential for containment pressurization, the 10 CFR 50, Appendix J leakage criteria and tests are not required.

The containment serves to contain fission product radioactivity that may be released from the reactor core following an accident, such that offsite radiation exposures are maintained well within the requirements of 10 CFR 100. Additionally, the containment provides radiation shielding from the fission products that may be present in the containment atmosphere following accident conditions.

The containment equipment hatch, which is part of the containment pressure boundary, provides a means for moving large equipment and components into and out of containment. If closed, the equipment hatch must be held in place by at least four bolts. Good engineering practice dictates that the bolts required by this LCO be approximately equally spaced. Alternatively, the equipment hatch can be open provided it can be installed with a minimum of four bolts holding it in place.

The containment air locks, which are also part of the containment pressure boundary, provide a means for personnel access during MODES 1, 2, 3, and 4 unit operation in accordance with LCO 3.6.2, "Containment Air Locks." Each air lock has a door at both ends. The doors are normally interlocked to prevent simultaneous opening when containment OPERABILITY is required. During periods of unit shutdown (continued)

Farley Units 1 and 2 B 3.9.3-1 Revision 26

NL-07-0067, Enclosure 2, Page 4 of 9 Containment Penetrations B 3.9.3 limited to maintain dose consequences within regulatory limits.

BASES BACKGROUND when refueling integrity is not required, the door interlock mechanism (continued) may be disabled, allowing both doors of an air lock to remain open for extended periods when frequent containment entry is necessary.

During CORE ALTERATIONS or movement of irradiated fuel assemblies within containment, refueling integrity is required; therefore, the door interlock mechanism may remain disabled, but one air lock door must always remain dosed.

/A The requirements for refueling integrity ensure that a rele se capableof fission roduct radioactivi within containment will be restricted from

escaping e evironment Tei g

rncions are sufficient to restrict fission product radioactivity release from containment due to afuel handling accident during refueling.

The Containment Purge and Exhaust System includes two subsystems. The normal subsystem includes a 48-inch purge penetration and a 48-inch exhaust penetration. The second subsystem, a minipurge system, includes an 8-inch purge and an 8 inch exhaust line that utilize the 48-inch penetrations. During MODES 1, 2, 3, and 4, the two 48-inch purge valves in each of the normal purge and exhaust penetrations are secured in the closed position. The two 8-inch minipurge valves in each of the two minipurge lines may be opened in these MODES in accordance with LCO 3.6.3, "Containment Isolation Valves," but are closed automatically by the Engineered Safety Features Actuation System (ESFAS) instrumentation specified in LCO 3.3.6, "Containment Purge and Exhaust Isolation Instrumentation." Neither of the subsystems is subject to a Specification in MODE 5.

In MODE 6, large air exchanges are necessary to conduct refueling operations. The normal 48-inch purge system is used for this purpose, and all four valves are closed by the ESFAS instrumentation specified in LCO 3.3.6, "Containment Purge and Exhaust Isolation Instrumentation."

The minipurge system is not normally used in MODE 6. However, if the minipurge valves are opened they are capable of being closed automatically by the instrumentation specified in LCO 3.3.6, "Containment Purge and Exhaust Isolation Instrumentation."

The other containment penetrations that provide direct access from containment atmosphere to outside atmosphere must be isolated on at least one side. Isolation may be achieved by a closed automatic (continued)

Farley Units 1 and 2 B 3.9.3-2 Revision 0

NL-07-0067, Enclosure 2, Page 5 of 9 Containment Penetrations B 3.9.3 BASES BACKGROUND (continued) isolation valve, a manual isolation valve, blind flange, or equivalent.

Equivalent isolation methods allowed under the provisions of 10 CFR 50.59 may include use of a material that can provide a temporary, atmospheric pressure, ventilation barrier for the other containment penetrations during CORE ALTERATIONS or movement of irradiated fuel assemblies within containment (Ref. 1).

APPLICABLE SAFETY ANALYSES During CORE ALTERATIONS or movement of irradiated fuel assemblies within containment, the most severe radiological consequences result from a fuel handling accident. The fuel handling accident is a postulated event that involves damage to irradiated fuel (Ref. 2). The fuel handling accident analyzed includes dropping a single irradiated fuel assembly. The requirements of LCO 3.9.6, "Refueling Cavity Water Level," and the minimum decay time of 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months prior to CORE ALTERATIONS ensure that the release of fission product radioactivity, subsequent to a fuel handling accident, results in doses that are well within the guideline values specified in 10 CFR 100. Standard Review Plan, Section 15.7.4, Rev. I (Ref. 3),

defines "well within" 10 CFR 100 to be 25% or less of the 10 CFR 100 values. The acceptance limits for offsite radiation exposure will be 25% of 10 CFR 100 values.

Containment penetrations satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii).

LCO This LCO limits the consequences of a fuel handling accident 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 purge and exhaust penetrationseo

-NrheteW For the OPERABLE containment purge and exhaust penetrations, this LCO ensures that these penetrations are isolable by the Containment Purge and Exhaust Isolation System. For the r-qWteh, closure capability is provided by a designated trained osure crew and the necessary equipment. The OPERABILITY requirements for LCO 3.3.6, "Containment Purge and m ent Exhaust Isolation Instrumentation," ensure that the automatic purge the and exhaust valve closure times specified in the FSAR can be air equipment hatch and personnel air

locks, Ithe equip hatch and personnel locks.

(continued)

Farley Units 1 and 2 B 3.9.3-3 Revision 26

NL-07-0067, Enclosure 2, Page 6 of 9 Containment Penetrations B 3.9.3 Iand Dersonnel airI an locks are and personnel air BASES locks LCO aachieved and, therefore, meet the assumptions used in the safety (continued) janalysis to ensure that releases through the valves are terminated,

.such that radiological doses are within the acceptance limit.

d personnel air v

locks door The equipment hatch rconsidered isolable when the following criteria openings are satisfied:

1.

the necessary equipment required to close the hatch is available,

2.

at least 23 feet of water is maintained over the top of the reactor vessel flange in accordance with Specification 3.9.6,

3.

a desianated trained hetclosure crew is available.

The equipment hatc 4*4n must be capable of being cleared of any obstruction so that closure can be achieved as soon as possible.

APPLICAE B ILITY /

s I

Insert B 3.9.3 I The containment penetration requirements are applicable during CORE ALTERATIONS or movement of irradiated fuel assemblies within containment because this is when there is a potential for a fuel handling accident. In MODES 1, 2, 3, and 4, containment penetration requirements are addressed by LCO 3.6.1. In MODES 5 and 6, when CORE ALTERATIONS or movement of irradiated fuel assemblies within containment are not being conducted, the potential for a fuel handling accident does not exist. Therefore, under these conditions no requirements are placed on containment penetration status.

ACTIONS A.1 and A.2 If the containment equipment hatch, air locks, 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 Purge and Exhaust Isolation System not capable of automatic actuation when the purge and exhaust valves are open, the unit must be placed in a condition where the isolation function is not needed. This is accomplished by immediately suspending CORE ALTERATIONS and movement of irradiated fuel assemblies within containment. Performance of these actions shall not preclude completion of movement of a component to a safe position.

(continued)

Farley Units 1 and 2 B 3.9.3-4 Revision 26

NL-07-0067, Enclosure 2, Page 7 of 9 Insert B 3.9.3 The containment personnel air lock and emergency personnel air lock doors may be open during movement of irradiated fuel in the containment and during CORE ALTERATIONS provided that one door in each air lock is capable of being closed in the event of a fuel handling accident. Should a fuel handling accident occur inside containment, one door in each personnel air lock will be closed following an evacuation of containment.

The closure of the equipment hatch and the personnel air locks will be completed promptly following a fuel handling accident within containment.

NL-07-0067, Enclosure 2, Page 8 of 9 Containment Penetrations B 3.9.3 BASES SURVEILLANCE REQUIREMENTS SR 3.9.3.1 This Surveillance demonstrates that each of the containment penetrations required to be in its dosed position is in that position. The Surveillance on the open purge and exhaust 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 purge and exhaust isolation signal.

The Surveillance is performed every 7 days during CORE ALTERATIONS or movement of irradiated fuel assemblies within containment. The Surveillance interval is selected to be commensurate with the normal duration of time to complete fuel handling operations. A surveillance before the start of refueling operations will provide two or three surveillance verifications during the applicable period for this LCO. As such, this Surveillance ensures that a postulated fuel handling accident that releases fission product radioactivity within the containment will not result in SR 3.9.3.2 radiological doses in excess of those recommended by RG 1.195 (Reference 4.)

This Surveillance demonstrates that each containment purge and exhaust valve actuates to its isolation position on manual initiation or on an actual or simulated high radiation signal from each of the containment purge radiation monitoring instrumentation channels.

The 18 month Frequency maintains consistency with other similar ESFAS instrumentation and valve testing requirements. In LCO 3.3.6, the Containment Purge and Exhaust Isolation instrumentation requires a CHANNEL CHECK every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and a COT every 92 days to ensure the channel OPERABILITY during refueling operations. Every 18 months a CHANNEL CALIBRATION is performed. The system actuation response time is demonstrated every 18 months, during refueling, on a STAGGERED TEST BASIS.

SR 3.6.3.4 demonstrates that the isolation time of each valve is in accordance with the Inservice Testing Program requirements. These Surveillances performed during MODE 6 will ensure that the valves are capable of closing after a postulated fuel handling accident to limit a release of fission product radioactivity from the containment.

(continued)

Farley Units 1 and 2 B 3.9.3-5 Revision 26 I

NL-07-0067, Enclosure 2, Page 9 of 9 Containment Penetrations B 3.9.3 BASES SURVEILLANCE SR 3.9.3.3 REQUIREMENTS The equipment hatch is provided with a set of hardware, tools, and equipment for moving the hatch from its storage location and installing it in the opening. The required set of hardware, tools, and equipment shall be inspected to ensure that they can perform the required functions.

The 7 day frequency is adequate considering that the hardware, tools, and equipment are dedicated to the equipment hatch and not used for any other functions.

The SR is modified by a Note which only requires that the surveillance be met for an open equipment hatch. If the equipment hatch is installed in its opening, the availability of the means to install the hatch is not required.

REFERENCES

1. GPU Nuclear Safety Evaluation SE-0002000-001, Rev. 0, May 20, 1988.

2. FSAR, Section 15.4.5.

3. NUREG-0800, Section 15.7.4, Rev. 1, July 1981.

4. Regulatory Guide 1.195, "Methods and Assumptions for Evaluating Radiological Consequences of Design Basis Accidents at Light-Water Nuclear Power Reactors," May 2003 Farley Units 1 and 2 B 3.9.3-6 Revision 26

Joseph M. Farley Nuclear Plant, Units 1 and 2 License Amendment Request to Technical Specification 3.9.3, Containment Penetrations, Personnel Air Locks Doors Open During Fuel Movement Technical Specifications and Bases Clean Typed Pages List of Effected Pages TS Pages 3.9.3-1 TS Bases Pages B 3.9.3-1 B 3.9.3-2 B 3.9.3-3 B 3.9.3-4 B 3.9.3-5 B 3.6.3-6

Containment Penetrations 3.9.3 3.9 REFUELING OPERATIONS 3.9.3 Containment Penetrations LCO 3.9.3 APPLICABILITY:

The containment penetrations shall be in the following status:

a.

The equipment hatch is capable of being closed and held in place by four bolts;

b.

One door in each air lock is capable of being closed; and

c.

Each penetration providing direct access from the containment atmosphere to the outside atmosphere either:

1.

closed by a manual or automatic isolation valve, blind flange, or equivalent, or

2.

capable of being closed by an OPERABLE Containment Purge and Exhaust Isolation System.

During CORE ALTERATIONS, During movement of irradiated fuel assemblies within containment.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

One or more containment A.1 Suspend CORE Immediately penetrations not in ALTERATIONS.

required status.

AND A.2 Suspend movement of Immediately irradiated fuel assemblies within containment.

Farley Units 1 and 2 3.9.3-1 Amendment No.

Amendment No.

(Unit 1)

(Unit 2)

Containment Penetrations B 3.9.3 B 3.9 REFUELING OPERATIONS B 3.9.3 Containment Penetrations BASES BACKGROUND During CORE ALTERATIONS or movement of irradiated fuel assemblies within containment, a release of fission product radioactivity within containment will be limited to maintain dose consequences within regulatory limits when the LCO requirements are met. In MODES 1, 2, 3, and 4, this is accomplished by maintaining containment OPERABLE as described in LCO 3.6.1, "Containment."

In MODE 6, the potential for containment pressurization as a result of an accident is not likely; therefore, requirements to isolate the containment from the outside atmosphere can be less stringent. The LCO requirements are referred to as "refueling integrity" rather than "containment OPERABILITY." Refueling integrity means that all potential escape paths are closed or capable of being closed. Since there is no potential for containment pressurization, the 10 CFR 50, Appendix J leakage criteria and tests are not required.

The containment serves to contain fission product radioactivity that may be released from the reactor core following an accident, such that offsite radiation exposures are maintained well within the requirements of 10 CFR 100. Additionally, the containment provides radiation shielding from the fission products that may be present in the containment atmosphere following accident conditions.

The containment equipment hatch, which is part of the containment pressure boundary, provides a means for moving large equipment and components into and out of containment. If closed, the equipment hatch must be held in place by at least four bolts. Good engineering practice dictates that the bolts required by this LCO be approximately equally spaced. Alternatively, the equipment hatch can be open provided it can be installed with a minimum of four bolts holding it in place.

The containment air locks, which are also part of the containment pressure boundary, provide a means for personnel access during MODES 1, 2, 3, and 4 unit operation in accordance with LCO 3.6.2, "Containment Air Locks." Each air lock has a door at both ends. The doors are normally interlocked to prevent simultaneous opening when containment OPERABILITY is required. During periods of unit shutdown (continued)

Farley Units 1 and 2 B 3.9.3-1 Revision

Containment Penetrations B 3.9.3 BASES BACKGROUND when refueling integrity is not required, the door interlock mechanism (continued) may be disabled, allowing both doors of an air lock to remain open for extended periods when frequent containment entry is necessary.

During CORE ALTERATIONS or movement of irradiated fuel assemblies within containment, refueling integrity is required; therefore, the door interlock mechanism may remain disabled, but one air lock door must always remain capable of being closed.

The requirements for refueling integrity ensure that a release of fission product radioactivity within containment will be limited to maintain dose consequences within regulatory limits.

The Containment Purge and Exhaust System includes two subsystems. The normal subsystem includes a 48-inch purge penetration and a 48-inch exhaust penetration. The second subsystem, a minipurge system, includes an 8-inch purge and an 8 inch exhaust line that utilize the 48-inch penetrations. During MODES 1, 2, 3, and 4, the two 48-inch purge valves in each of the normal purge and exhaust penetrations are secured in the closed position. The two 8-inch minipurge valves in each of the two minipurge lines may be opened in these MODES in accordance with LCO 3.6.3, "Containment Isolation Valves," but are closed automatically by the Engineered Safety Features Actuation System (ESFAS) instrumentation specified in LCO 3.3.6, "Containment Purge and Exhaust Isolation Instrumentation." Neither of the subsystems is subject to a Specification in MODE 5.

In MODE 6, large air exchanges are necessary to conduct refueling operations. The normal 48-inch purge system is used for this purpose, and all four valves are closed by the ESFAS instrumentation specified in LCO 3.3.6, "Containment Purge and Exhaust Isolation Instrumentation."

The minipurge system is not normally used in MODE 6. However, if the minipurge valves are opened they are capable of being closed automatically by the instrumentation specified in LCO 3.3.6, "Containment Purge and Exhaust Isolation Instrumentation."

The other containment penetrations that provide direct access from containment atmosphere to outside atmosphere must be isolated on at least one side. Isolation may be achieved by a closed automatic (continued)

Farley Units 1 and 2 B 3.9.3-2 Revision

Containment Penetrations B 3.9.3 BASES BACKGROUND (continued) isolation valve, a manual isolation valve, blind flange, or equivalent.

Equivalent isolation methods allowed under the provisions of 10 CFR 50.59 may include use of a material that can provide a temporary, atmospheric pressure, ventilation barrier for the other containment penetrations during CORE ALTERATIONS or movement of irradiated fuel assemblies within containment (Ref. 1).

APPLICABLE SAFETY ANALYSES During CORE ALTERATIONS or movement of irradiated fuel assemblies within containment, the most severe radiological consequences result from a fuel handling accident. The fuel handling accident is a postulated event that involves damage to irradiated fuel (Ref. 2). The fuel handling accident analyzed includes dropping a single irradiated fuel assembly. The requirements of LCO 3.9.6, "Refueling Cavity Water Level," and the minimum decay time of 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months prior to CORE ALTERATIONS ensure that the release of fission product radioactivity, subsequent to a fuel handling accident, results in doses that are well within the guideline values specified in 10 CFR 100. Standard Review Plan, Section 15.7.4, Rev. 1 (Ref. 3),

defines "well within" 10 CFR 100 to be 25% or less of the 10 CFR 100 values. The acceptance limits for offsite radiation exposure will be 25% of 10 CFR 100 values.

Containment penetrations satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii).

LCO This LCO limits the consequences of a fuel handling accident 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 purge and exhaust penetrations, the equipment hatch and the personnel air locks. For the OPERABLE containment purge and exhaust penetrations, this LCO ensures that these penetrations are isolable by the Containment Purge and Exhaust Isolation System.

For the equipment hatch and personnel air locks, closure capability is provided by a designated trained closure crew and the necessary equipment. The OPERABILITY requirements for LCO 3.3.6, "Containment Purge and Exhaust Isolation Instrumentation," ensure that the automatic purge and exhaust valve closure times specified in the FSAR can be achieved and, therefore, meet the assumptions used in the safety achieved and, therefore, meet the assumptions used in the safety analysis to ensure that releases through the valves (continued)

Farley Units 1 and 2 B 3.9.3-3 Revision

Containment Penetrations B 3.9.3 BASES LCO (continued) are terminated, such that radiological doses are within the acceptance limit.

The equipment hatch and personnel air locks are considered isolable when the following criteria are satisfied:

1.

the necessary equipment required to close the hatch and personnel air locks is available,

2.

at least 23 feet of water is maintained over the top of the reactor vessel flange in accordance with Specification 3.9.6,

3.

a designated trained closure crew is available.

The equipment hatch and personnel air locks door openings must be capable of being cleared of any obstruction so that closure can be achieved as soon as possible.

The containment personnel air lock and emergency personnel air lock doors may be open during movement of irradiated fuel in the containment and during CORE ALTERATIONS provided that one door in each air lock is capable of being closed in the event of a fuel handling accident. Should a fuel handling accident occur inside containment, one door in each personnel air lock will be closed following an evacuation of containment.

The closure of the equipment hatch and the personnel air locks will be completed promptly following a fuel handling accident within containment.

APPLICABILITY The containment penetration requirements are applicable during CORE ALTERATIONS or movement of irradiated fuel assemblies within containment because this is when there is a potential for a fuel handling accident. In MODES 1, 2, 3, and 4, containment penetration requirements are addressed by LCO 3.6.1. In MODES 5 and 6, when CORE ALTERATIONS or movement of irradiated fuel assemblies within containment are not being conducted, the potential for a fuel handling accident does not exist. Therefore, under these conditions no requirements are placed on containment penetration status.

Farley Units 1 and 2 B 3.9.3-4 Revision

Containment Penetrations B 3.9.3 BASES ACTIONS A.1 and A.2 If the containment equipment hatch, air locks, 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 Purge and Exhaust Isolation System not capable of automatic actuation when the purge and exhaust valves are open, the unit must be placed in a condition where the isolation function is not needed. This is accomplished by immediately suspending CORE ALTERATIONS and movement of 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 purge and exhaust 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 purge and exhaust isolation signal.

The Surveillance is performed every 7 days during CORE ALTERATIONS or movement of irradiated fuel assemblies within containment. The Surveillance interval is selected to be commensurate with the normal duration of time to complete fuel handling operations. A surveillance before the start of refueling operations will provide two or three surveillance verifications during the applicable period for this LCO. As such, this Surveillance ensures that a postulated fuel handling accident that releases fission product radioactivity within the containment will not result in radiological doses in excess of those recommended by RG 1.195 (Reference 4).

SR 3.9.3.2 This Surveillance demonstrates that each containment purge and exhaust valve actuates to its isolation position on manual initiation or on an actual or simulated high radiation signal from each of the containment purge radiation monitoring instrumentation channels.

The 18 month Frequency maintains consistency with other similar (continued)

Farley Units 1 and 2 B 3.9.3-5 Revision

Containment Penetrations B 3.9.3 BASES SURVEILLANCE SR 3.9.3.2 (continued)

REQUIREMENTS ESFAS instrumentation and valve testing requirements. In LCO 3.3.6, the Containment Purge and Exhaust Isolation instrumentation requires a CHANNEL CHECK every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and a COT every 92 days to ensure the channel OPERABILITY during refueling operations. Every 18 months a CHANNEL CALIBRATION is performed. The system actuation response time is demonstrated every 18 months, during refueling, on a STAGGERED TEST BASIS. SR 3.6.3.4 demonstrates that the isolation time of each valve is in accordance with the Inservice Testing Program requirements. These Surveillances performed during MODE 6 will ensure that the valves are capable of closing after a postulated fuel handling accident to limit a release of fission product radioactivity from the containment.

SR 3.9.3.3 The equipment hatch is provided with a set of hardware, tools, and equipment for moving the hatch from its storage location and installing it in the opening. The required set of hardware, tools, and equipment shall be inspected to ensure that they can perform the required functions.

The 7 day frequency is adequate considering that the hardware, tools, and equipment are dedicated to the equipment hatch and not used for any other functions.

The SR is modified by a Note which only requires that the surveillance be met for an open equipment hatch. If the equipment hatch is installed in its opening, the availability of the means to install the hatch is not required.

REFERENCES

1. GPU Nuclear Safety Evaluation SE-0002000-001, Rev. 0, May 20, 1988.

2. FSAR, Section 15.4.5.

3. NUREG-0800, Section 15.7.4, Rev. 1, July 1981.

4. Regulatory Guide 1.195, "Methods and Assumptions for Evaluating Radiological Consequences of Design Basis Accidents at Light-Water Nuclear Power Reactors," May 2003.

Farley Units 1 and 2 B 3.9.3-6 Revision

Joseph M. Farley Nuclear Plant, Units 1 and 2 License Amendment Request to Technical Specification 3.9.3, Containment Penetrations, Personnel Air Locks Doors Open During Fuel Movement List of Regulatory Commitments

Joseph M. Farley Nuclear Plant, Units 1 and 2 License Amendment Request to Technical Specification 3.9.3, Containment Penetrations, Personnel Air Locks Doors Open During Fuel Movement List of Regulatory Commitments The following table identifies those actions committed to by SNC in this document. Any other statements in this submittal are provided for information purposes and are not considered to be regulatory commitments.

Regulatory Commitments Due Date / Event SNC procedures will ensure that the containment Prior to the implementation equipment hatch is installed and containment date after approval of this personnel air locks have at least one door closed amendment request by the promptly following a fuel handling accident inside NRC.

containment.

Joseph M. Farley Nuclear Plant, Units 1 and 2 License Amendment Request to Technical Specification 3.9.3, Containment Penetrations, Personnel Air Locks Doors Open During Fuel Movement Excerpt from SNC Atmospheric Dispersion Factors Calculation

4A SOUTHERN 5..

COMPANY Ems ~yrjggoSeYou r Wrlir NL-07-0067, Enclosure 5, Page 1 of 88 Southern Nuclear Design Calculation Calculation Number:

BM-03-0018-001 Plant:

Unit:

Discipline:

Farley Nuclear Plant r-11 [:2 Z1

&2 Mechanical!

Environmental Subject/Title:

Job Number:

Control Room and Technical Support Center Air Intake X/Q Estimates RER 03-018 Objective:

To determine X/Qs at the control room and Technical Support Center air intakes due to accidental releases following a LOCA.

System or Equipment Tag Numbers:

Contents Attachments (Computer Printouts, Technical Papers,

of Topic Page Sketches, Correspondence)

Pages Purpose of Calculation 2

A. Farley letter Log: PS-04-0441 2

Summary of Conclusions 2

B. FarlMet Program Listing 2

Criteria 2

C. ARCON96 Output (4 years met data), Ver. 3 36 Methodology 2

D. ARCON96 Output (4.5 years met data), Ver. 3 36 Assumptions 3

Input Data 5-9 References 9-10 Last Page Number:

12 Nuclear Quality Level Safety-Related Safety Significant El Non-Safety-Significant Notes: File 46.16

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1.0 VERSION HISTORY Version 1 of this calculation was performed using the updated meteorological data (2000, 2001, 2002 and 2003) provided by Farley Nuclear Plant's Computer Services (Farley Nuclear Plant, 2004). This version of the calculation also adopted methodology included by Regulatory Guide 1.194 (NRC, 2003). Additionally, minor corrections were made in some of the vertical distances used as input to the ARCON96 model as shown in Version 0.

Version 2 of the calculation is performed per a Southern Company's request (External Work Authorization, EWA# BF0002, Revision3, August 13, 2004) to include an extra set of six months of hypothetical meteorological data for a case involving potential Unit 1 hatch door release to Unit 1 control room (CR) air intake. See Section 6.1 for a detailed description of how this data was included in the re-analysis. In this version, X/Qs were calculated based on temperature gradients measured between 35-and 200-foot levels, while upper wind measurements were also assumed to be measured at the 200-foot level.

Version 3 of the calculation was performed based on Southern Company's request (External Work Authorization, EWA# BF0002, Revision 4, September 8, 2004). At the onsite meteorological tower, the temperature gradients are measured between the 35-and 200-foot levels; the wind measurements are made at 35-and 150-foot levels. In Version 3, the correct upper wind measurement height of 150 ft was used. The horizontal and vertical distances between the release heights and the receptors were also refined to improve the accuracy of the X/Q estimates. Additionally in Version 3 of the calculation, besides 4 years (2000-2003) of meteorological data, the 6 months of hypothetical meteorological data were also considered for all the cases studied (see Section 6.1 for detail).

2.0 PURPOSE OF CALCULATIONS/

SUMMARY

OF CONCLUSIONS The purpose of this calculation is to determine the relative atmospheric dispersion factor (X/Q) at the main CR air intakes and the Technical Support Center (TSC) air intake of the Farley Nuclear Power Station due to postulated releases from the reactor buildings, vent stacks, and hatch doors following a Loss of Coolant Accident (LOCA).

The results indicate that among Units 1 and 2 CR air intakes and the TSC air intake, the maximum X/Q values are all associated with the uniform releases from the Unit 2 reactor building. Among the three air intakes, the TSC intake has the highest X/Q value for each averaging time period due to the relatively short distance between the TSC intake and the Unit 2 reactor.

3.0 CRITERIA In 1997 the NRC issued a new calculation procedure coded in the computer model, ARCON96 (NUREG/CR-633 1, Rev. 1, 1997), to estimate X/Q values in building wake regions. This new model results in the estimation of more representative X/Q values through the use of improved approaches in accounting for

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dispersion associated with complex wake effects in the vicinity of building structures. The analysis in this calculation uses the current version (Rev. 1, 1997) of the ARCON96 model.

4.0 METHODOLOGY The ARCON96 model uses hourly meteorological data to estimate dispersion in the vicinity of wake producing structures and to calculate relative concentrations (X/Qs) at air intakes that would be exceeded no more than five percent (5%) of the time. These concentrations are calculated for averaging periods ranging from one hour to 30 days.

ARCON96 evaluates ground level, vent, and elevated releases. A vent release is one that takes place through a rooftop vent with an uncapped vertical opening. Building wake effects are also considered in the model for estimating X/Q values from ground-level releases. Momentum rise and thermal plume rise are not' considered in calculating the effective release height in the model. Additionally, under calm wind conditions, the receptor location is assumed to be directly downwind of the release point. Considering the release height, the receptor height, and the horizontal distance from the release point to the receptor, the model will calculate a "slant range distance" as the straight-line distance between the release point and the receptor.

ARCON96 incorporates a method to handle area source releases. For area sources with a defined height and width (such as reactor building releases), the initial horizontal and vertical diffusion coefficients can be defined as (NRC 2003):

y = structure width / 6 (Y = structure height / 6 This approach is more conservative than that suggested in NUREG/CR-6331 (NRC, 1997) using a factor of 4.3

(+/-2.15) for estimating the above values. An averaging sector width constant of 4.3 suggested in Regulatory Guide 1.194 (NRC, 2003) was used in lieu of the default value of 4.0 as suggested in NUREG/CR-6331 (1997). Additionally, a surface roughness length of 0.2 suggested by Regulatory Guide 1. 194 was used in lieu of the default value of 0.1 as suggested in NUREG/CR-6331 (NRC, 1997).

A personal computer (Bechtel Configuration Number: FREW52526, Compaq Deskpro EN, Windows 2000) was used to run the ARCON96 program for this analysis. For verification purposes, the program was run using sample input data and a set of meteorological data provided in NUREG/CR-6331 (NRC, 1997, p. 22).

Since the results of the test run matched the output provided in NUREG/CR-6331 (NRC, 1997, p. 23), the ARCON96 program was properly validated.

5.0 ASSUMPTIONS 5.1 Ground-Level Release

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,I The postulated release locations at the reactor buildings, reactor vents, and hatch doors are shown schematically in Figure 1. These releases were assumed to be caught in the aerodynamic building wake and were treated as ground-level releases.

5.2 Distances between Release Points and the Air Intakes To be conservative, releases from the reactor buildings were assumed to occur at the same height as the CR and TSC air intakes and to be located directly upwind of them. With this orientation, potential releases from the reactor building will have the shortest horizontal distance to travel toward the air intakes.

Releases from the reactor vents were assumed to travel directly toward the air intakes following the slant range path calculated by the ARCON96 model. This approach is conservative because it results in a shorter overall distance between release point and air intake.

The potential releases from the vents could reach the TSC air intake by going around the southern side of the Unit 2 reactor building. The physical traveling distance between the Unit 2 Vent and the TSC air intake was accounted for instead of the straight-line distance between them. Because there is no structure to block the path between the TSC air intake and the Unit 1 Vent, a straight-line distance was used as the distance between them.

With respect to the CR and TSC air intakes, the hatch doors are located on the opposite side of the reactor buildings. In order for the hatch door releases to reach the air intakes, effluents need to travel around the reactor building into the wakes at the lee side of the buildings. However, straight-line distances between the hatch door and the air intakes were used in the dispersion modeling. Compared to traveling around the buildings, this assumption yields a shorter distance between the release points and the air intake and higher X/Q values; therefore, the assumption is conservative.

5.3 Building Cross-Sectional Areas Used for Wake Analysis For hatch door releases, only the portion of the reactor building area (A) that is higher than the release point was considered in the wake analysis. This approach is conservative, because smaller building cross-sectional areas would produce higher X/Q values.

For reactor releases, only the building area that is higher than the air intake height was considered in the wake analysis. The CR and TSC air intakes are located 37.5' and 22.5' above grade, respectively (see Drawing Nos.

D-1751 10, Sheet 1, Ver. 33 and D-205574, Ver. 6). Similarly, this approach is conservative because smaller building cross-sectional areas would produce higher X/Q values.

The releases described above could be recirculated in the wake region and brought into the CR air intakes and the TSC air intake. For vent releases, no building wake effect was considered because under very low wind speeds, the vent releases could reach the CR or TSC intakes without the influence of the wake effect. This

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approach is conservative as the releases under wake effect usually produce lower X/Q values than when no wake effect is assumed.

5.4 Source Types Two types of release modes were considered in this calculation. For an area release, ARCON96 allows the use of the initial diffusion coefficients to simulate the source as a virtual point source. In this study, for uniform releases occurring on the reactor surface, an area source release was assumed because there is no single point from which the release can be assumed on the relatively large reactor building. Similarly, because the vent diameter is relatively small, it is conservative to assume that the vent releases are point sources. For hatch door releases, although the hatch door has a diameter of about 6.45 m (see Drawing No. D-176120, Ver. 3), it was conservatively assumed as a point source with no initial diffusion coefficients.

5.5 Source Release Height For potential containment releases, the height of the release point was assumed to be the same as the receptor height. This approach is conservative because this assumption would result in the shortest distance between the release point and the corresponding receptor. For example, the CR and TSC air intake heights are 37.5 ft (11.4 m) and 22.5 ft (6.9 m), respectively. Therefore, for the potential reactor release scenarios, the release heights were assumed to be 37.5 ft and 22.5 ft for the CR and TSC air intakes, respectively.

6.0 INPUT DATA 6.1 Meteorological data For Version 1 of this calculation, four years (2000-2003) of continuous hourly on-site raw meteorological data (i.e., wind speed, wind direction from the 10-meter and 61-meter (35-and 200-foot) levels, and stability class) were provided by Plant Farley (see Attachment A). These hourly data were used in the prepared input to the ARCON96 to calculate X/Q values in this analysis. The raw monthly meteorological data were combined into a yearly data set and converted to the ASCII format using MS Word to get rid of unwanted characters, such as slashes (/) and colons (:). A Fortran code (FarlMet) was developed to process the ASCII meteorological data set and generate an ARCON96-ready meteorological input data set (see Attachment B).

Del H = 35'- 200' =165' = 50.292 m 1

2 3

4 DeIT (F)

DT/Dz (F/fl)

DT/Dz (C/m)

DT/Dz (C/i 00m)

-1.74

-0.0105

-0.0192

-1.9

-1.56

-0.0095

-0.0172

-1.7

-1.38

-0.0084

-0.0152

-1.5

-0.46

-0.0028

-0.0051

-0.5 1.38 0.0084 0.0152 1.5 3.6 0.0218 0.0398 4.0

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Raw meteorological data format is described in a Farley Nuclear Plant documentation (FNP-0-ENV-17). The wind speed is in miles per hour (mph) and the temperature gradient (degree F/ft) is measured between 35-ft and 200-foot levels above grade on the FNP meteorological tower. The atmospheric stability classes are determined by a set of criteria specified in FNP-0-ENV-17. This set of criteria (see table below) satisfies the guideline provided in Regulatory Guide 1.23 (NRC, 1980), and is also used in the FarlMet program to determine stability classes.

In the above table, Del H is the difference in height between the two measuring levels, DeIT is the difference in temperature between the two levels, and DT/Dz is the temperature gradient. Column 1 lists criteria used in the FarlMet program, while Column 4 lists values presented in Regulatory Guide 1.23 (NRC, 1980). Values presented in Column 1 are equivalent to values presented in Column 4. Atmospheric stability classes implemented in the FarlMet program are shown in the following table.

Stability Classification Pasquill Categories Temperature change between 35 ft and 200 ft Extremely unstable A

< -1.74 Moderately unstable B

-1.74 to -1.56 Slightly unstable C

-1.56 to -1.38 Neutral D

-1.38 to -0.46 Slightly stable E

-0.46 to 1.38 Moderately stable F

1.38 to 3.6 Extremely stable G

> 3.6 For Version 2 of this calculation and as mentioned in Section 1.0, an extra 6 months of hypothetical meteorological data were considered in the modeling run in addition to the 4-year (2000-2003) data analyzed for Version 1. This 6-month data were constructed based on 1-month of December 1999 data repeated for 6 times. The one month data for December 1999 meteorological data set was provided by Southern Company (E-mail from J. Wehrenberg to B. Singal). Data collected before 2000 seem to have slightly higher frequency of southerly winds. Because the Unit 1 hatch door is located to the SE of the Unit 1 CR air intake (see Figure 1), the inclusion of the extra southerly winds is expected to somewhat increase the X/Q values at the Unit 1 CR air intake due to releases from the Unit 1 hatch door. Version 2 of this calculation was to study the effects on X/Q values for the above source-receptor scenario with the increase of the southerly wind frequency.

Besides using the revised upper wind measurement height, both sets of meteorological data (4-year and 4.5-year) were used to study all cases involved in Version 3 of the calculation.

6.2 Plant Data Drawings used to obtained plant data are listed in the following table.

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Description Drawing Number CR Air Intake Locations D-1751 10, Sheet 1, Ver. 33 Reactor Vent and Hatch Door Locations D-176006, Ver. 33 (Unit 1) and D-206006, Ver. 36 (Unit 2)

Hatch Door Elevations D-176120, Ver. 3 (Unit 1) and D-206120, Ver. 2 (Unit 2)

TSC Air Intake Location D-205574, Ver. 6 Reactor building height Drwg No. D-205068, Ver. 4 Distances between the release points and air intakes are presented below.

Horizontal Vertical Horizontal Vertical Direction to Receptor Release Point Distance (ft) Distance (ft) Distance (m) Distance (m) Source (de ree)

Unit 1 CR Air Intake Unit 1 Vent 198.8 145.5 60.6 44.3 116 Reactor 128.5 37.5 39.2 11.4 136 Hatch Door 263.7 8.75 80.4 2.7 136 Unit 1 CR Air Intake Unit 2 Vent 200.6 145.5 61.1 44.3 66 Reactor 128.5 37.5 39.2 11.4 42 Hatch Door 269.2 8.75 82.1 2.7 43 Horizontal Vertical Horizontal Vertical Direction to Receptor Release Point Distance (ft) Distance (ft) Distance (m) Distance (m) Source (degree)

Unit 2 CR Air Intake Unit 1 Vent 200.6 145.5 61.1 44.3 116 Reactor 128.5 37.5 39.2 11.4 138 Hatch Door 269.2 8.75 82.1 2.7 137 Unit 2 CR Air Intake Unit 2 Vent 198.9 145.5 60.6 44.3 67 Reactor 128.5 37.5 39.2 11.4 45 Hatch Door 263.7 8.75 80.4 2.7 45 Receptor Horizontal Vertical Horizontal Vertical Direction to Release Point Distance (ft) Distance (ft) Distance (m) Distance (in) Source (degree)

TSC Air Intake Unit 1 Vent 230.3 145.5 70.2 44.3 141 Reactor 189.1 22.5 57.6 6.9 157 Hatch Door 330.4 8.75 100.7 2.7 157 TSC Air Intake Unit 2 Vent 153.5 145.5 46.8 44.3 96 Reactor 44.5 22.5 13.6 6.9 66 Hatch Door 176.9 8.75 53.9 2.7 56

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The elevations of air intakes and release points are presented in the following table.

Location Elevation (ft)

Elevation (m)

Ht above grade (ft)

Ht. above grade (m)

CR Intake 192 58.5 37.5 11.4 TSC Intake 177 53.9 22.5 6.9 Vent Stack 300 91.4 145.5 44.3 Hatch Door 163.25 49.8 8.75 2.7 6.3 Effective Wind Directions The direction to the source from the intake and "wind direction window width" are part of the input requirements for the ARCON96 program. As suggested in NUREG/CR-6331 (NRC, 1997), the wind direction window sector size was chosen as the direction from the receptor to the source, +/-45'. The "direction to source" for each combination of release point and receptor (air intake) is presented in the table shown on Sheet 7.

6.4 Building Areas Building areas used for wake analysis were calculated based on the assumptions described in Section 5.3.

Reactor Building Release (Receptor: CR air intake)

For building area (A) calculation, the dome-shaped reactor elevation was assumed to be 261.7'. This elevation is about the mid-point of the dome in vertical direction (see) Drwg No. D-205068, Ver. 4.

CR intake height = 192' (Drwg. No. D-1751 10, Sheet 1, Ver. 33)

Reactor diameter = 137' (Drwg. No. D-176006, Ver. 32)

A = (261.7' - 192') x 137' = 9548.9 ft2 = 887 M2 Reactor Building Release (Receptor: TSC air intake)

TSC intake height = 177' (Drwg. No. D-205574, Ver. 6)

Reactor diameter = 137' (Drwg. No. D-176006, Ver. 32)

A = (261.7' - 177') x 137' = 11603.9 ft2 = 1078 m2 Hatch Door Release (Receptor: CR & TSC air intakes)

NL-07-0067, Enclosure 5, Pa e 9 of 88 Southern Company Services Design Calculations - Nuclear Project Calculation Number Farley Units I and 2 BM-03-0018-001 Subject/Title Control Room and Technical Support Center Air Intake X/Q Estimates Sheet [9]

of

[12] ),I Hatch door elevation = 163.25' (Drwg. No. D-176120, Ver. 3)

A = (261.7' - 163.25') x 137' = 13487.6 ft2 = 1253 m 2 6.5 Initial Diffusion Coefficients No initial diffusion coefficients were used for vent or hatch door releases. This approach is conservative because the use of initial diffusion coefficients would reduce the X/Q values somewhat. Initial diffusion coefficients were used for reactor releases are presented in the following table.

Sources Reactor Receptor Receptor Reactor Sigma Y Sigma Z Height (ft)

Height (ft)

Diameter (ft)

(m)

Reactor 261.7 CR air 192 137 6.97 3.54 intake Reactor 261.7 TSC air 177 137 6.97 4.30 1_______

1___

intake I

I II_

I Notes: Sigma Y = Reactor Diameter / 6; Sigma Z = (Reactor height - Receptor height) / 6 CR = control room; TSC = Technical Support Center

7.0 REFERENCES

1)

NRC, 2003, Regulatory Guide 1.194, Atmospheric Relative Concentrations for Control Room Radiological Habitability Assessments at Nuclear Power Plants, U.S. NRC, June.

2)

NRC, 1997, Atmospheric Relative Concentrations in Building Wakes, NUREG/CR-6331, Rev. 1, Pacific Northwest Laboratory, prepared for U.S. NRC.

3)

FNP-0-ENV-17, Met Tower Data Review, Version 25.0, Farley Nuclear Plant.

4)

Letter from Mark J. Ajiluni (Farley) to B. K. Singal (Bechtel) "Farley Nuclear Plant - Unit 1 and 2, Meteorological Data", dated March 30, 2004, File: RER 03-018, Log: PS-04-0441, Southern Company (Attachment A).

5)

NRC, 1980, Regulatory Guide 1.23, proposed Revision 1, Meteorological Program in Support of Nuclear Power Plants, U.S. NRC, September.

6)

External Work Authorization, EWA# BF0002, Version 3,

Subject:

Farley Atmospheric Dispersion Factor Calculation BM-03-0018-001 for the MCR and Technical Support Center, 8/13/2004.

7)

E-mail from J. Wehrenberg (Southern Company) to B. Singal (Bechtel),

Subject:

six months of adjusted meteorological data, dated March 25, 2004.

Design Calculations - Nuclear NL-07-0067, Enclosure 5, Page10 of 88 Southern Company Services I Project Calculation Number Farley Units 1 and 2 BM-03-0018-001 Subject/Title Control Room and Technical Support Center Air Intake X/Q Estimates I sheet [10]

of

[12]

8)

External Work Authorization, EWA# BF0002, Version 4,

Subject:

Farley Atmospheric Dispersion Factor Calculation BM-03-0018-001 for the MCR and Technical Support Center, 9/8/2004.

8.0 RESULTS 8.1 Version 3 Results (based on 4 years of meteorological data)

The ARCON96 modeling outputs based on 4 years of meteorological data are presented in Attachment C for releases originating at the reactor buildings, reactor vents, and hatch doors. The results from the above outputs are summarized in the following table.

X/Q Values at the CR and TSC Air Intakes (4 years of meteorological data)

Release Point Receptor 0 - 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 2 - 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 8 - 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 1 - 4 days 4 - 30 days U1 Vent U1 CR 1.61 E-03 1.26E-03 5.65E-04 3.43E-04 2.32E-04 Ul Reactor U1 CR 1.51 E-03 9.01 E-04 3.95E-04 2.75E-04 1.91 E-04 U1 Hatch Door U1 CR 8.39E-04 5.11E-04 2.09E-04 1.47E-04 9.35E-05 U2 Vent U1 CR 1.64E-03 1.37E-03 7.17E-04 5.41 E-04 3.60E-04 U2 Reactor Ul CR 1.66E-03 1.36E-03 6.81 E-04 5.60E-04 4.21 E-04 U2 Hatch Door Ul CR 7.95E-04 6.73E-04 3.35E-04 2.48E-04 1.87E-04 U1 Vent U2 CR 1.59E-03 1.25E-03 5.54E-04 3.34E-04 2.27E-04 Ul Reactor U2 CR 1.51 E-03 8.91 E-04 3.91 E-04 2.71 E-04 1.87E-04 U1 Hatch Door U2 CR 8.04E-04 4.90E-04 2.01 E-04 1.39E-04 9.04E-05 U2 Vent U2 CR 1.65E-03 1.38E-03 7.20E-04 5.47E-04 3.63E-04 U2 Reactor U2 CR 1.65E-03 1.34E-03 6.75E-04 5.40E-04 4.03E-04 U2 Hatch Door U2 CR 8.33E-04 6.98E-04 3.43E-04 2.57E-04 1.91 E-04 U1 Vent TSC 1.15E-03 8.19E-04 3.44E-04 2.1OE-04 1.49E-04 Ul Reactor TSC 8.17E-04 4.64E-04 2.OOE-04 1.41 E-04 9.26E-05 U1 Hatch Door TSC 5.47E-04 2.95E-04 1.27E-04 8.84E-05 5.44E-05 U2 Vent TSC 2.11 E-03 1.72E-03 8.13E-04 5.25E-04 3.55E-04 U2 Reactor TSC 5.06E-03 3.19E-03 1.45E-03 1.31 E-03 1.01 E-03 U2 Hatch Door TSC 1.77E-03 1.51 E-03 7.26E-04 5.09E-04 3.95E-04 Notes:

Ul - Unit 1; U2 - Unit 2; CR - Control Room; TSC - Technical Support Center 8.2 Version 3 Results (based on 4.5 years of meteorological data)

Using 4.5 years of meteorological data (i.e., 4 year + 1-month x 6), ARCON96 modeling analyses were performed for all cases considered in Section 8.1. The results are presented in Attachment D and are summarized in the following table.

I I

t I

NL-07-0067, Enclosure 5, Page 11 of 88 Southern Company Services A Design Calculations - Nuclear Project Calculation Number Farley Units I and 2 BM-03-0018-001 Subject/Title Control Room and Technical Support Center Air Intake X/Q Estimates Sheet [11]

of

[12]

X/Q Values at the CR and TSC Air Intakes (4.5 years of meteorological data)

I Release Point Receptor 0 - 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 2 - 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 8 - 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 1 - 4 days 4 - 30 days U1 Vent Ul CR 1.62E-03 1.21 E-03 5.37E-04 3.35E-04 2.32E-04 U1 Reactor U1 CR 1.54E-03 9.62E-04 4.45E-04 2.61 E-04 3.09E-04 U1 Hatch Door Ul CR 8.79E-04 5.89E-04 2.63E-04 1.57E-04 1.93E-04 U2 Vent U1 CR 1.59E-03 1.35E-03 7.08E-04 5.16E-04 3.59E-04 U2 Reactor U1 CR 1.64E-03 1.34E-03 6.65E-04 5.45E-04 4.18E-04 U2 Hatch Door U1 CR 7.83E-04 6.52E-04 3.22E-04 2.43E-04 1.85E-04 U1 Vent U2 CR 1.60E-03 1.1 9E-03 5.23E-04 3.24E-04 2.28E-04 U1 Reactor U2 CR 1.54E-03 9.50E-04 4.43E-04 2.70E-04 3.11 E-04 U1 Hatch Door U2 CR 8.56E-04 5.67E-04 2.50E-04 1.52E-04 1.92E-04 U2 Vent U2 CR 1.60E-03 1.37E-03 7.1 OE-04 5.21 E-04 3.60E-04 U2 Reactor U2 CR 1.64E-03 1.32E-03 6.60E-04 5.27E-04 4.01 E-04 U2 Hatch Door U2 CR 8.1 8E-04 6.77E-04 3.32E-04 2.49E-04 1.89E-04 U1 Vent TSC 1.23E-03 8.68E-04 3.65E-04 2.17E-04 2.57E-04 U1 Reactor TSC 8.74E-04 5.97E-04 3.04E-04 2.59E-04 2.86E-04 U1 Hatch Door TSC 5.94E-04 4.43E-04 2.1 8E-04 1.73E-04 1.96E-04 U2 Vent TSC 2.07E-03 1.67E-03 7.90E-04 5.14E-04 3.54E-04 U2 Reactor TSC 5.03E-03 3.08E-03 1.43E-03 1.29E-03 9.95E-04 U2 Hatch Door TSC 1.75E-03 1.46E-03 7.0159E-04 4.98E-04 3.87E-04 Notes:

U1 - Unit 1; U2 - Unit 2; CR - Control Room; TSC - Technical Support Center As shown in the above table, compared to the use of 4 years of meteorological data, the X/Qs have decreased for releases involving Unit 2 vent, reactor and equipment hatch door. This is because the extra six months of meteorological data contain more southerly frequency, and Unit 2 vent, reactor and hatch door are located to the north of the receptors (Unit 1 CR, Unit 2 CR and TSC). Subsequently, the overall frequency for releases involving Unit 2 vent, reactor or hatch door has reduced accordingly. Compared to the use of 4 years of meteorological data, the X/Qs have increases somewhat for releases involving the Unit 1 vent, reactor and hatch door. Among the three release locations, Unit 1 hatch door release has resulted in the highest X/Q increase.

This is because the Unit 1 hatch door is located upwind of the considered receptors under southerly winds. For the most critical averaging period (0-2 hours) during the LOCA, the increases due to Unit 1 hatch door release are 4.8%, 6.5% and 8.6% at Unitl CR intake, Unit 2 CR intake and TSC intake, respectively.

Design Calculations - Nuclear NL-07-0067, Enclosure 5, Page 12 of 88 Southern Company Services A Project Calculation Number Farley Units I and 2 BM-03-0018-001 SubjectrTitle Control Room and Technical Support Center Air Intake x/Q Estimates Sheet [ 12) of,[12]

Figure 1 TSC Intake 0

Air Intake Locations and Release Points Hatch Door North Vent Unit 2 CR Intake 0

0 Unit 1 CR Intake Vent Hatch Door Note: Release point from the reactor was a point on the containment surface closest to the receptor location.

NL-07-0067, Enclosure 5, Page 13 of 88 Calculation Number: BM-03-0018-001, Version I Sheet I of 2 Attachment A.

SOUTHERN COMPANY F-nr'p to Serve.Uur WPrU-March 30, 2004 Farley Nuclear Plant - Unit 1 and 2 File: RER 03-018 Meteorololical Data Log: PS-04-0441 B. K. Singal Bechtel Power, Inc.

5275 Westview Dr.

Location BP4-1D2 Frederick, MD 21703-8306

Dear Mr. Singal:

Enclosed are data files containing meteorological data needed to revise the X/Q calculations for Farley Nuclear Plant RER 03-018. Previous transmittals of meteorological data in support of RER 03-018 per letters PS-03-1012, PS-03-1264 and PS-04-0182 contain significant data acquisition errors and are to be replaced with the attached new data.

A compact disc containing data files of raw meteorological data suitable for generating ARCON96 input is attached. Weather data is for years 2000 through 2003. Each file provides a month's worth of hourly weather data for a given year. The data units of measure are in English: wind speed is in miles per hour, wind direction is in degrees (true north); and air temperature delta is in degrees Fahrenheit between eevation 35 ft.

and 150 f L a*fc*T hlem 4

-W hftA4c, " 3g S

J Ataf t:V Some data points are in error or questionable as noted in the attached data review summary. For questionable data occurring over only a two hour period or less, a linear interpolation should be used to address these data points. For questionable data occurring for more than a two hour period of time, the following approaches should be applied. DOEJ SM-03-018-002 evaluated the three largest time periods of questionable data and provides justifications for acceptable. All other questionable data should be addressed in ACRON96 as bad data.

S:\\Workgroups\\SNC Soulhern NuctesaCo pale\\TechSupp\\Eng\\Products'tRER REAFarley ProjectF-o3-o18 Control RooM HabitablitD Analyses\\Mel 3 Final Data Used - with some backup tower data\\PS-04-0441.doc

NL-07-0067, Enclosure 5, Page 14 of 88 Calculation Number: BM-03-00 8-001, Version 1 Sheet 2 of 2 Attachment A.

Mr. B. K. Singal File: 03-018 March 30, 2004 Log: PS-04-0441 Page 2 Please use this new input to revise Calculation BM-03-0018-001. DOEJ SM-03-018-002 is attached and should be included as input in the revised calculation. The revised calculation should have a cautionary note added to the results section. The cautionary note should address the potential inaccuracy of the elevated stack releases since some questionable weather data was used for the 150 ft elevation. Therefore, only the results of the ground level releases are to be used for design basis evaluations.

If you have any questions or comments, please contact Steve Berryhill at extension (205) 992-7356 or Scott Soper at extension (205) 992-7045.

Sincerely, Mark J. Ajluni Project Support Manager - Farley SFB/crs Attachments cc:

S. F. Berryhill (copy of letter only)

J. A. Wehrenberg (copy of letter only)

NL-07-0067, Enclosure 5, Page 15 of 88 Calculation No. BM-03-0018-001, Version 1 Sheet 1 of 2 Attachment B FarlMet Program Listing Program FarlMet This code processes Farley onsite met data and generates an ARCON96-ready met data set.

The raw data have been converted to free format using MS Words to get rid of characters (i.e., /, :) for easy processing.

Replace / with a space for all; Replace: 00 with 2 spaces for all.

Save as filename.txt to be the input for this code.

Developed by: Y. J. Lin Date: 3/01/2004 Character*1 FNAME DATA FNAME/' '7 Integer Stab, UWS, UWD Print*, 'Enter raw data file name (Unit 5)'

Open (5, File=Fname)

Print*, 'Enter output file name Unit 6)'

Open (6, File=Fname)

Read raw data in free fromat DO 601 = 1,366 DO 50J = 1, 24 Read(5,*,END=80) Mon,LDay,IYear,IHour,DelT,WD2,WD 1,WS2,WS I Stability classification is based on FNP-0-ENV-17.

If (DelT.LE.-1.74) Stab = 1 If (DelT.GT.-1.74.and.DelT.LE.-1.56) Stab = 2 If (DelT.GT.-1.56.and.DelT.LE.-1.38) Stab = 3 If (DelT.GT.-1.38.and.DelT.LE.-0.46) Stab = 4 If (DelT.GT.-O.46.and.De]T.LE.1.38) Stab = 5 If (DelT.GT.1.38.and.DelT.LE.3.6) Stab = 6 If (DelT.GT.3.6) Stab = 7 Identify missing DeIT If (DelT.eq.-9999) Stab = 99

NL-07-0067, Enclosure 5, Page 16 of 88 Calculation No. BM-03-0018-001, Version 1 Sheet 2 of 2 Attachment B Convert wind speed from mph to mr/s and mutiply by 10 as required by ARCON96.

1 mph = 0.447 m/s If (WS1.ne.-9999) then LWS = WS1

0.447

10 Endlf If (WS2.ne.-9999) then UWS = WS2

0.447

10 Endlf LWD =WD1 UWD =VWD2 ARCON906 doesn't accept WD = 0 If (LWD.eq.0) LWD = 360 If (LWD.eq.0)

UWD = 360 Identify missing data using ARCON96 criteria If (WS 1.eq.-9999) LWS = 9999 If (WS2.eq.-9999) UTWS = 9999 If (LWD.eq.-9999) LWD = 999 If (UWD.eq.-9999) UWD = 999 Generate ARCON96-ready met data Write (6,9) I,lHourLWD,LWS,Stab,UWD,UWS 9 Format (lx,F'LMet',3x,13,I2,2x,13,14,lx,I2,2x,I3,I4) 50 Continue 60 Continue 80 END

00 cO 4-0 r-Calculation No.: BM-03-0018-001, Version 3 Attachment C sheet I of 36 a) 0)

co a-in W0 Co 0

I-a Unit 1 CR X/Q Estimates for Unit 1 Vent Release (4 years of meteorological data)

Program

Title:

ARCON96.

Developed For:

Date:

NRC Contacts:

U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation Division of Reactor Program Management June 25, 1997 11:00 a.m.

J.

Y. Lee Phone: (301) 415 1080 e-mail: jyll@nrc.gov J.

J. Hayes Phone: (301) 415 3167 e-mail: jjh@nrc.gov L.

A Brown Phone:

(301) 415 1232 e-mail: lab2onrc.gov Code Developer: J.

V. Ramsdell Phone: (509) 372 6316 e-mail: J ramsdell@pnl.gov Code Documentation:

NUREG/CR-6331 Rev. 1 The program was prepared for an agency of the United States Government. Neither the United States Government nor any agency thereof, nor any of their employees, makes any warranty, expressed or implied, or assumes any legal liability or responsibilities for any third party's use, or the results of such use, of any portion of this program or represents that its use by such third party would not infringe privately owned rights.

Program Run 9/ 4/2004 at 13:51:57 ARCON INPUT *****

Number of Meteorological Data Files Meteorological Data File Names Farleyco.met FarleyOl.met Farley02.met Farley03.met

=

4 Height of lower wind instrument (m) 10.0 Height of upper wind instrument (m)

=

45.7 Wind speeds entered as meters/second Ground-level release Release height (m)

=

44.3 Building Area (m^2)

=

.0 Effluent vertical velocity (m/s)

=

.00 Vent or stack flow (m^3/s)

=

.00 Vent or stack radius (m)

=

.00 Direction.. intake to source (deg)

=

115 Wind direction sector width (deg) 90 Wind direction window (deg)

=

070 -

160 Distance to intake (m)

=

60.6 Intake height (m)

=

11.4 Terrain elevation difference (m)

=

.0

00 00 4--

0) 00 (0

U C:

z Calculation No.: BM-03-0018-001, Version 3 Attachment C Sheet 2 of 36 Unit 1 CR X/Q Estimates for Unit 1 Vent Release (4 years of meteorological data)

Output file names CIV145.out ClV145.jfd Minimum Wind Speed (m/s)

.5 Surface roughness length (m)

.20 Sector averaging constant

=

4.3 Initial value of sigma y

.00 Initial value of sigma z

=

.00 Expanded output for code testing not selected Total number of hours of data processed =

35064 Hours of missing data

=

406 Hours direction in window

=

7414 Hours elevated plume w/ dir. in window

=

0 Hours of calm winds

=

2112 Hours direction not in window or calm 25132 DISTRIBUTION

SUMMARY

DATA BY AVERAGING INTERVAL AVER.

PER.

1 2

4 8

UPPER LIM.

1.00E-02 1.OOE-02 1.00E-02 1.00E-02 LOW LIM.

1.OOE-06 1.00E-06 1.OOE-06 1.00E-06 ABOVE RANGE

0.

IN RANGE 9526.

11788.

14876.

18569.

BELOW RANGE

0.

ZERO 25132.

22791.

19594.

15723.

TOTAL X/Qs 34658.

34579.

34470.

34292.

% NON ZERO 27.49 34.09 43.16 54.15 95th PERCENTILE X/O VALUES 1.61E-03 1.52E-03 1.43E-03 1.35E-03 95% X/Q for standard averaging intervals 0 to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 1.61E-03 2 to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 1.26E-03 8 to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 5.65E-04 1 to 4 days 3.43E-04 4 to 30 days 2.32E-04 12 1.00E-02 1.00E-06 0.

20974.

0.

13392.

34366.

61.03 24 1.00E-02 1.00E-06 0.

25403.

0.

8786.

34189, 74.30 96 1.00E-02 1.00E-06 0.

33288.

0.

810.

34098.

97.62 168 1.00E-02 1.00E-06 0.

33636.

0.

189.

33825.

99.44 360 1.00-E02 1.OOE-06 0.

33657.

0.

33657.

100.00 720 1.OOE-02 1.OOE-06 0.

34289.

0.

34289.

100.00 1.11E-03 8.26E-04 4.64E-04 3.90E-04 3.15E-04 2.63E-04 HOURLY VALUE RANGE MAX X/Q 3.07E-03 1.79E-03 CENTERLINE SECTOR-AVERAGE NORMAL PROGRAM COMPLETION MIN X/1 2.31E-04 1.34E-04

00 00 4-0 0Y)

(1)

L6 0~

03 C:)

0:

LU Calculation No.: BM-03-0018-001, Version 3 Attachment C Sheet 3 of 36 Unit 1 CR X/Q Estimates for Unit 1 Reactor Release (4 years of meteorological data)

Program

Title:

ARCON96.

Developed For:

Date:

NRC Contacts:

U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation Division of Reactor Program Management June 25, 1997 11:00 a.m.

J.

Y. Lee Phone:

(301) 415 1080 e-mail: jyllonrc.gov J.

J.

Hayes Phone:

(301) 415 3167 e-mail: jJh@nrc.gov L.

A Brown Phone:

(301) 415 1232 e-mail: lab2@nrc.gov Code Developer: J. V. Ramsdell Phone: (509) 372 6316 e-mail: j ramsdellopnl.gov Code Documentation:

NUREG/CR-6331 Rev.

1 The program was prepared for an agency of the United States Government. Neither the United States Government nor any agency thereof, nor any of their employees, makes any warranty, expressed or implied, or assumes any legal liability or responsibilities for any third party's use, or the results of such use, of any portion of this program or represents that its use by such third party would not infringe privately owned rights.

Program Run 9/

4/2004 at 13:51:58

-*,*,* ARCON INPUT ********

Number of Meteorological Data Files

=

Meteorological Data File Names FarleyO0.met Farley0l.met FarleyO2.met Farley03.met Height of lower wind instrument (m)

Height of upper wind instrument (m)

Wind speeds entered as meters/second Ground-level release Release height (m)

Building Area (m^2)

=

Effluent vertical velocity (m/s)

=

Vent or stack flow (m^3/a)

Vent or stack radius (m) 4 10.0 45.7 11.4 887.0

.00

.00 Direction..

intake to source (deg)

Wind direction sector width (deg)

Wind direction window (deg)

Distance to intake (m)

Intake height (m)

Terrain elevation difference (m)

=

136

=

90

=

091 -

181

=

39.2 11.4

.0

00 4-000 CN C4 0_

c-,

W 0)

M a-o

_J CD z

Calculation No.: BM-03-0018-001, Version 3 Attachment C Sheet 4 of 36 Unit 1 CR X/Q Estimates for Unit 1 Reactor Release (4 years of meteorological data)

Output file names ClR145.out CIR145.jfd Minimum Wind Speed (m/s)

=

.5 Surface roughness length (m)

.20 Sector averaging constant

=

4.3 Initial value of sigma y 6.97 Initial value of sigma z 3.54 Expanded output for code testing not selected Total number of hours of data processed =

35064 Hours of missing data

=

406 Hours direction in window

=

4777 Hours elevated plume w/ dir. in window

=

0 Hours of calm winds 1676 Hours direction not in window or calm

=

29205 DISTRIBUTION

SUMMARY

DATA BY AVERAGING INTERVAL AVER.

PER.

1 2

4 8

UPPER LIM.

1.OOE-02 1.00E-02 1.00E-02 1.00E-02 LOW LIM.

1.00E-06 1.00E-06 1.00E-06 1.00E-06 ABOVE RANGE

0.

IN RANGE 6453.

8552.

11610.

15661.

BELOW RANGE

0.

ZERO 28205.

26027.

22860.

18631.

TOTAL X/Qs 34658.

34579.

34470.

34292.

% NON ZERO 18.62 24.73 33.68 45.67 95th PERCENTILE X/Q VALUES 1.51E-03 1.39E-03 1.20E-03 1.05E-03 95% X/Q for standard averaging intervals 0 to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 1.51E-03 2 to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 9.01E-04 8 to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 3.95E-04 1 to 4 days 2.75E-04 4 to 30 days 1.91E-04 12 1.00E-02 1.00E-06 0.

18378.

0.

15988.

34366.

53.48 24 1.00E-02 1.00E-06 0.

23240.

0.

10949.

34189.

67.98 96 1.00E-02 1.OOE-06 0.

32492.

0.

1606.

34098.

95.29 168 1.00E-02 1.00E-06 0.

33563.

0.

262.

33825.

99.23 360 1.00-E02 1.00E-06 0.

33657.

0.

33657.

100.00 720 1.00E-02 1.OOE-06 0.

34289.

0.

34289.

100.00 8.56E-04 6.158-04 3.60E-04 3.01E-04 2.53E-04 2.14E-04 HOURLY VALUE RANGE MAX X/Q 2.01E-03 1.17E-03 CENTERLINE SECTOR-AVERAGE NORMAL PROGRAM COMPLETION MIN X/Q 4.32E-04 2.52E-04

0O 4-0 Calculation No.: BM-03-0018-001, Version 3 Attachment C Sheet 5 of 36 Unit 1 CR X/Q Estimates for Unit 1 Hatch Door Release (4 years of meteorological data)

L6 Program

Title:

ARCON96.

Developed For:

U.S. Nuclear Regulatory Commission U)

Office of Nuclear Reactor Regulation O

Division of Reactor Program Management 0

Date:

June 25, 1997 11:00 a.m.

w r-NRC Contacts:

J. Y. Lee Phone: (301) 415 1080 (0D e-mail: jyllanrc.gov D

J.

J. Hayes Phone:

(301) 415 3167 CO e-mail: jjh@nrc.gov o1L L. A Brown Phone:

(301) 415 1232 O

e-mail: lab2@nrc.gov Z

Code Developer: J.

V. Ramsdell Phone: (509) 372 6316 e-mail: jramadellopnl.gov Code Documentation:

NUREG/CR-6331 Rev.

1 The program was prepared for an agency of the United States Government. Neither the United States Government nor any agency thereof, nor any of their employees, makes any warranty, expressed or implied, or assumes any legal liability or responsibilities for any third party's use, or the results of such use, of any portion of this program Or represents that its use by such third party would not infringe privately owned rights.

Program Run 9/ 4/2004 at 13:51:59

.***ARCON INPUT

Number of Meteorological Data Files 4

Meteorological Data File Names FarleyOO.met Farley0l.met Farley02.met Farley03.met Height of lower wind instrument (i)

=

10.0 Height of upper wind instrument (i) =

45.7 Wind speeds entered as meters/second Ground-level release Release height Wm) 2.7 Building Area (m'2)

=

1253.0 Effluent vertical velocity (m/s)

=

.00 Vent or stack flow (m^3/s)

=

.00 Vent or stack radius Wm)

=

.00 Direction.. intake to source (deg)

=

136 Wind direction sector width (deg)

=

90 Wind direction window (deg) 091 - 181 Distance to intake (W) 80.4 Intake height (i) 11.4 Terrain elevation difference Wm)

.0

0o

'4-0 04 L6 0

Cw C)

I'-.

C)

Calculation No.: BM-03-0018-001, Version 3 Attachment C Sheet 6 of 36 Unit 1 CR X/Q Estimates for Unit I Hatch Door Release (4 years of meteorological data)

Output file names CiH145.out C1H145.jfd Minimum Wind Speed (m/s)

Surface roughness length (m)

Sector averaging constant Initial value of sigma y Initial value of sigma z

.5

.20 4.3

.00

.00 Expanded output for code testing not selected Total number of hours of data processed =

35064 Hours of missing data 406 Hours direction in window

=

4750 Hours elevated plume w/ dir. in window

=

0 Hours of calm winds 1874 Hours direction not in window or calm

=

28034 DISTRIBUTION

SUMMARY

DATA BY AVERAGING INTERVAL AVER.

PER.

1 2

4 8

UPPER LIM.

1.00E-02 1.00E-02 1.OOE-02 1.OOE-02 LOW LIM.

1.00E-06 1.00E-06 1.00E-06 1.00E-06 ABOVE RANGE

0.

IN RANGE 6624.

8784.

11910.

15991.

BELOW RANGE

0.

ZERO 28034.

25795.

22560.

18301.

TOTAL X/Qs 34658.

34579.

34470.

34292.

% NON ZERO 19.11 25.40 34.55 46.63 95th PERCENTILE X/Q VALUES 8.39E-04 7.92E-04 6.92E-04 5.93E-04 12 1.00E-02 1.00E-06 0.

18725.

0.

15641.

34366.

54.49 24 1.OOE-02 1.00E-06 0.

23530.

0.

10659.

34189.

68.82 96 1.OOE-02 1.00E-06 0.

32530.

0.

1568.

34098.

95.40 168

1. OOE-02 1.00E-06 0.

33572.

0.

253.

33825.

99.25 360 1.00E-02 1.008-06 0.

33657.

0.

33657.

100.00 720 1.00E-02 1.00E-06 0.

34289.

0.

34289.

100.00 4.72E-04 3.37E-04 1.94E-04 1.60E-04 1.29E-04 1.07E-04 95% X/Q for standard averaging intervals 0 to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 8,39E-04 2 to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 5.11E-04 8 to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 2.09E-04 1 to 4 days 1.47E-04 4 to 30 days 9.35E-05 HOURLY VALUE RANGE MAX X/Q 1.22E-03 7.11E-04 CENTERLINE SECTOR-AVERAGE NORMAL PROGRAM COMPLETION MIN X/Q 2.46E-04 1.43E-04

cO cO I+--

0 (NJ Calculation No.: BM-03-O018-O01, Version 3 Attachment C Sheet 7 of 36 0)

Unit 1 CR X/Q Estimates for Unit 2 Vent Release (4 years of meteorological data) 0_

(o Program

Title:

ARCON96.

Developed For:

U.S. Nuclear Regulatory Commission 0

Office of Nuclear Reactor Regulation Dae:Division of Reactor Program Management Date:

June 25, 1997 11:00 a.m.

CO NRC Contacts:

J.

Y. Lee Phone: (301) 415 1080 0

e-mail: jyll@nrc.gov 0

J. J. Hayes Phone:

(301) 415 3167 a

e-mail: jjh@nrc.gov L. A Brown Phone:

(301) 415 1232 z

e-mail: lab2@nrc.gov Code Developer: J. V. Ramsdell Phone: (509) 372 6316 e-mail: jramsdell@pnl.gov Code Documentation:

NUREG/CR-6331 Rev. 1 The program was prepared for an agency of the United States Government. Neither the United States Government nor any agency thereof, nor any of their employees, makes any warranty, expressed or implied, or assumes any legal liability or responsibilities for any third party's use, or the results of such use, of any portion of this program or represents that its use by such third party would not infringe privately owned rights.

Program Run 9/

4/2004 at 13:51:58 ARCON INPUT Number of Meteorological Data Files

=

4 Meteorological Data File Names FarleyOO.met Farley0l.met Farley02.met FarleyO3.met Height of lower wind instrument (m) 10.0 Height of upper wind instrument (m)

=

45.7 Wind speeds entered as meters/second Ground-level release Release height (m)

=

44.3 Building Area (m-2)

=

.0 Effluent vertical velocity (m/s)

=

.00 Vent or stack flow (mC3/s)

.00 Vent or stack radius (m)

=

.00 Direction.. intake to source (deg)

=

066 Wind direction sector width (deg)

=

90 Wind direction window (deg)

=

021 - 111 Distance to intake (m)

=

61.1 Intake height (m) 11.4 Terrain elevation difference (m)

=

.0

00 4o-0 CN4 ai) 0Y)

Calculation No.: BM-03-0018-001, Version 3 Attachment C Sheet 8 of 36 Unit 1 CR X/Q Estimates for Unit 2 Vent Release (4 years of meteorological data)

U) 0 C

ý.'

Co z

Output file names CIV245.out CiV245.jfd Minimum Wind Speed (m/s)

=

.5 Surface roughness length (m)

=

.20 Sector averaging constant

=

4.3 Initial value of sigma y

.00 Initial value of sigma z

.00 Expanded output for code testing not selected Total number of hours of data processed =

35064 Hours of missing data

=

406 Hours direction in window a

9813 Hours elevated plume w/ dir. in window a

0 Hours of calm winds

=

2112 Hours direction not in window or calm

=

22733 DISTRIBUTION

SUMMARY

DATA BY AVERAGING INTERVAL AVER.

PER.

1 2

4 8

UPPER LIM.

1.00E-02 1.00E-02 1.00E-02 1.00E-02 LOW LIM.

1.00E-06 1.OOE-06 1.00E-06 1.00E-06 ABOVE RANGE

0.

0, 0.

IN RANGE 11925.

14190, 17121.

20639.

BELOW RANGE

0.

ZERO 22733.

20389.

17349.

13653.

TOTAL X/Qs 34658.

34579.

34470.

34292.

% NON ZERO 34.41 41.04 49.67 60.19 95th PERCENTILE X/Q VALUES 1.64E-03 1.58E-03 1.52E-03 1.44E-03 95% X/O for standard averaging intervals 0 to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 1.64E-03 2 to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 1.37E-03 8 to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 7.17E-04 I to 4 days 5.41E-04 4 to 30 days 3.60E-04 12 1.00E-02

1. OOE-06 0.

22964.

0.

11402.

34366.

66.82 24 1.00E-02 1.00E-06 0.

27082.

0.

7107.

34189.

79.21 96 l.OOE-02 1.00E-06 0.

33570.

0.

528.

34098.

98,45 168 1.00E-02 1.00E-06 0.

33746.

0.

79.

33825.

99.77 360 1.OOE-02 1.00E-06 0.

33657.

0.

33657.

100.00 720 1.00E-02 1.00E-06 0.

34289.

0.

34289.

100.00 1.20E-03 9.58E-04 6.45E-04 5.31E-04 4.51E-04 3.98E-04 HOURLY VALUE RANGE MAX X/O MIN X/O 3.17E-03 1.54E-04 1.85E-03 9.00E-05 CENTERLINE SECTOR-AVERAGE NORMAL PROGRAM COMPLETION

00 4--

C14 Calculation No.: BM-03-0018-001, Version 3 Attachment C Sheet 9 of 36 Unit 1 CR X/Q Estimates for Unit 2 Reactor Release (4 years of meteorological data)

Q.

Program

Title:

ARCON96.

11)

Q)

Developed For:

U.S. Nuclear Regulatory Commission O

Office of Nuclear Reactor Regulation

0)

Division of Reactor Program Management Cw Date*

June 25, 1997 11:00 a.m.

NRC Contacts:

J.

Y. Lee Phone:

(301) 415 1080 0

(:

e-mail: jyll@nrc.gov J.

J. Hayes Phone:

(301) 415 3167 o*

e-mail: jjh@nrc.gov L. A Brown Phone: (301) 415 1232 z

e-mail: lab20nrc.gov Code Developer: J.

V. Ramsdell Phone: (509) 372 6316 e-mail: j_ramsdell@pnl.gov Code Documentation:

NUREG/CR-6331 Rev.

1 The program was prepared for an agency of the United states Government. Neither the United States Government nor any agency thereof, nor any of their employees, makes any warranty, expressed or implied, or assumes any legal liability or responsibilities for any third party's use, or the results of such use, of any portion of this program or represents that its use by such third party would not infringe privately owned rights.

Program Run 9/

4/2004 at 13:51:59 ARCON INPUT *********

Number of Meteorological Data Files

=

4 Meteorological Data File Names FarleyO0.met Farley0l.met FarleyO2.met FarleyO3.met Height of lower wind instrument (m) 10.0 Height of upper wind instrument (m) 45.7 Wind speeds entered as meters/second Ground-level release Release height (m) 11.4 Building Area (m^2) 887.0 Effluent vertical velocity (m/s)

=

.00 Vent or stack flow (m^3/s)

=

.00 Vent or stack radius (m)

=

.00 Direction..

intake to source (deg)

=

042 Wind direction sector width (deg)

=

90 Wind direction window (deg)

=

357 -

087 Distance to intake (m)

=

39.2 Intake height (m)

=

11.4 Terrain elevation difference (m)

=

.0

OD 0

C40 (1) 0)

L6 03 0

IC CD C:

C:

z Calculation No.: BM-03-0018-001, Version 3 Attachment C Sheet 10 of 36 Unit 1 CR X/Q Estimates for Unit 2 Reactor Release (4 years of meteorological data)

Output file names CIR245.out CiR245.jfd Minimum Wind Speed (m/s)

Surface roughness length (m)

Sector averaging constant Initial value of sigma y Initial value of sigma z

.5

.20 4.3 6.97 3.54 Expanded output for code testing not selected Total number of hours of data processed =

35064 Hours of missing data

=

406 Hours direction in window 10539 Hours elevated plume w/ dir. in window

=

0 Hours of calm winds

=

1676 Hours direction not in window or calm

=

22443 DISTRIBUTION

SUMMARY

DATA BY AVERAGING INTERVAL AVER.

PER.

1 2

4 8

UPPER LIM.

1.00E-02 1.00E-02 1.00E-02 1.005-02 LOW LIM.

1.00E-06 1.00E-06 1.00E-06 I.O0E-06 ABOVE RANGE

0.

IN RANGE 12215.

15134.

18672.

22694.

BELOW RANGE

0.

ZERO 22443.

19445.

15798.

11598.

TOTAL X/Qs 34658.

34579.

34470.

34292.

% NON ZERO 35.24 43.77 54.17 66.18 95th PERCENTILE X/Q VALUES 1.66E-03 1.62E-03 1.55E-03 1.44E-03 95% X/Q for standard averaging intervals 0 to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 1.66E-03 2 to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 1.36E-03 8 to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 6.81E-04 1 to 4 days 5.60E-04 4 to 30 days 4.21E-04 12 1.00E-02

1. OOE-06 0.

25261.

0.

9105.

34366.

73.51 24 1.00E-02 1.00E-06 0.

29151.

0.

5038.

34189.

85.26 96 1.00E-02 1.00E-06 0.

33693.

0.

405.

34098.

98.81 168 1.00E-02 1.00E-06 0.

33754.

0.

71.

33825.

99.79 360 1.00E-02 1.00E-06 0.

33657.

0.

33657.

100.00 720 1.00E-02

1. O0E-06 0.

34289.

0.

34289.

100.00 1.20E-03 9.33E-04 6.53E-04 5.62E-04 4.82E-04 4.52E-04 HOURLY VALUE RANGE MAX X/Q 2.01E-03 1.17E-03 CENTERLINE SECTOR-AVERAGE NORMAL PROGRAM COMPLETION MIN X/Q 3.39E-04 1.98E-04

4-0 04 L6 C)

C:)

0ý

.5I z

Calculation No.: BM-03-0018-001, Version 3 Attachment C Sheet 11 of 36 Unit 1 CR X/Q Estimates for Unit 2 Hatch Door Release (4 years of meteorological data)

Program

Title:

ARCON96.

Developed For:

Date:

NRC Contacts:

Code Developer:

U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation Division of Reactor Program Management June 25, 1997 11:00 a.m.

J.

Y. Lee Phone:

(301) 415 1080 e-mail: jyll@nrc.gov J.

J. Hayes Phone:

(301) 415 3167 e-mail: jjh@nrc.gov L. A Brown Phone:

(301) 415 1232 e-mail: lab2@nrc.gov J.

V. Ramsdell Phone: (509) 372 6316 e-mail: jramsdellopnl.gov Code Documentation:

NUREG/CR-6331 Rev.

1 The program was prepared for an agency of the United States Government. Neither the United States Government nor any agency thereof, nor any of their employees, makes any warranty, expressed or implied, or assumes any legal liability or responsibilities for any third party's use, or the results of such use, of any portion of this program or represents that its use by such third party would not infringe privately owned rights.

Program Run 9/

4/2004 at 13:51:59 ARCON INPUT Number of Meteorological Data Files

=

4 Meteorological Data File Names FarleyOO.met Farley0l.met Farley02.met Farley03.met Height of lower wind instrument (m)

Height of upper wind instrument (m) wind speeds entered as meters/second Ground-level release Release height (m)

Building Area (m^2)

Effluent vertical velocity (m/s)

Vent or stack flow (m^3/s)

Vent or stack radius (m)

Direction.. intake to source (deg)

Wind direction sector width (deg)

Wind direction window (deg)

Distance to intake Wm)

Intake height (m)

Terrain elevation difference (m) 10.0 45.7 2.7 1253.0

.00

.00 043 90 358 -

088 82.1 11.4

.0

00 4-0 03 a)

(L 1.6 CD 0

CuJ C:

CD z

Calculation No.: BM-03-0018-001, Version 3 Attachment C Sheet 12 of 36 Unit 1 CR X/Q Estimates for Unit 2 Hatch Door Release (4 years of meteorological data) output file names clH245.out C1H245.jfd Minimum Wind Speed (m/s)

=

.5 Surface roughness length (m)

=

.20 Sector averaging constant

=

4.3 Initial value of sigma y

.00 Initial value of sigma z

.00 Expanded output for code testing not selected Total number of hours of data processed =

35064 Hours of missing data

=

406 Hours direction in window

=

10367 Hours elevated plume w/ dir. in window

=

0 Hours of calm winds

=

1874 Hours direction not in window or calm

=

22417 DISTRIBUTION

SUMMARY

DATA BY AVERAGING INTERVAL AVER. PER.

1 2

4 8

UPPER LIM.

1.00E-03 1.00E-03 1.OOE-03 1.00E-03 LOW LIM.

1.00E-07 1.OOE-07 l.O0E-07 1.OOE-07 ABOVE RANGE

1.

0.

IN RANGE 12240.

15182.

18756.

22773.

BELOW RANGE

0.

ZERO 22417.

19397.

15714.

11519.

TOTAL X/QS 34658.

34579.

34470.

34292.

% NON ZERO 35.32 43.91 54.41 66.41 95th PERCENTILE X/Q VALUES 7.95E-04 7.72E-04 7.42E-04 7.04E-04 95% X/Q for standard averaging intervals 0 to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 7.95E-04 2 to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 6.73E-04 8 to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 3.35E-04 1 to 4 days 2.48E-04 4 to 30 days 1.87E-04 12 1.OOE-03 1.00E-07 0.

25333.

0.

9033.

34366.

73.72 24 1.00E-03 1.00E-07 0.

29237.

0.

4952.

34189.

85.52 96 1.00E-03 1.OOE-07 0.

33689.

147.

262.

34098.

99.23 168 1.OOE-03 1.00E-07 0.

33754.

9.

62.

33825.

99.82 360 1.00E-03 1.00E-07 0.

33657.

0.

33657.

100.00 720 1.00E-03 1.00E-07 0.

34289.

0.

34289.

100.00 5.88E-04 4.58E-04 3.01E-04 2.58E-04 2.16E-04 2.02E-04 HOURLY VALUE RANGE MAX X/Q 1.20E-03 6.98E-04 CENTERLINE SECTOR-AVERAGE NORMAL PROGRAM COMPLETION MIN X/Q 1.90E-04 1.11E-04

co

.4-0 03)

C14 Calculation No.: BM-03-0018-001, Version 3 Attachment C Sheet 13 of 36 Unit 2 CR X/Q Estimates for Unit 1 Vent Release (4 years of meteorological data) 0_

Program

Title:

ARCON96.

Developed For:

U.S. Nuclear Regulatory Commission (n

Office of Nuclear Reactor Regulation 0

~

Division of Reactor Program Management LU Date:

June 25, 1997 11:00 a.m.

NRC Contacts; J.

Y. Lee Phone:

(301) 415 1080 (0

O e-mail: jyll@nrc.gov O

J.

J. Hayes Phone: (301) 415 3167 rlý e-mail: jjh@nrc.gov O

L.

A Brown Phone:

(301) 415 1232 e-mail: lab2@nrc.gov z

Code Developer: J.

V. Ramsdell Phone: (509) 372 6316 e-mail: j ramsdell@pnl.gov Code Documentation:

NUREG/CR-6331 Rev. 1 The program was prepared for an agency of the United States Government. Neither the United States Government nor any agency thereof, nor any of their employees, makes any warranty, expressed or implied, or assumes any legal liability or responsibilities for any third party's use, or the results of such use, of any portion of this program or represents that its use by such third party would not infringe privately owned rights.

Program Run 9/ 4/2004 at 13:52:00

- - - ... ARCON INPUT ***..

Number of Meteorological Data Files 4

Meteorological Data File Names FarleyOO.met FarleyOl.met Farley02.met Farley03.met Height of lower wind instrument (m) 10.0 Height of upper wind instrument (m) 45.7 Wind speeds entered as meters/second Ground-level release Release height (m)

=

44.3 Building Area (mW2)

.0 Effluent vertical velocity (m/s)

=

.00 Vent or stack flow (mC3/s)

=

.00 Vent or stack radius (m)

.00 Direction.. intake to source (deg) 116 Wind direction sector width (deg) 90 Wind direction window (deg) 071 -

161 Distance to intake (m) 61.1 Intake height (m) 11.4 Terrain elevation difference (m)

.0

00 co 4-0 C) 09 Co

3 to U) 0)

(0 (00o 00 z*

Calculation No.: BM-03-0018-001, Version 3 Attachment C Sheet 14 of 36 Unit 2 CR X/Q Estimates for Unit I Vent Release (4 years of meteorological data)

Output file names C2V145.out C2V145.jfd Minimum Wind Speed (m/s)

=

.5 Surface roughness length (m)

=

.20 Sector averaging constant

=

4.3 Initial value of sigma y

.00 Initial value of sigma z

.00 Expanded output for code testing not selected Total number of hours of data processed =

35064 Hours of missing data

=

406 Hours direction in window

=

7329 Hours elevated plume w/ dir. in window

=

0 Hours of calm winds

=

2112 Hours direction not in window or calm

=

25217 DISTRIBUTION

SUMMARY

DATA BY AVERAGING INTERVAL AVER.

PER.

1 2

4 8

UPPER LIM.

1.00E-02 1.00E-02 1.00E-02 1.OOE-02 LOW LIM.

1.00E-06 1.00E-06 1.OOE-06 1.OOE-06 ABOVE RANGE

0.

IN RANGE 9441.

11688.

14770.

18449.

BELOW RANGE

0.

ZERO 25217.

22891.

19700.

15843.

TOTAL X/Os 34658.

34579.

34470.

34292.

% NON ZERO 27.24 33.80 42.8S 53.80 95th PERCENTILE X/Q VALUES 1.59E-03 1.50E-03 1.41E-03 1.33E-03 95% X/Q for standard averaging intervals 0 to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 1.59E-03 2 to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 1.25E-03 8 to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 5.54E-04 1 to 4 days 3.34E-04 4 to 30 days 2.27E-04 12 1.00E-02 1.00E-06 0.

20850.

0.

13516.

34366.

60.67 24 1.00E-02 1.OOE-06 0.

25276.

0.

8913.

34189.

73.93 96 1.00E-02 1.00E-06 0.

33185.

0.

913.

34098.

97.32 168

1. OOE-02 1.00E-06 0.

33616.

0.

209.

33825.

99.38 360 1.OOE-02 1.00E-06 0.

33657.

0.

33657.

100.00 720 1.00E-02 1.00E-06 0.

34289.

0.

34289.

100.00 1.09E-03 8.13E-04 4.54E-04 3.82E-04 3.08E-04 2.58E-04 HOURLY VALUE RANGE MAX X/Q 3.03E-03 1.77E-03 CENTERLINE SECTOR-AVERAGE NORMAL PROGRAM COMPLETION MIN X/O 2.28E-04 1.33E-04

00 4-0

%T-co Calculation No.: BM-03-0018-001 Version 3 Attachment C Sheet 15 of 36 03 Unit 2 CR X/Q Estimates for Unit 1 Reactor Release (4 years of meteorological data)

L6 Program

Title:

ARCON96.

Z Developed For: U.S. Nuclear Regulatory Commission co Office of Nuclear Reactor Regulation Division of Reactor Program Management C:

W Date:

June 25, 1997 11:00 a.m.

NRC Contacts:

J. Y. Lee Phone:

(301) 415 1080 o

e-mail; jyll@nrc.gov oC J. 0. Hayes Phone.(301) 415 3167 N-ý e-mail: jjhknrc.gov O

L. A Brown Phone:

(301) 415 1232

__1 e-mail: lab2@nrc.gov z

Code Developer; J. V. Ramsdell Phone:

(509) 372 6316 e-mail: jramsdell@pnl.gov Code Documentation:

NUREG/CR-6331 Rev.

1 The program was prepared for an agency of the United States Government. Neither the United States Government nor any agency thereof, nor any of their employees, makes any warranty, expressed or implied, or assumes any legal liability or responsibilities for any third party's use, or the results of such use, of any portion of this program or represents that its use by such third party would not infringe privately owned rights.

Program Run 9/ 4/2004 at 13:52:00 ARCON INPUT **********

Number of Meteorological Data Files 4

Meteorological Data File Names FarleyOO.met Farley0l.met Farley02.met Farley03.met Height of lower wind instrument (m)

=

10.0 Height of upper wind instrument (m)

=

45.7 Wind speeds entered as meters/second Ground-level release Release height (m)

=

11.4 Building Area (m^2)

=

887.0 Effluent vertical velocity (m/s)

=

.00 Vent or stack flow (m^3/s)

=

.00 Vent or stack radius (m)

=

.00 Direction.. intake to source (deg)

=

138 Wind direction sector width (deg)

=

90 Wind direction window (deg)

=

093 -

183 Distance to intake (m) 39.2 Intake height (m)

=

11.4 Terrain elevation difference (m)

=

.0

00 00 4-0)

L6)

C/)

0 C)

C:)

z Calculation No.: BM-03-0018-001, Version 3 Attachment C Sheet 16 of 36 Unit 2 CR X/Q Estimates for Unit 1 Reactor Release (4 years of meteorological data)

Output file names C2R145.out C2R145.jfd Minimum Wind Speed (m/s)

=

.5 Surface roughness length (m)

.20 Sector averaging constant

=

4.3 Initial value of sigma y 6.97 Initial value of sigma z 3.54 Expanded output for code testing not selected Total number of hours of data processed =

35064 Hours of missing data 406 Hours direction in window

=

4678 Hours elevated plume w/ dir. in window

=

0 Hours of calm winds

=

1676 Hours direction not in window or calm

=

28304 DISTRIBUTION

SUMMARY

DATA BY AVERAGING INTERVAL AVER.

PER.

1 2

4 8

UPPER LIM.

1.00E-02 1.00R-02 1.00E-02 l.0E-02 LOW LIM.

1.00E-06 1.00E-06 1.00E-06 1.00-E06 ABOVE RANGE

0.

IN RANGE 6354.

8443.

11487.

15524.

BELOW RANGE

0.

ZERO 28304.

26136.

22983.

18768.

TOTAL X/Qs 34658.

34579.

34470.

34292.

% NON ZERO 18.33 24.42 33.32 45.27 95th PERCENTILE X/Q VALUES 1.51E-03 1.39t-03 1.20E-03 1.05E-03 95% X/Q for standard averaging intervals 0 to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 1.51E-03 2 to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 8.91E-04 8 to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 3.91E-04 1 to 4 days 2.71E-04 4 to 30 days 1.87E-04 12 1.00E-02 1.OOE-06 0.

18224.

0.

16142.

34366.

53.03 24 1.00E-02 1.00E-06 0.

23053, 0.

11136.

34189.

67.43 96 1.00E-02 1.00E-06 0.

32357.

0.

1741.

34098.

94.89 168 1.00E-02 1.00E-06 0.

33535.

0.

290.

33825.

99.14 360 1.00E-02 1.00E-06 0.

33657.

0.

33657.

100.00 720 1.00E-02 1.00E-06 0.

34289.

0.

34289.

100.00 8.48E-04 6.09E-04 3.56E-04 2.96E-04 2.50E-04 2.09E-04 HOURLY VALUE RANGE MAX xlQ 2.01E-03 1.17E-03 CENTERLINE SECTOR-AVERAGE NORMAL PROGRAM COMPLETION MIN X/Q 4.32E-04 2.52E-04

cO 0O 4-0 Co) co Calculation No.: BM-03-0018-001, Version 3 Attachment C Sheet 17 of 36 MD Unit 2 CR X/Q Estimates for Unit 1 Hatch Door Release (4 years of meteorological data)

CL

(

Program

Title:

ARCON96.

Developed For:

U.S. Nuclear Regulatory Commission O

Office of Nuclear Reactor Regulation

-0 Division of Reactor Program Management C

W Date:

June 25, 1997 11:00 a.m.

(*

NRC Contacts:

J. Y. Lee Phone:

(301) 415 1080 O

e-mail: jyll@nrc.gov o

J. J. Hayes Phone: (301) 415 3167 Ie-mail:

jjh@nrc.gov O

L.

A Brown Phone:

(301) 415 1232 e-mail: lab2@nrc.gov z

Code Developer: J. V. Ramsdell Phone:

(509) 372 6316 e-mail: jramsdellopnl.gov Code Documentation:

NUREG/CR-6331 Rev.

1 The program was prepared for an agency of the United States Government. Neither the United States Government nor any agency thereof, nor any of their employees, makes any warranty, expressed or implied, or assumes any legal liability or responsibilities for any third party's use, or the results of such use, of any portion of this program or represents that its use by such third party would not infringe privately owned rights.

Program Run 9/ 4/2004 at 13:52:00

- - - - ARCON INPUT Number of Meteorological Data Files

=

4 Meteorological Data File Names FarleyOO.met FarleyOl.met Farley02.met FarleyO3.met Height of lower wind instrument Wm) 10.0 Height of upper wind instrument (i) 45.7 Wind speeds entered as meters/second Ground-level release Release height (W) 2.7 Building Area (m-2) 1253.0 Effluent vertical velocity (m/s)

=

.00 Vent or stack flow (m^3/s)

=

.00 Vent or stack radius (i)

=

.00 Direction.. intake to source (deg)

=

137 Wind direction sector width (deg)

=

90 wind direction window (deg)

=

092 -

192 Distance to intake (m)

=

82.1 Intake height (W)

=

11.4 Terrain elevation difference (W)

.0

00 00

'4-0 It CY) ca)

LU S)

(1) 0 o

C:)

rý-

C I:I K

Co

-Jz Calculation No.: BM-03-0018-001, Version 3 Attachment C Sheet 18 of 36 Unit 2 CR X/Q Estimates for Unit 1 Hatch Door Release (4 years of meteorological data)

Output file names C2H145.out C2Hl45.jfd Minimum Wind Speed (m/s)

=

.5 Surface roughness length (m)

=

.20 Sector averaging constant

=

4.3 Initial value of sigma y

.00 Initial value of sigma z

.00 Expanded output for code testing not selected Total number of hours of data processed =

35064 Hours of missing data

=

406 Hours direction in window

=

4702 Hours elevated plume w/ dir. in window

=

0 Hours of calm winds

=

1874 Hours direction not in window or calm

=

28082 DISTRIBUTION

SUMMARY

DATA BY AVERAGING INTERVAL AVER.

PER.

1 2

4 8

UPPER LIM.

1.OOE-02 1.00E-02 1.00E-02 1.00E-02 LOW LIM.

1.00E-06 1.00E-06 1.00E-06 1.00E-06 ABOVE RANGE

0.

IN RANGE 6576.

8733.

11854.

15932, BELOW RANGE

0.

ZERO 28082.

25846.

22616.

18360.

TOTAL X/Qs 34658.

34579.

34470.

34292.

% NON ZERO 18.97 25.26 34.39 46.46 95th PERCENTILE X/Q VALUES 8.04E-04 7.67E-04 6.65E-04 5.68E-04 95% X/Q for standard averaging intervals 0 to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 8.04E-04 2 to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 4.90E-04 8 to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 2.01E-04 1 to 4 days 1.39E-04 4 to 30 days 9.04E-05 12

1. 00E-03 1.00E-07 0.

18660.

0.

15706.

34366.

54.30 24 1.00E-03 1.00E-07 0.

23436.

0.

10753.

34189.

68.55 96 1.00E-03 1.00E-07 0.

32452.

0.

1646.

34098.

95.17 168 1.00E-03 1.00E-07 0.

33555.

0.

270.

33825.

99.20 360

1. 0OE-03 1.OGE-07 0.

33657.

0.

33657.

100.00 720 1.00E-03 1.00E-07 0.

34289.

0.

34289.

100.00 4.53E-04 3.23E-04 1.85E-04 1.54E-04 1.25E-04 1.03E-04 HOURLY VALUE RANGE MAX X/0 1.17E-03 6.84E-04 CENTERLINE SECTOR-AVERAGE NORMAL PROGRAM COMPLETION MIN X/0 2.36E-04 1.38E-04

00 00 4-0)

C0 0~

W C) 0 Calculation No.: BM-03-0018-001, Version 3 Attachment C Sheet 19 of 36 Unit 2 CR X/Q Estimates for Unit 2 Vent Release (4 years of meteorological data)

Program

Title:

ARCON96.

Developed For:

Date:

NRC Contacts:

Code Developer:

U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation Division of Reactor Program Management June 25, 1997 11:00 a.m.

J.

Y. Lee Phone: (301) 415 1080 e-mail: jyll@nrc.gov J.

J. Hayes Phone: (301) 415 3167 e-mail: jjh@nrc.gov L.

A Brown Phone:

(301) 415 1232 e-mail: lab2tnrc.gov J.

V. Ramsdell Phone:

(509) 372 6316 e-mail: j ramsdell@pnl.gov Code Documentation:

NUREG/CR-6331 Rev.

1 The program was prepared for an agency of the United States Government. Neither the United States Government nor any agency thereof, nor any of their employees, makes any warranty, expressed or implied, or assumes any legal liability or responsibilities for any third party's use, or the results of such use, of any portion of this program or represents that its use by such third party would not infringe privately owned rights.

Program Run 9/ 4/2004 at 13:52:00 ARCON INPUT Number of Meteorological Data Files

=

Meteorological Data File Names FarleyOO.met Farley0l.met Farley02.met Farley03.met Height of lower wind instrument (m)

=

Height of upper wind instrument (m)

Wind speeds entered as meters/second Ground-level release Release height (m)

Building Area (m2)

Effluent vertical velocity (m/s)

=

Vent or stack flow (m^3/s)

=

Vent or stack radius (m)

Direction..

intake to source (deg)

=

Wind direction sector width (deg)

=

Wind direction window (deg)

Distance to intake (m)

Intake height (m)

Terrain elevation difference (m)

=

4 10.0 45.7 44.3

.0

.00

.00 067 90 022 -

112 60.6 11.4

.0

4-0 (0

1a) 0)

(D 0~

rz..

C:

Calculation No.: BM-03-0018-001, Version 3 Attachment C Sheet 20 of 36 Unit 2 CR X/Q Estimates for Unit 2 Vent Release (4 years of meteorological data)

Output file names C2V245.out C2V245.jfd Minimum Wind Speed (m/a)

Surface roughness length (m)

Sector averaging constant Initial value of sigma y Initial value of sigma z

=

.5

=

.20

=

4.3

.00

.00 Expanded output for code testing not selected Total Hours Hours Hours Hours Hours number of hours of data processed =

35064 of missing data 406 direction in window

=

9788 elevated plume w/ dir. in window

=

0 of calm winds

=

2112 direction not in window or calm

=

22758 DISTRIBUTION

SUMMARY

DATA BY AVERAGING INTERVAL AVER.

PER.

UPPER LIM.

LOW LIM.

ABOVE RANGE IN RANGE BELOW RANGE ZERO TOTAL X/Qa V NON ZERO 1

1.00E-02 1.ODE-06 0.

11900.

0.

22758.

34658.

34.34 2

1.00E-02

1. 00E-06 0.

14156.

0.

20423.

34579.

40.94 4

1.00E-02 1.00E-06 0.

17092.

0.

17378.

34470.

49.59 8

1.00E-02 1.00E-06 0.

20589.

0.

13703.

34292.

60.04 12 1.OOE-02

1. OOE-06 0.

22917.

0.

11449.

34366.

66.69 24 1.00E-02 1.00E-06 0.

27034.

0.

7155.

34189.

79.07 96 1.00E-02 1.DOE-06 0.

33570.

0.

528.

34098.

98.45 168 1.00E-02

1. 00E-06 0.

33745.

0.

80.

33825.

99.76 360 1.00E-02 1.00E-06 0.

33657.

0.

33657.

100.00 720 1.OOE-02 1.OOE-06 0.

34289.

0.

34289.

100.00 95th PERCENTILE X/Q VALUES 1.65E-03 1.60E-03 1.53E-03 1.45E-03 95% X/O for standard averaging intervals 1.21E-03 9.63E-04 6.51E-04 5.36E-04 4.53E-04 4.01E-04 0

2 8

1 4

to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to 4 days to 30 days 1.65E-03 1.38E-03 7.20E-04 5.47E-04 3.63E-04 HOURLY VALUE RANGE MAX X/Q 3.21E-03 1.87E-03 CENTERLINE SECTOR-AVERAGE MIN x/O 1.56E-04 9.11E-05 NORMAL PROGRAM COMPLETION

0o 00 4-0 CY Calculation No.: BM-03-0018-001, Version 3 Attachment C Sheet 21 of 36 (W

Unit 2 CR X/Q Estimates for Unit 2 Reactor Release (4 years of meteorological data)

CL L6 Program

Title:

ARCON96.

Developed For:

U.S. Nuclear Regulatory Commission (U)

O Office of Nuclear Reactor Regulation Division of Reactor Program Management W

Date:

June 25, 1997 11:00 a.m.

NRC Contacts:

J.

Y. Lee Phone: (301) 415 1080 oC e-mail: jyll@nrc.gov o

J.

J Hayes Phone; (301) 415 3167 Ný e-mail: jjh@nrc.gov o

L.

A Brown Phone: (301) 415 1232 e-mail: lab2enrc.gov z

Code Developer: J.

V. Ramsdell Phone: (509) 372 6316 e-mail: J_ramadell'Spnl.gov Code Documentation:

NUREG/CR-6331 Rev.

1 The program was prepared for an agency of the United States Government. Neither the United States Government nor any agency thereof, nor any of their employees, makes any warranty, expressed or implied, or assumes any legal liability or responsibilities for any third party's use, or the results of such use, of any portion of this program or represents that its use by such third party would not infringe privately owned rights.

Program Run 9/

4/2004 at 13:52:01 ARCON INPUT Number of Meteorological Data Files

=

4 Meteorological Data File Names FarleyOO.met FarleyOl.met Farley02.met FarleyG3.met Height of lower wind instrument (m) 10.0 Height of upper wind instrument (m) 45.7 Wind speeds entered as meters/second Ground-level release Release height (m)

=

11.4 Building Area (m^2)

=

887.0 Effluent vertical velocity (m/s)

=

.00 Vent or stack flow (ma3/s)

=

.00 Vent or stack radius (m)

=

.00 Direction..

intake to source (deg)

=

045 Wind direction sector width (deg)

=

90 Wind direction window (deg)

=

000 -

090 Distance to intake (m)

=

39.2 Intake height (m)

=

11.4 Terrain elevation difference (m)

=

.0

0o

.4-0 co (l_

L6 0~

(0 C:

CD z

Calculation No.: BM-03-0018-001, Version 3 Attachment C Sheet 22 of 36 Unit 2 CR X/Q Estimates for Unit 2 Reactor Release (4 years of meteorological data)

Output file names C2R245.out C2R245.jfd Minimum Wind Speed (m/s)

=

.5 Surface roughness length (m)

=

.20 Sector averaging constant

=

4.3 Initial value of sigma y 6.97 Initial value of sigma z 3.54 Expanded output for code testing not selected Total number of hours of data processed =

35064 Hours of missing data

=

406 Hours direction in window

=

10049 Hours elevated plume w/ dir. in window

=

0 Hours of calm winds

=

1676 Hours direction not in window or calm

=

22933 DISTRIBUTION

SUMMARY

DATA BY AVERAGING INTERVAL AVER.

PER.

1 2

4 8

UPPER LIM.

1.00E-02 1.OOE-02 1.OOE-02 1.00E-02 LOW LIM.

1.00E-06 1.00E-06 1.00E-06 1.00E-06 ABOVE RANGE

0.

IN RANGE 11725.

14606.

18167.

22238.

BELOW RANGE

0.

ZERO 22933.

19973.

16303.

12054.

TOTAL X/Qs 34658.

34579.

34470.

34292.

% NON ZERO 33.83 42.24 52.70 64.85 95th PERCENTILE X1Q VALUES 1.65E-03 1.61E-03 1.53E-03 1.42E-03 95% X/Q for standard averaging intervals 0 to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 1.65E-03 2 to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 1.34E-03 8 to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 6.75E-04 1 to 4 days 5.40E-04 4 to 30 days 4.03E-04 12 1.OOE-02 1.00E-06 0.

24827.

0.

9539.

34366.

72.24 24 1.00E-02 1.DOE-06 0.

28819.

0.

5370.

34189.

84.29 96 1.00E-02

1. ODE-06 0.

33660.

0.

438.

34098.

98.72 168 1.00E-02

1. 00E-06 0.

33754.

0.

71.

33825.

99.79 360 1.OOE-02 1.00E-06 0.

33657.

0.

33657.

100.00 720

1. 00E-02 1.00E-06 0.

34289.

0.

34289.

100.00 1.19E-03 9.24E-04 6.36E-04 5.49E-04 4.68E-04 4.34E-04 HOURLY VALUE RANGE MAX X/O 2.01E-03 1.17E-03 CENTERLINE SECTOR-AVERAGE NORMAL PROGRAM COMPLETION MIN X/Q 3.39E-04 1.98E-04

00 4-0 0)

Calculation No.: BM-03-0018-001, Version 3 Attachment C Sheet 23 of 36 0)

Unit 2 CR X/Q Estimates for Unit 2 Hatch Door Release (4 years of meteorological data)

L6o) Program

Title:

ARCON96.

(w Developed For:

U.S. Nuclear Regulatory Commission O

Office of Nuclear Reactor Regulation C.

Division of Reactor Program Management Wjj Date:

June 25, 1997 11:00 a.m.

(0 NRC Contacts:

J.

Y. Lee Phone:

(301) 415 1080 S

e-mail: jyllcnrc.gov C

J.

J. Hayes Phone: (301) 415 3167 C:

e-mail: jjh@nrc.gov O1 L. A Brown Phone: (301) 415 1232 e-mail; lab2@nrc.gov z

Code Developer: J.

V. Ramsdell Phone: (509) 372 6316 e-mail: J_ramadell@pnl.gov Code Documentation:

NUREG/CR-6331 Rev. 1 The program was prepared for an agency of the United States Government. Neither the United States Government nor any agency thereof, nor any of their employees, makes any warranty, expressed or implied, or assumes any legal liability or responsibilities for any third party's use, or the results of such use, of any portion of this program or represents that its use by such third party would not infringe privately owned rights.

Program Run 9/ 4/2004 at 13:52:01

- - - ~**

ARCON INPUT **********

Number of Meteorological Data Files

=

4 Meteorological Data File Names FarleyOO.met FarleyOl.met Farley02.met Farley03.met Height of lower wind instrument (m) 10.0 Height of upper wind instrument (m) 45.7 Wind speeds entered as meters/second Ground-level release Release height (m) 2.7 Building Area (mW2)

=

1253.0 Effluent vertical velocity (m/s)

=

.00 Vent or stack flow (mA3/s)

=

.00 Vent or stack radius (m)

=

.00 Direction.. intake to source (deg)

=

045 Wind direction sector width (deg)

=

90 Wind direction window (deg)

=

000 -

090 Distance to intake (m)

=

80.4 Intake height (m)

=

11.4 Terrain elevation difference (m)

=

.0

00~

00 0

CD CL 0

0 CD z

Calculation No.: BM-03-0018-001, Version 3 Attachment C Sheet 24 of 36 Unit 2 CR X/Q Estimates for Unit 2 Hatch Door Release (4 years of meteorological data)

Output file names C2H245.out C2H245.jfd Minimum Wind Speed (m/s)

Surface roughness length (m)

Sector averaging constant Initial value of sigma y Initial value of sigma z

.5

.20 4.3

.00

.00 Expanded output for code testing not selected Total number of hours of data processed =

35064 Hours of missing data

=

406 Hours direction in window

=

9977 Hours elevated plume w/ dir. in window

=

0 Hours of calm winds

=

1874 Hours direction not in window or calm

=

22807 DISTRIBUTION

SUMMARY

DATA BY AVERAGING INTERVAL AVER.

PER.

1 2

4 8

UPPER LIM.

1.O0E-02 1.00E-02 1.00E-02 1.00E-02 LOW LIM.

1.OOE-06 1.OOE-06 1.00E-06 1.00E-06 ABOVE RANGE

0.

IN RANGE 11851.

14761.

18346.

22415.

BELOW RANGE

0.

ZERO 22807.

19818.

16124.

11877.

TOTAL X/Qs 34658.

34579.

34470.

34292.

! NON ZERO 34.19 42.69 53.22 65.37 95th PERCENTILE X/Q VALUES 8.33E-04 8.04E-04 7.72E-04 7.32E-04 95% X/Q for standard averaging intervals 0 to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 8.33E-04 2 to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 6.98E-04 8 to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 3.43E-04 1 to 4 days 2.57E-04 4 to 30 days 1.91E-04 12

1. OOE-03

1. OOE-07 0.

24996.

0.

9370.

34366.

72.73 24 1.00E-03 1.00E-07 0.

28978.

0.

5211.

34189.

84.76 96

1. OOE-03 1.00E-07 0.

33683.

0.

41S.

34098.

98.78 168

1. OOE-03 1.00E-07 0.

33754.

0.

71.

33825.

99.79 360 1.00E-03 1.00E-07 0.

33657.

0.

33657.

100.00 720 1.00E-03 1.00E-07 0.

34289.

0.

34289.

100.00 6.12E-04 4.73E-04 3.11E-04 2.65E-04 2.18E-04 2.07E-04 HOURLY VALUE RANGE MAX X/O 1.25E-03 7.26E-04 CENTERLINE SECTOR-AVERAGE NORMAL PROGRAM COMPLETION MIN X/Q 1.97E-04 1.15E-04

00 00 4--

0 Calculation No.: BM-03-0018-001, Version 3 Attachment C Sheet 25 of 36 WTSC X/Q Estimates for Unit 1 Vent Release (4 years of meteorological data) aO Q)D Program

Title:

ARCON96.

Developed For:

U.S. Nuclear Regulatory Commission U) o Office of Nuclear Reactor Regulation aD Division of Reactor Program Management C-WLJ Date:

June 25, 1997 11:00 a.m.

t-:

NRC Contacts:

J.

Y. Lee Phone:

(301) 415 1080 D

e-mail: jyll@nrc.gov o

J. J. Hayes Phone:

(301) 415 3167 s-.

e-mail: jjh@nrc.gov o

L. A Brown Phone:

(301) 415 1232 e-mail: lab2@nrc.gov z

Code Developer: J.

V. Ramsdell Phone:

(509) 372 6316 e-mail: j ramsdell@pnl.gov Code Documentation:

NUREG/CR-6331 Rev. 1 The program was prepared for an agency of the United States Government. Neither the United States Government nor any agency thereof, nor any of their employees, makes any warranty, expressed or implied, or assumes any legal liability or responsibilities for any third party's use, or the results of such use, of any portion of this program or represents that its use by such third party would not infringe privately owned rights.

Program Run 9/ 4/2004 at 13:52:01

- - - - ARCON INPUT Number of Meteorological Data Files

=

4 Meteorological Data File Names FarleyOO.met Farley01.met FarleyO2.met Farley03.met Height of lower wind instrument (m) 10.0 Height of upper wind instrument (m) 45.7 Wind speeds entered as meters/second Ground-level release Release height (m) 44.3 Building Area (m2)

.0 Effluent vertical velocity (m/s)

.00 Vent or stack flow (m^3/s)

=

.00 Vent or stack radius (m)

.00 Direction..

intake to source (deg)

=

141 Wind direction sector width (deg) 90 wind direction window (deg)

=

096 -

186 Distance to intake (m)

=

70.2 Intake height (m)

=

6.9 Terrain elevation difference (m)

=

.0

00 00

'-4 0

N L6 C) 03 C)

Cl w

cc 0

0

_Jz Calculation No.: BM-03-0018-001, Version 3 Attachment C Sheet 26 of 36 Unit 2 CR X/Q Estimates for Unit 1 Vent Release (4 years of meteorological data) output file names T2V145.out T2V145.jfd Minimum Wind Speed (m/s)

=

.5 Surface roughness length (m)

=

.20 Sector averaging constant

=

4.3 Initial value of sigma y

=

.00 Initial value of sigma z

=

.00 Expanded output for code testing not selected Total number of hours of data processed =

35064 Hours of missing data

=

406 Hours direction in window

=

5548 Hours elevated plume w/ dir. in window

=

0 Hours of calm winds

=

2112 Hours direction not in window or calm

=

26998 DISTRIBUTION

SUMMARY

DATA BY AVERAGING INTERVAL AVER.

PER.

1 2

4 8

12 24 96 168 360 720 UPPER LIM.

1.00E-02 1.OOE-02 1.00E-02 1.00E-02 1.00E-02 1.00E-02 1.OOE-02 1.OOE-02 1.00E-02 1.00E-02 LOW LIM.

1.00E-06 1.OOE-06 1.OOE-06 1.00Eý06 1.00E-06 l.OOE-06 1.00E-06 l.OOE-06 1.00E-06 1.00E-06 ABOVE RANGE

0.

IN RANGE 7660.

9772.

12807.

16615.

19132.

23670.

32215.

33399.

33657.

34289.

BELOW RANGE

0.

ZERO 26998.

24807.

21663.

17677.

15234.

10519.

1883.

426.

0.

TOTAL X/Qs 34658.

34579.

34470.

34292.

34366.

34189.

34098.

33825.

33657.

34289.

% NON ZERO 22.10 28.26 37.15 48.45 55.67 69.23 94.48 98.74 100.00 100.00 95th PERCENTILE X/Q VALUES 1.15S-03 1.08E-03 1.01E-03 9.01E-04 95% X/Q for standard averaging intervals 0 to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 1.15E-03 2 to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 8.19E-04 8 to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 3.44E-04 1 to 4 days 2.10E-04 4 to 30 days 1.49E-04 7.31E-04 5.30E-04 2.90E-04 2.34E-04 1.89E-04 1.68E-04 HOURLY VALUE RANGE MAX X/Q MIN X/Q 2.33E-03 2.39E-04 1.368-03 1.40E-04 CENTERLINE SECTOR-AVERAGE NORMAL PROGRAM COMPLETION

co co 4-0 Calculation No.: BM-03-0018-001, Version 3 Attachment C Sheet 27 of 36 (3)

CTSC X/Q Estimates for Unit 1 Reactor Release (4 years of meteorological data)

L6 P

Program

Title:

ARCON96.

=

Developed For:

U.S. Nuclear Regulatory Commission U9 O

Office of Nuclear Reactor Regulation Division of Reactor Program Management W

Date:

June 25, 1997 11:00 a.m.

(0 NRC Contacts:

J.

Y. Lee Phone: (301) 415 1080 O

e-mail: jyll@nrc.gov

(

J.

Hayes Phone: (301) 415 3167 11-e-mail: jjhenrc.gov CD L. A Brown Phone:

(301) 415 1232

-I e-mail: lab2@nrc.gov z

Code Developer: J.

V. Ramsdell Phone: (509) 372 6316 e-mail: j ramsdellopnl.gov Code Documentation:

NUREG/CR-6331 Rev.

1 The program was prepared for an agency of the United States Government. Neither the United States Government nor any agency thereof, nor any of their employees, makes any warranty, expressed or implied, or assumes any legal liability or responsibilities for any third party's use, or the results of such use, of any portion of this program or represents that its use by such third party would not infringe privately owned rights.

Program Run 9/

4/2004 at 13:52:02

- - - - t ARCON INPUT Number of Meteorological Data Files

=

4 Meteorological Data File Names FarleyOO.met Farley0l.met Farley02.met Farley03.met Height of lower wind instrument (m)

=

10.0 Height of upper wind instrument (m)

=

45.7 Wind speeds entered as meters/second Ground-level release Release height (m)

=

6.9 Building Area (m2) 1078.0 Effluent vertical velocity (m/s)

=

.00 Vent or stack flow (m^3/s)

.00 Vent or stack radius (m)

=

.00 Direction.. intake to source (deg)

=

157 Wind direction sector width (deg)

=

90 Wind direction window (deg)

=

112 - 202 Distance to intake (m)

=

57.6 Intake height (m)

=

6.9 Terrain elevation difference (m)

=

.0

00 0) 0)

0) 0-

"6 0) 0 C:

w CD z

Calculation No.: BM-03-0018-001, Version 3 Attachment C Sheet 28 of 36 Unit 2 CR X/Q Estimates for Unit 1 Reactor Release (4 years of meteorological data)

Output file names T2R145.out T2R145.jfd Minimum Wind Speed (M/S)

=

.5 Surface roughness length (m)

.20 Sector averaging constant

=

4.3 Initial value of sigma y 6.97 Initial value of sigma z 4.30 Expanded output for code testing not selected Total number of hours of data processed =

35064 Hours of missing data

=

406 Hours direction in window

=

3864 Hours elevated plume w/ dir. in window

=

0 Hours of calm winds

=

1874 Hours direction not in window or calm

=

28920 DISTRIBUTION

SUMMARY

DATA BY AVERAGING INTERVAL AVER.

PER.

1 2

4 8

UPPER LIM.

1.OOE-02 1.OOE-02 1.00E-02 1.OOE-02 LOW LIM.

1.OOE-06 1.OOE-06 1.00E-06 l.OOE-06 ABOVE RANGE

0.

IN RANGE 5738.

7804.

10804.

14714.

BELOW RANGE

0.

ZERO 28920.

26775.

23666.

19578.

TOTAL X/Qs 34658.

34579.

34470.

34292.

% NON ZERO 16.56 22.57 31.34 42.91 95th PERCENTILE X/Q VALUES 8.17E-04 7.46E-04 6.33E-04 5.52E-04 95% X/O for standard averaging intervals 0 to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 8.17E-04 2 to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 4.64E-04 8 to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 2.OOE-04 1 to 4 days 1.41E-04 4 to 30 days 9.26E-05 12

1. OOE-03 1.OOE-07 0.

17403.

0.

16963.

34366.

50.64 24 1.OOE-03

1. OOE-07 0.

22275.

0.

11914.

34189.

65.15 96 1,OOE-03 1.OOE-07 0.

31655.

0.

2443.

34098.

92.84 168 1.OOE-03

1. OOE-07 0.

33249.

0.

576.

33825.

98.30 360 1.00E-03 1.00E-07 0.

33657.

0.

33657.

100.00 720 1.00E-03 1.00E-07 0.

34289.

0.

34289.

100.00 4.45E-04 3.17E-04 1.85E-04 1.52E-04 1.27E-04 1.05E-04 HOURLY VALUE RANGE MAX X/Q 1.15E-03 6.73E-04 CENTERLINE SECTOR-AVERAGE NORMAL PROGRAM COMPLETION MIN X/Q 2.26E-04

1. 32E-04

00 4--

0)

Calculation No.: BM-03-0018-001, Version 3 Attachment C Sheet 29 of 36 TSC X/Q Estimates for Unit 1 Hatch Door Release (4 years of meteorological data) 0_

U5 Program

Title:

ARCON96.

Developed For:

U.S. Nuclear Regulatory Commission U)

O Office of Nuclear Reactor Regulation 76 Division of Reactor Program Management C:

W Date:

June 25, 1997 11:00 a.m.

NRC Contacts:

J.

Y. Lee Phone:

(301) 415 1080 o

e-mail: jyllonrc.gov O

J. J. Hayes Phone: (301) 415 3167 N-.

e-mail: jjhenrc.gov O

L. A Brown Phone:

(301) 415 1232

-1 e-mail: lab2@nrc.gov z

Code Developer: J.

V. Ramsdell Phone:

(509) 372 6316 e-mail: jramsdell@pnl.gov Code Documentation:

NUREG/CR-6331 Rev. 1 The program was prepared for an agency of the United States Government. Neither the United States Government nor any agency thereof, nor any of their employees, makes any warranty, expressed or implied, or assumes any legal liability or responsibilities for any third party's use, or the results of such use, of any portion of this program or represents that its use by such third party would not infringe privately owned rights.

Program Run 9/ 4/2004 at 13:52:02

....***, ARCON INPUT Number of Meteorological Data Files 4

Meteorological Data File Names Farley0O.met Farley0l.met Farley02.met FarleyO3.met Height of lower wind instrument Wm) 10.0 Height of upper wind instrument Wm) 45.7 Wind speeds entered as meters/second Ground-level release Release height (i)

=

2.7 Building Area (m^2)

=

1253.0 Effluent vertical velocity (m/s)

=

.00 Vent or stack flow (m'3/s)

=

.00 Vent or stack radius Wm)

=

.00 Direction.. intake to source (deg)

=

157 Wind direction sector width (deg) 90 Wind direction window (deg) 112 - 202 Distance to intake Wm) 100.7 Intake height (m) 6.9 Terrain elevation difference Wm)

.0

co co 00 4--

0)

(D 0)

Co CU o

o (0) 0 (0

N-0 C)z Calculation No.: BM-03-0018-001, version 3 Attachment C Sheet 30 of 36 TSC CR X/Q Estimates for Unit 1 Hatch Door Release (4 years of meteorological data)

Output file names T2H145.out T2H145.jfd Minimum Wind Speed (m/s)

Surface roughness length (m)

Sector averaging constant Initial value of sigma y Initial value of sigma z

=

.5

=

.20 f

4.3

.00

.00 Expanded output for code testing not selected Total number of hours of data processed =

35064 Hours of missing data 406 Hours direction in window 3864 Hours elevated plume w/ dir. in window

=

0 Hours of calm winds

=

1874 Hours direction not in window or calm

=

28920 DISTRIBUTION

SUMMARY

DATA BY AVERAGING INTERVAL AVER.

PER.

1 2

4 8

UPPER LIM.

1.00E-03 1.OOE-03 1.00E-03 l.OOE-03 LOW LIM.

1.00E-07 1.00E-07 1.00E-07 1.00E-07 ABOVE RANGE

0.

IN RANGE 5738.

7804.

10804.

14714.

BELOW RANGE

0.

ZERO 28920.

26775.

23666.

19578.

TOTAL X/Os 34658.

34579.

34470.

34292.

% NON ZERO 16.56 22.57 31.34 42.91 95th PERCENTILE X/Q VALUES 5.47E-04 5.05E-04 4.23E-04 3.58E-04 95% X/Q for standard averaging intervals 0 to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 5.47E-04 2 to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 2.95E-04 8 to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 1.27E-04 1 to 4 days 8.84E-05 4 to 30 days 5.44E-05 12

1. OOE-03 1.00E-07 0.

17403.

0.

16963.

34366.

50.64 24 1.00E-03 1.00E-07 0.

22275.

0.

11914.

34189.

65.15 96 1.00E-03 1.00E-07 0.

31655.

0.

2443.

34098.

92.84 168 1.00E-03

1. O0E-07 0.

33249.

0.

576, 33825.

98.30 360 1.OOE-03 1.00E-07 0,

33657.

0.

33657.

100.00 720 1.00E-03 1.00E-07 0.

34289.

0.

34289.

100.00 2.85E-04 2.04E-04 1.17E-04 9.54E-05 7.65E-05 6.28E-05 HOURLY VALUE RANGE MAX XIG 8.04E-04 4.68E-04 CENTERLINE SECTOR-AVERAGE NORMAL PROGRAM COMPLETION MIN X/l 1.46E-04

8. 53E-05

00 00 I-.

Calculation No.: BM-03-0018-001, Version 3 Attachment C Sheet 31 of 36 WTSC X/Q Estimates for Unit 2 Vent Release (4 years of meteorological data)

P program

Title:

ARCON96.

U)

Developed For:

U.S. Nuclear Regulatory Commission o

Office of Nuclear Reactor Regulation OD Division of Reactor Program Management LI]

Date:

June 25, 1997 11:00 a.m.

N.1 (0

NRC Contacts:

J. Y. Lee Phone: (301) 415 1080 o

e-mail: jyllwnrc.gov 0

J. J. Hayes Phone:

(301) 415 3167

.e-mail:

jih@nrc.gov 0

L. A Brown Phone:

(301) 415 1232 "1

e-mail: lab2@nrc.gov z

Code Developer: J. V. Ramsdell Phone:

(509) 372 6316 e-mail: jramsdell@pnl.gov Code Documentation:

NUREG/CR-6331 Rev.

1 The program was prepared for an agency of the United States Government. Neither the United States Government nor any agency thereof, nor any of their employees, makes any warranty, expressed or implied, or assumes any legal liability or responsibilities for any third party's use, or the results of such use, of any portion of this program or represents that its use by such third party would not infringe privately owned rights.

Program Run 9/

4/2004 at 13:52:02 ARCON INPUT Number of Meteorological Data Files

=

4 Meteorological Data File Names Farley00.met Farley0l.met FarleyD2.met Farley03.met Height of lower wind instrument (m) 10.0 Height of upper wind instrument (m) 45.7 Wind speeds entered as meters/second Ground-level release Release height (m)

=

44.3 Building Area (m^2)

=

.0 Effluent vertical velocity (mWs)

.00 Vent or stack flow (m^3/s)

=

.00 Vent or stack radius (m)

.00 Direction.. intake to source (deg)

=

096 Wind direction sector width (deg)

=

90 Wind direction window (deg)

=

051 -

141 Distance to intake (m) 46.8 Intake height (m)

=

6.9 Terrain elevation difference (m)

=

.0

c0 00 4-03 00 ao-U)

C:)

0:

Calculation No.: BM-03-0018-001, Version 3 Attachment C Sheet 32 of 36 TSC X/Q Estimates for Unit 2 Vent Release (4 years of meteorological data)

Output file names T2V245.out T2V245.jfd Minimum Wind Speed (m/s)

Surface roughness length (m)

Sector averaging constant Initial value of sigma y Initial value of sigma z

=

.5

=

.20

=

4.3

.00

.00 Expanded output for code testing not selected Total number of hours of data processed =

35064 Hours of missing data 406 Hours direction in window

=

8779 Hours elevated plume w/ dir. in window

=

0 Hours of calm winds 2112 Hours direction not in window or calm

=

23767 DISTRIBUTION

SUMMARY

DATA BY AVERAGING INTERVAL AVER.

PER.

UPPER LIM.

LOW LIM.

ABOVE RANGE IN RANGE BELOW RANGE ZERO TOTAL X/Qs

% NON ZERO 1

1.00E-02 1.OOE-06 0.

10891.

0.

23767.

34658.

31.42 2

1.00E-02 1.OOE-06 0.

13205.

0.

21374.

34579.

38.19 4

1. OOE-02 1.00E-06 0.

16272.

0.

18198.

34470.

47.21 8

1. OOE-02 1.00E-06 0.

19941.

0.

14351.

34292.

58.15 12 1.OOE-02 1.00E-06 0.

22308.

0.

12058.

34366.

64.91 24 1.00E-02 1.O0E-06 0.

26434.

0.

7755.

34189.

77.32 96 1.00E-02

1. OE-06 0.

33389.

0.

709.

34098.

97.92 168

1. OOE-02 1.OOE-06 0.

33722.

0.

103.

33825.

99.70 360

1. OOE-02 1.OOE-06 0.

33657.

0.

33657.

100.00 720 1.OOE-02 1.OOE-06 0.

34289.

0.

34289.

100.00 95th PERCENTILE X/Q VALUES 2.11E-03 2.02E-03 1.91E-03 1.81E-03 1.S5OE-03 1.15E-03 6.81E-04 S. 92E-04 4.66E-04 3.98E-04 95% x/Q for standard averaging intervals 0

2 8

1 4

to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to 4 days to 30 days 2.11E-03 1.72E-03 8.13E-04 5.25E-04 3.55E-04 HOURLY VALUE RANGE MAX X/Q 3.85E-03 2.24E-03 CENTERLINE SECTOR-AVERAGE MIN X/Q 2.99E-04 1.74E-04 NORMAL PROGRAM COMPLETION

cO cO 4--

0 0)

Calculation No.: BM-03-0018-001, Version 3 Attachment C Sheet 33 of 36 0)

Cu TSC X/Q Estimates for Unit 2 Reactor Release (4 years of meteorological data)

L6 Program

Title:

ARCON96.

Developed For:

U.S. Nuclear Regulatory Commission 0)

O Office of Nuclear Reactor Regulation Division of Reactor Program Management C-WU Date:

June 25, 1997 11:00 a.m.

NRC Contacts:

J. Y. Lee Phone:

(301) 415 1080 o

e-mail: jyllenrc.gov O

J. J. Hayes Phone:

(301) 415 3167 N'-

e-mail: jjh@nrc.gov OC L.

A Brown Phone: (301) 415 1232 e-mail: lab2@nrc.gov z

Code Developer: J.

V. Ramsdell Phone: (509) 372 6316 e-mail: j ramsdellepnl.gov Code Documentation:

NUREG/CR-6331 Rev. 1 The program was prepared for an agency of the United States Government. Neither the United States Government nor any agency thereof, nor any of their employees, makes any warranty, expressed or implied, or assumes any legal liability or responsibilities for any third party's use, or the results of such use, of any portion of this program or represents that its use by such third party would not infringe privately owned rights.

Program Run 9/

4/2004 at 13:52:03 ARCON INPUT Number of Meteorological Data Files

=

4 Meteorological Data File Names Farley00.met Farley0l.met FarleyO2.met FarleyO3.met Height of lower wind instrument (m) 10.0 Height of upper wind instrument (m) 45.7 Wind speeds entered as meters/second Ground-level release Release height (m) 6.9 Building Area (m'2)

=

1078.0 Effluent vertical velocity (m/s)

=

.00 Vent or stack flow Wm3/s)

.00 Vent or stack radius (m)

=

.00 Direction.. intake to source (deg)

=

066 Wind direction sector width (deg)

=

90 Wind direction window (deg) 021 -

111 Distance to intake (m)

=

13.6 Intake height (m) 6.9 Terrain elevation difference (m)

=

.0

c0 co

'4-0 LO co 0-75 Co 0-0 Co rl:

0 Calculation No.: BM-03-0018-001, Version 3 Attachment C Sheet 34 of 36 TSC X/Q Estimates for Unit 2 Reactor Release (4 years of meteorological data)

Output file names T2R245.out T2R245.jfd Minimum Wind speed (m/s)

.5 Surface roughness length Wm)

.20 Sector averaging constant 4.4 Initial value of sigma y 6.97 Initial value of sigma z 4.30 Expanded output for code testing not selected Total number of hours of data processed =

35064 Hours of missing data

=

406 Hours direction in window

=

8607 Hours elevated plume w/ dir, in window

=

0 Hours of calm winds

=

1874 Hours direction not in window or calm

=

24177 DISTRIBUTION

SUMMARY

DATA BY AVERAGING INTERVAL AVER.

PER.

1 2

4 8

UPPER LIM.

1.00E-02 1.OOE-02 1.00E-02 1.00E-02 LOW LIM.

1.00E-06 1.00E-06 1.00E-06 1.00E-06 ABOVE RANGE

0.

IN RANGE 10481.

13124.

16530.

20599.

BELOW RANGE

0.

ZERO 24177.

21455.

17940.

13693.

TOTAL X/Os 34658.

34579.

34470.

34292.

% NON ZERO 30.24 37.95 47.95 60.07 95th PERCENTILE x/C VALUES 5.06E-03 4.85E-03 4.07E-03 3.66E-03 95% X/Q for standard averaging intervals 0 to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 5.06E-03 2 to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 3.19E-03 8 to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 1.45E-03 1 to 4 days 1.31E-03 4 to 30 days 1.01E-03 12 1.00E-02

1. 00E-06 0.

23267.

0.

11099.

34366.

67.70 24 1.00E-02 1.00E-06 0.

27535.

0.

6654.

34189.

80.54 96 1.OOE-02

1. 00E-06 0.

33641.

0.

457.

34098.

98.66 168 1.OOE-02 1.00E-06 0.

33768, 0.

57.

33825.

99.83 360 1,OOE-02 1.OOE-06 0.

33657.

0.

33657.

100.00 720 1.00E-02 1.00E-06 0.

34289.

0.

34289.

100.00 2.94E-03 2.19E-03 1.53E-03 1.36E-03 1.18E-03 1.08E-03 HOURLY VALUE RANGE MAX x/O 5.54E-03 3.15E-03 CENTERLINE SECTOR-AVERAGE NORMAL PROGRAM COMPLETION MIN XA/

8.41E-04 4.79E-04

00 00 L0 Calculation No.: BM-03-0018-001, Version 3 Attachment C Sheet 35 of 36 (D

0)

M TSC X/Q Estimates for Unit 2 Hatch Door Release (4 years of meteorological data)

)

Program

Title:

ARCON96.

-)

Developed For:

U.S. Nuclear Regulatory Commission O

Office of Nuclear Reactor Regulation 0D Division of Reactor Program Management C"

WLJ Date:

June 25, 1997 11:00 a.m.

NRC Contacts:

J. Y. Lee Phone: (301) 415 1080 O

e-mail! jyllonrc.gov 0

J. J. Hayes Phone:

(301) 415 3167 e-mail: jjh@nrc.gov C

L. A Brown Phone:

(301) 415 1232 e-mail: lab2@nrc.gov z

Code Developer: J. V. Ramsdell Phone:

(509) 372 6316 e-mail: jramsdell@pnl.gov Code Documentation:

NUREG/CR-6331 Rev. 1 The program was prepared for an agency of the United States Government. Neither the United States Government nor any agency thereof, nor any of their employees, makes any warranty, expressed or implied, or assumes any legal liability or responsibilities for any third party's use, or the results of such use, of any portion of this program or represents that its use by such third party would not infringe privately owned rights.

Program Run 9/

8/2004 at 14:08:59 ARCON INPUT **********

Number of Meteorological Data Files 4

Meteorological Data File Names FarleyOO.met Farley0l.met Farley02.met Parley03.met Height of lower wind instrument (m)

=

10.0 Height of upper wind instrument (m)

=

45.7 Wind speeds entered as meters/second Ground-level release Release height (m) 2.7 Building Area (m^2)

=

1253.0 Effluent vertical velocity (m/s)

=

.00 Vent or stack flow (ma3/s)

=

.00 Vent or stack radius (m)

.00 Direction.. intake to source (deg)

=

056 wind direction sector width (deg)

=

90 Wind direction window (deg)

=

011 - 101 Distance to intake (m)

=

53.9 Intake height (m)

=

6.9 Terrain elevation difference (m)

=

.0

00 cO 4-0 C14 LO (U

a) co in; Z3 0

C

(.0 0

r-z Calculation No.: BM-03-0018-001, Version 3 Attachment C Sheet 36 of 36 TSC X/Q Estimates for Unit 2 Hatch Door Release (4 years of meteorological data)

Output file names T2H245.out T2H245.jfd Minimum Wind Speed (m/s)

=

.5 Surface roughness length (m)

.20 Sector averaging constant

=

4.3 Initial value of sigma y

.00 Initial value of sigma z

.00 Expanded output for code testing not selected Total number of hours of data processed =

35064 Hours of missing data

=

406 Hours direction in window

=

9206 Hours elevated plume w/ dir. in window

=

0 Hours of calm winds

=

1874 Hours direction not in window or calm

=

23578 DISTRIBUTION

SUMMARY

DATA BY AVERAGING INTERVAL AVER.

PER.

1 2

4 a

UPPER LIM.

1.00E-02 1.OOE-02 1.00E-02 1.00E-02 LOW LIM.

1.00E-06 1.00E-06 1.00E-06 1.OOE'06 ABOVE RANGE

0.

IN RANGE 11080.

13812.

17291.

21432.

BELOW RANGE

0.

ZERO 23578.

20767.

17179.

12860.

TOTAL X/Qs 34658.

34579.

34470.

34292.

1 NON ZERO 31.97 39.94 50.16 62.50 95th PERCENTILE X/Q VALUES 1.77E-03 1.73E-03 1.66E-03 1.57E-03 95% X/Q for standard averaging intervals 0 to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 1.77E-03 2 to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 1.51E-03 8 to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 7.26E-04 1 to 4 days 5.09E-04 4 to 30 days 3.95E-04 12 1.00E-02 1.00E-06 0.

24162.

0.

10204.

34366.

70.31 24

1. 00E-02 1.00E-06 0.

28377.

0.

5812.

34189.

83.00 96 1.OOE-02 1.OOE-06 0.

33701.

0.

397.

34098.

98.84 168

1. OOE-02

1. OE-06 0.

33756.

0.

69.

33825.

99.80 360 1.OOE-02 1.00E-06 0.

33657.

0.

33657.

200.00 720

1. OOE-02 1.00E-06 0.

34289.

0.

34289.

100.00 1.31E-03 1.01E-03 6.34E-04 5.47E-04 4.47E-04 4.27E-04 HOURLY VALUE RANGE MAX X/Q 2.70E-03 1.57E-03 CENTERLINE SECTOR-AVERAGE NORMAL PROGRAM COMPLETION MIn X/Q 4.15E-04 2.42E-04

cO 4-0 L0 Calculation No.: BM-03-0018--001, Version 3 (U

Unit 1 CR X/Q Es 0

P

~5 Program

Title:

ARCON96, Attachment D Sheet 1 of 36 timates for Unit I Vent Release (4.5 years of meteorological data) 0 0

w P-(D 0U 0

z Developed For:

Date:

NRC Contacts:

U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation Division of Reactor Program Management June 25, 1997 11:00 a.m.

J.

Y. Lee Phone: (301) 415 1080 e-mail: jyll@nrc.gov J.

J.

Hayes Phone:

(301) 415 3167 e-mail: jjh@nrc.gov L. A Brown Phone:

(301) 415 1232 e-mail: lab2@nrc.gov Code Developer: J.

V. Ramsdell Phone: (509) 372 6316 e-mail: j ramsdell@pnl.gov Code Documentation:

NUREG/CR-6331 Rev. 1 The program was prepared for an agency of the United States Government. Neither the United States Government nor any agency thereof, nor any of their employees, makes any warranty, expressed or implied, or assumes any legal liability or responsibilities for any third party's use, or the results of such use, of any portion of this program or represents that its use by such third party would not infringe privately owned rights.

Program Run 9/

7/2004 at 17:29:35 ARCON INPUT Number of Meteorological Data Files

=

5 Meteorological Data File Names FarleyOO.met Farley0l.met Farley02.met Farley03.met southern.met Height of lower wind instrument (m)

Height of upper wind instrument (m)

Wind speeds entered as meters/second Ground-level release Release height (m)

Building Area (mW2)

Effluent vertical velocity (m/s)

Vent or stack flow (m^3/s)

Vent or stack radius (m)

Direction.. intake to source (deg)

Wind direction sector width (deg)

Wind direction window (deg)

Distance to intake (m)

Intake height (a)

Terrain elevation difference (m)

=

10.0 45.7

=

44.3

=

.0

=

.00

=

.00

=

.00

=

115

=

90

=

070 -

160

=

60.6

=

11.4

=

.0

00 00 4-0 LO M

ar L6 V) 0 (0

C (0

0 Ci)

Co 00l 0

Calculation No.: BM-03-0018-001, Version 3 Attachment D Sheet 2 of 36 Unit 1 CR X/Q Estimates for Unit 1 Vent Release (4.5 years of meteorological data)

Output file names ClV145s.out ClV145S.jfd Minimum Wind Speed (m/s)

=

.5 Surface roughness length (m)

=

.20 Sector averaging constant

=

4.3 Initial value of sigma y

.00 Initial value of sigma z

.00 Expanded output for code testing not selected Total number of hours of data processed =

39528 Hours of missing data

=

406 Hours direction in window

=

8254 Hours elevated plume w/ dir. in window

=

0 Hours of calm winds 2112 Hours direction not in window or calm 28756 DISTRIBUTION

SUMMARY

DATA BY AVERAGING INTERVAL AVER.

PER.

1 2

4 8

UPPER LIM.

1.00E-02 1.OOE-02 1.00E-02 1.OOE-02 LOW LIM.

1.00E-06 1.00E-06 1.00E-06 1.OOE-06 ABOVE RANGE

0.

IN RANGE 10366.

13036.

16891.

21516.

BELOW RANGE

0.

ZERO 28756.

26007.

22043.

17240.

TOTAL X/Qs 39122.

39043.

38934.

38756.

% NON ZERO 26.50 33.39 43.38 55.52 95th PERCENTILE X/Q VALUES 1.62E-03 1.51E-03 1.41E-03 1.31E-03 95% X/Q for standard averaging intervals 0 to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 1.62E-03 2 to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 1.21E-03 8 to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 5.37E-04 1 to 4 days 3.35E-04 4 to 30 days 2.32E-04 12 1.00E-02

1. 00E-06 0.

24661.

0.

14169.

38830.

63.51 24 1.00E-02 1.0OE-06 0.

29858.

0.

8795.

38653.

77.25 96 1.00E-02 1.00E-06 0.

37752.

0.

810.

38562.

97.90 168 1.00E-02

1. 00E-06 0.

38100.

0.

189.

38289.

99.51 360 1.00E-02 1.00E-06 0.

38121.

0.

38121.

100.00 720 1.OOE-02

1. 00E-06 0.

38753.

0.

38753.

100.00 1.08E-03 7.95E-04 4.50E-04 3.83E-04 3.09E-04 2.61E-04 HOURLY VALUE RANGE MAX X/Q 3.07E-03 1.79E-03 CENTERLINE SECTOR-AVERAGE NORMAL PROGRAM COMPLETION MIN X/Q 2.31E-04 1.34E-04

cO 4--

00 LO_

03 0Unit 1 CR X/Q Est:

L6' Program

Title:

ARCON96.

Attachment D Sheet 3 of 36 imates for Unit 1 Reactor Release (4.5 years of meteorological data) 0-6 CwL r-to 0:

z Developed For:

Date:

U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation Division of Reactor Program Management June 25, 1997 11:00 a.m.

NRC Contacts:

J.

Y. Lee Phone:

(301) 415 1080 e-mail: jyll@nrc.gov J. J.

Hayes Phone:

(301) 415 3167 e-mail: jjh@nrc.gov L. A Brown Phone:

(301) 415 1232 e-mail: lab2@nrc.gov Code Developer: J.

V. Ramsdell Phone: (509) 372 6316 e-mail: j ramsdell@pnl.gov Code Documentation:

NUREG/CR-6331 Rev. 1 The program was prepared for an agency of the United States Government. Neither the United States Government nor any agency thereof, nor any of their employees, makes any warranty, expressed or implied, or assumes any legal liability or responsibilities for any third party's use, or the results of such use, of any portion of this program or represents that its use by such third party would not infringe privately owned rights.

Program Run 9/

7/2004 at 17:29:35 ARCON INPUT**********

Number of Meteorological Data Files

=

Meteorological Data File Names FarleyoO.met Farley0l.met Farley02.met Farley03.met southern.met Height of lower wind instrument (m)

=

Height of upper wind instrument (m)

=

Wind speeds entered as meters/second Ground-level release Release height (m)

Building Area (m2)

Effluent vertical velocity (m/s)

=

Vent or stack flow (m^3/s)

Vent or stack radius (m)

Direction.. intake to source (deg)

=

Wind direction sector width (deg)

=

Wind direction window (deg)

Distance to intake (m)

Intake height (m)

Terrain elevation difference (m)

=

5 10.0 45.7 11.4 887.0

.00

.00 136 90 091 -

181 39.2 11.4

.0

00 4--

L(0L0 Calculation No.: BM-03-0018-O0l, Version 3 Attachment D Sheet 4 of 36 Unit 1 CR X/Q Estimates for Unit 1 Reactor Release (4.5 years of meteorological data)

CO L6 Output file names CIR145s.out CiR145s.jfd O

Minimum Wind Speed (m/s)

.5 Surface roughness length (m)

=

.20 LU Sector averaging constant

=

4.3

(.0 Initial value of sigma y 6.97 O

Initial value of sigma z

=

3.54 Expanded output for code testing not selected Total number of hours of data processed =

39528 Hours direction in window

=

6613 Hours elevated plume w/ dir. in window

=

0 Hours of calm winds

=

1676 Hours direction not in window or calm

=

30833 DISTRIBUTION

SUMMARY

DATA BY AVERAGING INTERVAL AVER.

PER.

1 2

4 8

UPPER LIM.

1.00E-02 I.OOE-02 1.00E-02 1.OOE-02 LOW LIM.

1.OOE-06 I.OOE-06 1.OOE-06 1.00E-06 ABOVE RANGE

0.

0, 0.

IN RANGE 8289.

10988.

14879.

19691.

BELOW RANGE

0.

ZERO 30833.

28055.

24055.

19065.

TOTAL X/Qs 39122.

39043.

38934.

38756.

% NON ZERO 21.19 28.14 38.22 50.81 95th PERCENTILE X/Q VALUES 1.54E-03 1.41E-03 1.26E-03 1.11E-03 95% X/Q for standard averaging intervals 0 to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 1.54E-03 2 to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 9.62E-04 8 to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 4.45E-04 1 to 4 days 2.61E-04 4 to 30 days 3.09E-04 12 1.OOE-02

1. OOE-06 0.

22792.

0.

16038.

38830.

58.70 24 1.00E-02 1.O0E-06 0.

27702.

0.

10951.

38653.

71.67 96 1.00E-02 1.00E-06 0.

36956.

0.

1606.

38562.

95.84 168

1. OOE-02 1.00E-06 0.

38027.

0.

262.

38289.

99.32 360 1.00E-02 1.00E-06 0.

38121.

0.

38121.

100.00 720

1. OOE-02 1.00E-06 0.

38753.

0.

38753.

100.00 8.92E-04 6.65E-04 3.62E-04 3.39E-04 3.19E-04 3.16E-04 HOURLY VALUE RANGE MAX X/O 2.01E-03 1.17E-03 CENTERLINE SECTOR-AVERAGE NORMAL PROGRAM COMPLETION MIN X/Q 4.32E-04 2.52E-04

cO cO 00 4--

tO Calculation No.: BM-03-0018-001, Version 3 Unit 1 CR X/Q Estim Program

Title:

ARCON96.

Developed For:

U.S. Nuclear Regulatory Commiss:

Office of Nuclear Reactor Regull o

Division of Reactor Program Man C

Date:

June 25, 1997 11:00 a.m.

w NRC Contacts; J.

Y. Lee Phone: (301) 41 e-mail: jyll@nr O

J.

Hayes Phone:

(301) 41 O

e-mail: jjh@nrc r,,

L.

A Brown Phone:

(301) 41 OC e-mail: lab2@nr Attachment D Sheet 5 of 36 ates for Unit 1 Hatch Door Release (4.5 years of meteorological data) ion ation agement 5 1080 c.gov 5 3167

.gov 5 1232 c.gov

-jz Code Developer: J. V. Ramsdell Phone:

(509) 372 6316 e-mail: jramsdell*pnl.gov Code Documentation:

NUREG/CR-6331 Rev. 1 The program was prepared for an agency of the United States Government. Neither the United States Government nor any agency thereof, nor any of their employees, makes any warranty, expressed or implied, or assumes any legal liability or responsibilities for any third party's use, or the results of such use, of any portion of this program or represents that its use by such third party would not infringe privately owned rights.

Program Run 9/

7/2004 at 17:29:36 ARCON INPUT *******..

Number of Meteorological Data Files Meteorological Data File Names FarleyOO.met Farley0l.met Farley02.met Farley03.met Southern.met

=

5 Height of lower wind instrument (m)

=

Height of upper wind instrument (m)

=

Wind speeds entered as meters/second Ground-level release Release height (m)

Building Area (m'2)

Effluent vertical velocity (m/s)

=

Vent or stack flow (m^3/s)

=

Vent or stack radius (m)

Direction.. intake to source (deg)

=

Wind direction sector width (deg)

=

Wind direction window (deg)

Distance to intake (m)

=

Intake height (m)

Terrain elevation difference (m)

=

10.0 45.7 2.7 1253.0

.00

.00 136 90 091 -

181 80.4 11.4

.0

0o Co 4-0 LO Calculation No.: BM-03-0018--001, Version 3 0) co Unit 1 CR X/Q Estim L6 Output file names CiH145s.out ClH145s.jfd Q

Minimum Wind Speed (m/a) c" Surface roughness length (m)

=

W Sector averaging constant Attachment D Sheet 6 of 36 ates for Unit 1 Hatch Door Release (4.5 years of meteorological data)

.5

.20 4.3

.00

.00 Cr, 0!

0 Initial value of sigma y Initial value of sigma z Expanded output for code testing not selected Total number of hours of data processed =

39528 Hours of missing data

=

406 Hours direction in window

=

6586 Hours elevated plume w/ dir. in window

=

0 Hours Of calm winds 1874 Hours direction not in window or calm

=

30662 DISTRIBUTION

SUMMARY

DATA BY AVERAGING INTERVAL AVER.

PER.

1 2

4 8

UPPER LIM.

1.00E-02 1.00E-02 1.00E-02 1.00E-02 LOW LIM.

1.00E-06 1.00E-06 1.00E-06 1.00E-06 ABOVE RANGE

0.

IN RANGE 8460.

11220.

15179.

20021.

BELOW RANGE

0.

ZERO

30662, 27823.

23755.

18735.

TOTAL X/Qs 39122.

39043.

38934.

38756.

% NON ZERO 21.62 28.74 38.99 51.66 95th PERCENTILE X/Q VALUES 8.79E-04 8.57E-04 7.66E-04 6.62E-04 95% X/Q for standard averaging intervals 12 1.OOE-02 1.00E-06 0.

23139.

0.

15691.

38830.

59.59 24 1.00E-02 1.00E-06 0.

27992.

0.

10661.

38653.

72.42 96 1.00E-02 1.00E-06 0.

36994.

0.

1568.

38562.

95.93 168 I.OE-02 1.00E-06 0.

38036.

0.

253.

38289.

99.34 360 1.00E-02 1.00E-06 0.

38121.

0.

38121.

100.00 720 1.00E-02 1.00E-06 0.

38753.

0.

38753.

100.00 5.23E-04 3.96E-04 2.17E-04 2.06E-04 1.99E-04 1.97E-04 0 to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 2 to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 8 to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 1 to 4 days 4 to 30 days 8.79E-04 5.89E-04 2.63E-04 1.57E-04 1.93E-04 HOURLY VALUE RANGE MAX X/Q MIN X/Q 1.22E-03 2.46E-04 7.11E-04 1.43E-04 CENTERLINE SECTOR-AVERAGE NORMAL PROGRAM COMPLETION

cO 4--0 0')

a)Calculation NO.: BM-03-0018-001, Version 3 co Unit 1 CR X/Q Es 0O Attachment D Sheet 7 of 36 timates for Unit 2 Vent Release (4.5 years of meteorological data)

L6 Program

Title:

ARCON96.

'2 Developed For:

U.S. NuC U)

Office o O

Division r_

Date:

June 25, w

NRC Contacts:

J.

Y. Le O

J.

Ha 0,

lear Regulatory Commission f Nuclear Reactor Regulation

of Reactor Program Management 1997 11:00 a.m.

e Phone: (301) 415 1080 e-mail: jyll@nrc.gov yes Phone:

(301) 415 3167 e-mail: jjh@nrc.gov own Phone:

(301) 415 1232 e-mail: lab2@nrc.oov o

L.

A Br Z

Code Developer: J. V. Ramsdell Phone:

(509) 372 6316 e-mail: j ramsdell@pnl.gov Code Documentation:

NUREG/CR-6331 Rev. 1 The program was prepared for an agency of the United States Government. Neither the United States Government nor any agency thereof, nor any of their employees, makes any warranty, expressed or implied, or assumes any legal liability or responsibilities for any third party's use, or the results of such use, of any portion of this program or represents that its use by such third party would not infringe privately owned rights.

Program Run 9/ 7/2004 at 17:29:36 ARCON INPUT Number of Meteorological Data Files Meteorological Data File Names FarleyO0.met FarleyOl.met FarleyO2.met Farley03.met southern.met

- 5 Height of lower wind instrument (m)

=

Height of upper wind instrument (m)

=

wind speeds entered as meters/second Ground-level release Release height (m)

Building Area (m2)

Effluent vertical velocity (m/s)

=

Vent or stack flow (m'3/s)

=

Vent or stack radius (m)

=

10.0 45.7 44.3

.0

.00

.00 Direction..

intake to source (deg)

Wind direction sector width (deg)

Wind direction window (deg)

Distance to intake (m)

Intake height (m)

Terrain elevation difference (m)

=

066

=

90

=

021 -

111 61.1 11.4

.0

cO 00

.4-0 (D

Calculation No.: BM-03-0018-001, Version 3 Attachment D Sheet 8 of 36 aUnit 1 CR X/Q Estimates for Unit 2 Vent Release (4.5 years of meteorological data)

Output file names a(1 C1V245s.out CIV245s.jfd Ci) 0 Minimum Wind Speed (m/s)

=

.5 Surface roughness length (m)

=

.20 WU Sector averaging constant

=

4.3 Initial value of sigma y

.00 Co o

Initial value of sigma z

.00 Expanded output for code testing not selected n

_1 Total number of hours of data processed =

39528 Z

Hours of missing data 406 Hours direction in window

=

9861 Hours elevated plume w/ dir. in window

=

0 Hours of calm winds

=

2112 Hours direction not in window or calm

=

27149 DISTRIBUTION

SUMMARY

DATA BY AVERAGING INTERVAL AVER.

PER.

1 2

4 8

12 24 96 168 360 720 UPPER LIM.

1.00E-02 1.OOE-02 1.00E-02 l.G0E-02 1.00E-02 1.00E-02 1.00E-02 1.00E-02 1.00E-02 1,00E-02 LOW LIM.

1.00E-06 1.OOE-06 1.00E-06 1.OOE-06 1.00E-06 1.00E-06 1.00R-06 1.00E-06 1.00R-06 1.00E-06 ABOVE RANGE

0.

IN RANGE 11973.

14287.

17316, 21029.

23546.

28240.

38034.

38210.

38121.

38753.

BELOW RANGE

0.

ZERO 27149.

24756.

21618.

17727.

15284.

10413.

528.

79.

0.

TOTAL X/Qs 39122.

39043.

38934.

38756.

38830.

38653.

38562.

38289.

38121.

38753.

% NON ZERO 30.60 36.59 44.48 54.26 60.64 73.06 98.63 99.79 100.00 100.00 95th PERCENTILE X/Q VALUES 1.59E-03 1.54E-03 1.47E-03 1.41E-03 1.18E-03 9.43E-04 6.23E-04 5.18E-04 4.42E-04 3.94E-04 95% X/Q for standard averaging intervals 0 to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 1.59E-03 2 to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 1.35E-03 8 to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 7.08E-04 1 to 4 days 5.16E-04 4 to 30 days 3.59E-04 HOURLY VALUE RANGE MAX X/O MIN X/O CENTERLINE 3.17E-03 1.54E-04 SECTOR-AVERAGE 1.85E-03 9.00E-05 NORMAL PROGRAM COMPLETION

c0 00 4--0 CO Calculation No.: BM-03-0018-001, Version 3 (D

0)

M Unit 1 CR X/Q Est.

a.

Attachment D Sheet 9 of 36 imates for Unit 2 Reactor Release (4.5 years of meteorological data) o-Ld 0

CJ C)

Co z-o

-Jz Program

Title:

ARCON96.

Developed For:

Date:

NRC Contacts:

U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation Division of Reactor Program Management June 25, 1997 11:00 a.m.

J.

Y. Lee Phone:

(301) 415 1080 e-mail: jyllenrc.gov J.

J.

Hayes Phone:

(301) 415 3167 e-mail: jjhtnrc.gov L.

A Brown Phone: (301) 415 1232 e-mail: lab2@nrc.gov Code Developer: J.

V. Ramsdell Phone: (509) 372 6316 e-mail: j ramsdellepnl.gov Code Documentation:

NUREG/CR-6331 Rev. 1 The program was prepared for an agency of the United States Government. Neither the United States Government nor any agency thereof, nor any of their employees, makes any warranty, expressed or implied, or assumes any legal liability or responsibilities for any third party's use, or the results of such use, of any portion of this program or represents that its use by such third party would not infringe privately owned rights.

Program Run 9/

7/2004 at 17:29:37 ARCON INPUT Number of Meteorological Data Files Meteorological Data File Names FarleyOO.met Farley0l.met Farley02.met Farley03.met southern.met

=

5 Height of lower wind instrument (m)

=

Height of upper wind instrument (m)

=

Wind speeds entered as meters/second Ground-level release Release height (m)

Building Area (MA 2)

Effluent vertical velocity (m/s)

=

Vent or stack flow (m^3/s)

Vent or stack radius (m)

Direction..

intake to source (deg)

=

Wind direction sector width (deg)

Wind direction window (deg)

Distance to intake (m)

Intake height (m)

Terrain elevation difference (m) 10.0 45.7 11.4 887.0

.00

.00 042 90 357 -

087 39.2 11.4

.0

00

'4-0 C\\1 CU 0)

C/)

0~

0)

(n 0l CD I.

z Calculation No.: BM-03-0018--001, Version 3 Attachment D Sheet 10 of 36 Unit 1 CR X/Q Estimates for Unit 2 Reactor Release (4.5 years of meteorological data)

Output file names ClR245s.out CIR245s.jfd Minimum Wind Speed (m/8)

=

.5 Surface roughness length (m)

=

.20 Sector averaging constant

=

4.3 Initial value of sigma y 6.97 Initial value of sigma z 3.54 Expanded output for code testing not selected Total number of hours of data processed =

39528 Hours of missing data

=

406 Hours direction in window

=

10587 Hours elevated plume w/ dir. in window

=

0 Hours of calm winds 1676 Hours direction not in window or calm

=

26859 DISTRIBUTION

SUMMARY

DATA BY AVERAGING INTERVAL AVER.

PER.

1 2

4 8

UPPER LIM.

1.00E-02 1.00E-02 1.00E-02 1.OOE-02 LOW LIM.

1.00E-06 1.00E-06 1.00E-06 1.00E-06 ABOVE RANGE

0.

IN RANGE 12263.

15231, 18867.

23084.

BELOW RANGE

0.

ZERO 26859.

23812.

20067.

15672.

TOTAL X/Qs 39122.

39043.

38934.

38756.

% NON ZERO 31.35 39.01 48.46 59.56 95th PERCENTILE X/Q VALUES 1.64E-03 l.60E-03 1.53E-03 1.42E-03 95% X/Q for standard averaging intervals 0 to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 1.64E-03 2 to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 1.34E-03 8 to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 6.65E-04 1 to 4 days 5.45E-04 4 to 30 days 4.18E-04 12 1.00E-02 1.OOE-06 0.

2S843.

0.

12987.

38830.

66.55 24 1.00E-02 1.00E-06 0.

30309.

0.

8344.

38653.

78.41 96 1.00E-02

1. 00E-06 0.

38157.

0.

405.

38562.

98.95 168 1.OOE-02 1.00E-06 0.

38218.

0.

71.

38289, 99.81 360 1.00E-02 1.00E-06 0.

38121.

0.

38121.

100.00 720 1.00E-02 1.00E-06 0.

38753.

0.

38753.

100.00 1.18E-03 9.15E-04 6.37E-04 5.50E-04 4.75E-04 4.47E-04 HOURLY VALUE RANGE MAX X/O 2.01E-03 1.17E-03 CENTERLINE SECTOR-AVERAGE NORMAL PROGRAM COMPLETION MIN X/Q 3.39E-04 1.98E-04

00 4-0 Cr)

(0 Calculation No.: BM-03-0018-001, Version 3 0) oUnit 1 CR X/Q Estim ui Program

Title:

ARCON96.

Developed For:

U.S. Nuclear Regulatory Commiss:

Office of Nuclear Reactor Regul:

Division of Reactor Program Man Date!

June 25, 1997 11:00 a.m.

w NRC Contacts:

J.

Y. Lee Phone; (301) 41!

t-7 e-mail: jyllanri (0

o J. J. Hayes Phone: (301) 41!

o e-mail: jjh@nrc 1,L L. A Brown Phone: (301) 41' o>

e-mail: lab2@nr Attachment D Sheet 11 of 36 ates for Unit 2 Hatch Door Release (4.5 years of meteorological data) ion ation agement 5 1080 c.gov 5 3167

.gov 5 1232 c.gov 2 6316 ellspnl.gov

-j Code Developer:

J.

V. Ramsdell Phone: (509) 37:

e-mail: jramsd:

Code Documentation:

NUREG/CR-6331 Rev. 1 The program was prepared for an agency of the United States Government. Neither the United States Government nor any agency thereof, nor any of their employees, makes any warranty, expressed or implied, or assumes any legal liability or responsibilities for any third party's use, or the results of such use, of any portion of this program or represents that its use by such third party would not infringe privately owned rights.

Program Run 9/

7/2004 at 17:29:37 ARCON INPUT Number of Meteorological Data Files

=

5 Meteorological Data File Names FarleyOO.met Farley0l.met FarleyO2.met Farley03.met southern.met Height of lower wind instrument (m)

=

Height of upper wind instrument (m)

=

Wind speeds entered as meters/second Ground-level release Release height (m)

Building Area (m^2)

Effluent vertical velocity (m/s)

=

Vent or stack flow (m^3/s)

Vent or stack radius (m)

Direction..

intake to source (deg)

=

Wind direction sector width (deg)

=

Wind direction window (deg)

Distance to intake (m)

Intake height (m)

Terrain elevation difference (m)

=

10.0 45.7 2.7 1253.0

.00

.00 043 90 358 -

088 82.1 11.4

.0

0 0

C (C

LL C

C D

DI.

Calculation No.: BM-03-0018-001, Version 3 Attachment D Sheet 12 of 36 Unit 1 CR X/Q Estimates for Unit 2 Hatch Door Release (4.5 years of meteorological data)

Output file names CiH245s.out 3CIH245s.jfd 3

3 Minimum Wind Speed (m/s)

.5 Surface roughness length (m)

=

.20 Sector averaging constant

=

4.3 Initial value of sigma y

.00 3

Initial value of sigma z

.00 Expanded output for code testing not selected Total number of hours of data processed 39528 Hours of missing data 406 Hours direction in window 10415 Hours elevated plume w/ dir. in window

=

0 Hours of calm winds 1874 Hours direction not in window or calm 26833 DISTRIBUTION

SUMMARY

DATA BY AVERAGING INTERVAL AVER.

PER.

1 2

4 8

UPPER LIM.

1.00E-03 1.OOE-03 1.00E-03 1.OOE-03 LOW LIM.

1.00E-07 1.00E-07 1.OOE-07 1.00E-07 ABOVE RANGE

1.

0,

0.

IN RANGE 12288.

15279.

18951.

23163.

BELOW RANGE

0.

ZERO 26833.

23764.

19983.

15593.

TOTAL X/os 39122.

39043.

38934.

38756.

% NON ZERO 31.41 39.13 48.67 59.77 95th PERCENTILE X/Q VALUES 7.83E-04 7.59E-04 7.26E-04 6.85E-04 95% X/Q for standard averaging intervals 0 to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 7.83E-04 2 to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 6.52E-04 8 to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 3.22E-04 1 to 4 days 2.43E-04 4 to 30 days 1.85E-04 12 1.OOE-03 1.00E-07 0.

25915.

0.

12915.

38830.

66.74 24 1.00E-03 1.OOE-07 0.

30395.

0.

8258.

38653.

78.64 96 1.OOE-03 1.OOE-07 0.

38153.

147.

262.

38562.

99.32 168 1.OOE-03 1.00E-07 0.

38218.

9.

62.

38289.

99.84 360 1.OOE-03 1.00E-07 0.

38121.

0.

38121.

100.00 720 1.OOE-03 1.00E-07 0.

38753.

0.

38753.

100.00 5.72E-04 4.43E-04 2.93E-04 2.SOE-04 2.11E-04 1.99E-04 HOURLY VALUE RANGE MAX X/O 1.20E-03 6.98E-04 CENTERLINE SECTOR-AVERAGE NORMAL PROGRAM COMPLETION MIN X/Q 1.90E-04 1.11E-04

00 4--

L0)

Calculation No.: BM-03-0018-001, Version 3 0) oUnit 2 CR X/Q Es c

Program

Title:

ARCON96.

Attachment D Sheet 13 of 36 timates for Unit 1 Vent Release (4.5 years of meteorological data)

=U U) 0 LUJ Developed For:

Date:

NRC Contacts:

C0 00 C,

U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation Division of Reactor Program Management June 25, 1997 11:00 a.m.

J.

Y. Lee Phone: (301) 415 1080 e-mail: jyll@nrc.gov J.

J. Hayes Phone: (301) 415 3167 e-mail: jjh@nrc.gov L.

A Brown Phone:

(301) 415 1232 e-mail: lab2@nrc.gov

-J Code Developer: J.

V. Ramsdell Phone:

(509) 372 6316 e-mail: jramsdellepnl.gov Code Documentation:

NUREG/CR-6331 Rev.

1 The program was prepared for an agency of the United States Government. Neither the United States Government nor any agency thereof, nor any of their employees, makes any warranty, expressed or implied, or assumes any legal liability or responsibilities for any third party's use, or the results of such use, of any portion of this program or represents that its use by such third party would not infringe privately owned rights.

Program Run 9/

7/2004 at 17:29:37 ARCON INPUT Number of Meteorological Data Files Meteorological Data File Names Farley00.met Farley0l.met Farley02.met FarleyO3.met southern.met

=

5 Height of lower wind instrument (m)

=

Height of upper wind instrument (m)

=

Wind speeds entered as meters/second Ground-level release Release height (m) suilding Area fm'2)

Effluent vertical velocity (m/s)

=

Vent or stack flow (m^3/s)

Vent or stack radius (m)

Direction..

intake to source (deg)

Wind direction sector width (deg)

=

Wind direction window (deg)

=

Distance to intake (m)

=

Intake height (m)

Terrain elevation difference (m)

=

10.0 45.7 44.3

.0

.00

.00 116 90 071 -

161 61.1 11.4

.0

00 0O 14-0 CO W-Co U,

C) 0-tai C:

Co C

Calculation No.: BM-03-0018-001, Version 3 Attachment D Sheet 14 of 36 Unit 2 CR X/Q Estimates for Unit 1 Vent Release (4.5 years of meteorological data)

Output file names C2V145s.out C2Vl45s.jfd Minimum Wind Speed (m/s)

=

.5 Surface roughness length (m)

=

.20 Sector averaging constant

=

4.3 Initial value of sigma y

.00 Initial value of sigma z

.00 Expanded output for code testing not selected Total number of hours of data processed 39528 Hours of missing data

=

406 Hours direction in window

=

8169 Hours elevated plume w/ dir. in window

=

0 Hours of calm winds 2112 Hours direction not in window or calm

=

28841 DISTRIBUTION

SUMMARY

DATA BY AVERAGING INTERVAL AVER.

PER.

1 2

4 8

UPPER LIM.

1.OOE-02 1.OOE-02 1.OOE-02 1.OOE-02 LOW LIM.

1.00E-06 1.OOE-06 1.OOE-06 1.OOE-06 ABOVE RANGE

0.

IN RANGE 10281.

12936.

16785.

21396.

BELOW RANGE

0.

ZERO 28841.

26107.

22149.

17360.

TOTAL X/Qs 39122.

39043.

38934.

38756.

% NON ZERO 26.28 33.13 43.11 55.21 95th PERCENTILE X/Q VALUES 1.60E-03 1.49E-03 1.39E-03 1.29E-03 95% X/Q for standard averaging intervals 0 to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 1.60E-03 2 to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 1.19E-03 8 to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 5.23E-04 1 to 4 days 3.24E-04 4 to 30 days 2.28E-04 12 1.OOE-02 1.00E-06 0.

24537.

0.

14293.

38830.

63.19 24 1.00E-02 1.OOE-06 0.

29731.

0.

8922.

38653.

76.92 96 1.00E-02 1.O0E-06 0.

37649.

0.

913.

38562.

97.63 168 1.00S-02 1.00E-06 0.

38080.

0.

209.

38289.

99.45 360 1.00E-02 1.00E-06 0.

38121.

0.

38121.

100.00 720 1.00E-02 1.OOE-06 0.

38753.

0.

38753.

100.00 1.06E-03 7.80E-04 4.38E-04 3.75E-04 3.02E-04 2.56E-04 HOURLY VALUE RANGE MAX X/Q 3.03E-03 1.77E-03 CENTERLINE SECTOR-AVERAGE NORMAL PROGRAM COMPLETION MIN X/Q 2.28E-04 1.33E-04

00 00 4--

0 I-Calculation No.: BM-03-0018-0Ol, Version 3 0)

(U Unit 2 CR 2C/Q Est.

0-

,,j Program Title; ARCON96.

L_

Developed For:

U.S. Nuclear Regulatory CommisS Z) office of Nuclear Reactor Regul 0

Division of Reactor Program Man C:

Date:

June 25. 1997 11:00 s.m.

LU Attachment D Sheet 15 of 36 imates for Unit 1 Reactor Release (4.5 years of meteorological data) ion ation agement CD I-

-_J z

NRC Contacts:

J.

Y. Lee Phone:

(301) 415 1080 e-mail: jyll@nrc.gov J.

J. Hayes Phone.

(301) 415 3167 e-mail: jjhonrc.gov L. A Brown Phone:

(301) 415 1232 e-mail; lab2onrc.gov Code Developer: J.

V. Ramsdell Phone:

(509) 372 6316 e-mail: jramsdell@pnl.gov Code Documentation:

NUREG/CR-6331 Rev. 1 The program was prepared for an agency of the United States Government. Neither the United States Government nor any agency thereof, nor any of their employees, makes any warranty, expressed or implied, or assumes any legal liability or responsibilities for any third party's use, or the results of such use, of any portion of this program or represents that its use by such third party would not infringe privately owned rights.

Program Run 9/

7/2004 at 17:29:38 ARCON INPUT Number of Meteorological Data Files Meteorological Data File Names FarleyO0.met Farley0l.met Farley02.met Farley03.met southern.met Height of lower wind instrument (m)

Height of upper wind instrument (m) wind speeds entered as meters/second Ground-level release Release height (m)

Building Area (m^2)

Effluent vertical velocity (m/s)

Vent or stack flow (m^3/s)

Vent or stack radius (m)

Direction.. intake to source (deg)

Wind direction sector width (deg)

Wind direction window (deg)

Distance to intake (m)

Intake height (m)

Terrain elevation difference (m)

-=5 10.0 45.7 11.4 887.0

.00

.00 Z

=

138

=

90

=

093 -

183

=

39.2

=

11.4

.0

0o cO

'4-0Co (0

Calculation No.: BM-03-0018-001, Version 3 Unit 2 CR X/Q Est:

Attachment D Sheet 16 of 36 imates for Unit I Reactor Release (4.5 years of meteorological data) 0 w

CD (D0 0*

0,

-J Output file names C2R145s.out C2R145s.jfd Minimum Wind Speed (m/s)

=

.5 Surface roughness length (m)

=

.20 Sector averaging constant

=

4.3 Initial value of sigma y 6.97 Initial value of sigma z 3.54 Expanded output for code testing not selected Total number of hours of data processed =

39528 Hours of missing data

=

406 Hours direction in window

=

6604 Hours elevated plume w/ dir, in window

=

0 Hours of calm winds

=

1676 Hours direction not in window or calm

=

30842 DISTRIBUTION

SUMMARY

DATA BY AVERAGING INTERVAL AVER.

PER.

1 2

4 8

UPPER LIM.

1.00E-02 1.OOE-02 1.00E-02 1.OOE-02 LOW LIM.

1.00E-06 1.OOE-06 1.00E-06 1.00E-06 ABOVE RANGE

0.

IN RANGE 8280.

11017.

14864.

19584.

BELOW RANGE

0.

ZERO 30842.

28026.

24070.

19172.

TOTAL X/Qs 39122.

39043.

38934.

38756.

% NON ZERO 21.16 28.22 38.18 50.53 95th PERCENTILE X/Q VALUES 1.54E-03 1.41E-03 1.25E-03 1.10E-03 95% X/Q for standard averaging intervals 0 to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 1.54E-03 2 to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 9.50E-04 8 to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 4.43E-04 1 to 4 days 2.70E-04 4 to 30 days 3,11E-04 12 1.00E-02 1.00E-06 0.

22644.

0.

16186.

38830.

58.32 24 1.OOE-02 1.00E-06 0.

27515.

0.

11138.

38653.

71.18 96 1.00E-02 1.OOE-06 0.

36821.

0.

1741.

38562.

95.49 168 1.OOE-02 1.00E-06 0.

37999.

0.

290.

38289.

99.24 360 1.00E-02 1.0OE-06 0.

38121.

0.

38121.

100.00 720 1.00E-02 1.00E-06 0.

38753.

0.

38753.

100.00 8.86E-04 6.61E-04 3.68E-04 3.50E-04 3.28E-04 3.19E-04 HOURLY VALUE RANGE MAX X/Q 2.01E-03 1.17E-03 CENTERLINE SECTOR-AVERAGE NORMAL PROGRAM COMPLETION MIN X/Q 4.32E-04 2.52E-04

00 00 0

0')

(0 Calculation No.: BM-03-0018-O0l, Version 3 0)

W Unit 2 CR X/Q Estini CL Attachment D Sheet 17 of 36 ates for Unit 1 Hatch Door Release (4.5 years of meteorological data)

(C6 a) 0 C:

C)

C..-

z Program

Title:

ARCON96, Developed For:

Date:

U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation Division of Reactor Program Management June 25, 1997 11:00 a.m.

NRC Contacts:

J.

Y. Lee Phone: (301) 415 1080 e-mail: jyllonrc.gov J.

J.

Hayes Phone: (301) 415 3167 e-mail: jjhonrc.gov L. A Brown Phone: (301) 415 1232 e-mail: lab2@nrc.gov Code Developer: J.

V. Ramsdell Phone: (509) 372 6316 e-mail; jramsdellgpnl.gov Code Documentation:

NUREG/CR-6331 Rev.

1 The program was prepared for an agency of the United States Government. Neither the United States Government nor any agency thereof, nor any of their employees, makes any warranty, expressed or implied, or assumes any legal liability or responsibilities for any third party's use, or the results of such use, of any portion of this program or represents that its use by such third party would not infringe privately owned rights.

Program Run 9/

7/2004 at 17:29:38 ARCON INPUT Number of Meteorological Data Files Meteorological Data File Names FarleyOO.met Farley01.met Farley02.met FarleyO3.met southern.met S

5 Height of lower wind instrument (m) =

Height of upper wind instrument (m)

=

Wind speeds entered as meters/second Ground-level release Release height (m)

Building Area (m^2)

Effluent vertical velocity (m/s)

=

Vent or stack flow (m-3/s)

=

Vent or stack radius (m)

Direction..

intake to source (deg)

=

Wind direction sector width (deg)

=

wind direction window (deg)

=

Distance to intake (m)

=

Intake height (m)

Terrain elevation difference (m)

=

10.0 45.7 2.7 1253.0

.00

.00 137 90 092 -

182 82.1 11.4

.0'

cO cO DO 4-0 N--

Calculation No.: BM-03-0018--001, Version 3 (Unit 2 CR X/Q Estim Attachment D Sheet 18 of 36 ates for Unit 1 Hatch Door Release (4.5 years of meteorological data)

(/)

0 CD LU Co 0

I.

z Output file names C2H145s.out C2Hl45s.jfd Minimum Wind Speed (m/s)

Surface roughness length (m)

Sector averaging constant Initial value of sigma y Initial value of sigma z

.5

.20 4.3

.00

.00 Expanded output for code testing not selected Total number of hours of data processed =

39528 Hours of missing data

=

406 Hours direction in window

=

6586 Hours elevated plume w/ dir. in window

=

0 Hours of calm winds

=

1874 Hours direction not in window or calm 30662 DISTRIBUTION

SUMMARY

DATA BY AVERAGING INTERVAL AVER.

PER.

1 2

4 8

UPPER LIM.

1.00E-02 1.OOE-02 1.00E-02 1.00E-02 LOW LIM.

1.002-06 1.00E-06 1.002-06 1.002-06 ABOVE RANGE

0.

IN RANGE 8460.

11223.

15147.

19992.

BELOW RANGE

0.

ZERO 30662.

27820.

23787.

18764.

TOTAL X/Qs 39122.

39043.

38934.

38756.

% NON ZERO 21.62 28.75 38.90 51.58 95th PERCENTILE X/Q VALUES 8.56E-04 8.202-04 7.45E-04 6.39E-04 95% X/Q for standard averaging intervals 12 1.00E-03 1.00E-07 0.

23080.

0.

15750.

38830.

59.44 24 1.00E-03 1.002-07 0.

27898.

0.

10755.

38653.

72.18 96 1.00E-03 1.002-07 0.

36916.

0.

1646.

38562.

95.73 168 1.00E-03 1.00E-07 0.

38019.

0.

270.

38289.

99.29 360 1.00E-03

1. OOE-07 0.

38121.

0.

38121.

100.00 720 1.00E-03 1.00E-07 0.

38753.

0.

38753.

100.00 5.072-04 3.80E-04 2.092-04 2.03E-04 1.95E-04 1.94E-04 0 to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 2 to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 8 to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 1 to 4 days 4 to 30 days 8.56E-04 5.672-04 2.50E-04 1.522-04 1.92E-04 HOURLY VALUE RANGE MAX x/O 1.17E-03 6.84E-04 CENTERLINE SECTOR-AVERAGE MIN X/O 2.36E-04 1.382-04 NORMAL PROGRAM COMPLETION

0000

.4-0 T--

1-Calculation No.: BM-03-0018-001, Version 3 Attachment D Sheet 19 of 36 0)

MD Unit 2 CR X/Q Estimates for Unit 2 Vent Release (4.5 years of meteorological data) 0_

Program

Title:

ARCON96.

0)

Developed For:

U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation W9 Division of Reactor Program Management Date:

June 25, 1997 11:00 a.m.

w _

NRC Contacts:

J.

Y. Lee Phone:

(301) 415 1080 N-:

e-mail: jyll@nrc.gov O

J. J. Hayes Phone: (301) 415 3167 o

e-mail: jjh@nrc.gov N4.

L.

A Brown Phone: (301) 415 1232 o

e-mail: lab2@nrc.gov

-J Z

Code Developer: J.

V. Ramsdell Phone: (509) 372 6316 e-mail: jramsdell@pnl.gov Code Documentation:

NUREG/CR-6331 Rev. 1 The program was prepared for an agency of the United States Government. Neither the United States Government nor any agency thereof, nor any of their employees, makes any warranty, expressed or implied, or assumes any legal liability or responsibilities for any third party's use, or the results of such use, of any portion of this program or represents that its use by such third party would not infringe privately owned rights.

Program Run 9/

7/2004 at 17:29:38 ARCON INPUT*

Number of Meteorological Data Files

=

5 Meteorological Data File Names FarleyOO.met Farley0l.met Farley02.met Farley03.met southern.met Height of lower wind instrument Wm) =

10.0 Height of upper wind instrument (m)

=

45.7 Wind speeds entered as meters/second Ground-level release Release height (a)

=

44.3 Building Area (m2)

=

.0 Effluent vertical velocity (m/s)

=

.00 Vent or stack flow (m^3/s)

=

.00 Vent or stack radius Wm)

=

.00 Direction.. intake to source (deg)

=

067 Wind direction sector width (deg)

=

90 Wind direction window (deg)

=

022 - 112 Distance to intake Wm)

=

60.6 Intake height (m)

=

11.4 Terrain elevation difference Wm)

=

.0,

DO 00

'4-0 C14 r__(1) 0)

CU (0

00 0

Calculation No.: BM-03-0018-O01, Version 3 Attachment D Sheet 20 of 36 Unit 2 CR X/Q Estimates for Unit 2 Vent Release (4.5 years of meteorological data)

Output file names C2V245s.out C2V245s.jfd Minimum Wind Speed (m/s)

.5 Surface roughness length (m)

.20 Sector averaging constant

=

4.3 Initial value of sigma y

=

.00 Initial value of sigma z

=

.00 Expanded output for code testing not selected Total number of hours of data processed =

39528 Hours of missing data

=

406 Hours direction in window

=

9836 Hours elevated plume w/ dir. in window

=

0 Hours of calm winds

=

2112 Hours direction not in window or calm

=

27174 DISTRIBUTION

SUMMARY

DATA BY AVERAGING INTERVAL AVER.

PER.

1 2

4 8

UPPER LIM.

1.OOE-02 1.OOE-02 1.00E-02 1.00E-02 LOW LIM.

1.00E-06 1.00E-06 1.00-E06 1.00E-06 ABOVE RANGE

0.

IN RANGE 11948.

14253.

17287.

20979.

BELOW RANGE

0.

ZERO 27174.

24790.

21647.

17777.

TOTAL X/Qs 39122.

39043.

38934.

38756.

% NON ZERO 30.54 36.51 44.40 54.13 95th PERCENTILE X/Q VALUES 1.60E-03 1.56E-03 1.49E-03 1.42E-03 95% X/Q for standard averaging intervals 0 to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 1.60E-03 2 to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 1.37E-03 8 to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 7.10E-04 1 to 4 days 5.21E-04 4 to 30 days 3.60E-04 12 1.00E-02 1.00E-06 0.

23499.

0.

15331.

38830.

60.52 24 1.00E-02 1.00E-06 0.

28192.

0.

10461.

38653.

72.94 96 1.00E-02

1. 00E-06 0.

38034.

0.

528.

38562.

98.63 168 1.00E-02 1.00E-06 0.

38209.

0.

80.

38289.

99.79 360 1.00E-02 1.00E-06 0.

38121.

0.

38121.

100.00 720 1.OOE-02 1.00E-06 0.

38753.

0.

38753.

100.00 1.19E-03 9.48E-04 6.28E-04 5.22E-04 4.43E-04 3.96E-04 HOURLY VALUE RANGE MAX X/Q 3.21E-03 1.87E-03 CENTERLINE SECTOR-AVERAGE NORMAL PROGRAM COMPLETION MIN X/O

1. 56E-04 9.11E-05

00 4-0 CO, Calculation No.: BM-03-0018-001, Version 3 a,

CO Unit 2 CR X/Q Est

-)

Deeoe Or:

iU.S.

Nuclar Reguatory Commiss U0o Division of Reactor Program Manm

-6 a-Date:

June 25, 1997 11:00 a.m.

w NRC Contacts:

J.

Y. Lee Phone:

(301) 41

(/*

e-mail : jyllonrc o

J. J. Hayes Phone:

(301) 41!

o e-mail: jjhsnrc C..

L. A Brown Phone:

(301) 41!

o e-mail: lab2@nri Attachment D Sheet 21 of 36 imates for Unit 2 Reactor Release (4.5 years of meteorological data) ion ation agement 5 1080 c.gov 5 3167

.gov 5 1232 c.gov

-jz7 Code Developer:

J.

V. Ramsdell Phone:

(509) 372 6316 e-mail: jramsdell@pnl.gov Code Documentation:

NUREG/CR-6331 Rev. 1 The program was prepared for an agency of the United States Government. Neither the United States Government nor any agency thereof, nor any of their employees, makes any warranty, expressed or implied, or assumes any legal liability or responsibilities for any third party's use, or the results of such use, of any portion of this program or represents that its use by such third party would not infringe privately owned rights.

Program Run 9/

7/2004 at 17:29:39

- - - - ARCON INPUT Number of Meteorological Data Files

=

Meteorological Data File Names FarleyOO.met FarleyOl.met Farley02.met FarleyO3.met southern.met Height of lower wind instrument (m)

Height of upper wind instrument (m)

Wind speeds entered as meters/second Ground-level release Release height (m)

Building Area (m^2)

Effluent vertical velocity (m/s)

Vent or stack flow (m^3/s)

Vent or stack radius (m)

Direction..

intake to source (deg)

=

Wind direction sector width (deg)

=

Wind direction window (deg)

Distance to intake (m)

Intake height (m)

Terrain elevation difference (m)

=

5 10.0 45.7 11.4 887.0

.00

.00 045 90 000 -

090 39.2 11.4

co coO 4-0 Calculation No.: BM-03-0018--001, Version 3 Attachment D Sheet 22 of 36 0)

MUnit 2 CR X/Q Estimates for Unit 2 Reactor Release (4.5 years of meteorological data) 0_

L(n L.J Ci) 0 0

Output file names C2R245s.out C2R245s.jfd Minimum Wind Speed (m/s)

=

.5 Surface roughness length (m)

.20 Sector averaging constant 4.3 Initial value of sigma y 6.97 Initial value of sigma z 3.54 Expanded output for code testing not selected Total number of hours of data processed 39528 Hours of missing data

=

406 Hours direction in window 10097 Hours elevated plume w/ dir. in window 0

Hours of calm winds 1676 Hours direction not in window or calm

=

27349 DISTRIBUTION

SUMMARY

DATA BY AVERAGING INTERVAL AVER. PER.

1 2

4 8

UPPER LIM.

1.00E-02 1.OOE-02 1.00E-02 1.OOE-02 LOW LIM.

1.00E-06 1.00E-06 1.00E-06 1.00E-06 ABOVE RANGE

0.

IN RANGE 11773.

14702.

18359.

22624.

BELOW RANGE

0.

ZERO 27349.

24341.

20575.

16132.

TOTAL X/Qs 39122.

39043.

38934.

38756.

% NON ZERO 30.09 37.66 47.15 58.38 95th PERCENTILE X/Q VALUES 1.64E-03 1.59E-03 1.51E-03 1.40E-03 12 1.OOE-02 1100E-06 0.

25405.

0.

13425.

38830.

65.43 24 I.OOE-02 1.OOE-06 0.

29973.

0.

8680.

38653.

77.54 96 1.00E-02 1.00E-06 0.

38124.

0.

438.

38562.

98.86 168 1.00E-02 I.OOE-06 0.

38218.

0.

71.

38289.

99.81 360 1.OOE-02 1.00E-06 0.

38121.

0.

38121.

100.00 720 1.00E-02 1.00E-06 0.

38753.

0.

38753.

100.00 1.17E-03 9-06E-04 6.22E-04 5.37F-04 4.61E-04 4.31E-04 95% X/Q for standard averaging intervals 0 to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 1.64E-03 2 to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 1.32E-03 8 to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 6.60E-04 1 to 4 days 5.27E-04 4 to 30 days 4.01E-04 HOURLY VALUE RANGE MAX X/Q 2.01E-03 1.17E-03 CENTERLINE SECTOR-AVERAGE NORMAL PROGRAM COMPLETION MIN X/Q 3.39E-04 1.98E-04

co 00 4--

U')

Calculation No.: BM-03-0018-001, Version 3 0) 0Unit 2 CR X/Q Estim 06 Program

Title:

ARCON96.

Attachment D Sheet 23 of 36 ates for Unit 2 Hatch Door Release (4.5 years of meteorological data) 0 w

C)

CI z

Developed For:

Date:

NRC Contacts:

U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation Division of Reactor Program Management June 25, 1997 11:00 a.m.

J.

Y. Lee Phone:

(301) 415 1080 e-mail: jyll@nrc.gov J.

J.

Hayes Phone:

(301) 415 3167 e-mail: jjh@nrc.gov L. A Brown Phone:

(301) 415 1232 e-mail: lab2@nrc.gov Code Developer: J. V. Ramsdell Phone:

(509) 372.6316 e-mail: jramsdellopnl.gov Code Documentation:

NUREG/CR-6331 Rev. 1 The program was prepared for an agency of the United States Government. Neither the United States Government nor any agency thereof, nor any of their employees, makes any warranty, expressed or implied, or assumes any legal liability or responsibilities for any third party's use, or the results of such use, of any portion of this program or represents that its use by such third party would not infringe privately owned rights.

Program Run 9/

7/2004 at 17:29:39

-****** ARCON INPUT Number of Meteorological Data Files Meteorological Data File Names FarleyO0.met Farley01.met Farley02.met Farley03.met southern.met Height of lower wind instrument (m)

Height of upper wind instrument Wm) wind speeds entered as meters/second Ground-level release Release height (W)

Building Area (m^2)

Effluent vertical velocity (m/s)

Vent or stack flow (m^3/s)

Vent or stack radius Wm)

Direction.. intake to source (deg)

Wind direction sector width (deg)

Wind direction window (deg)

Distance to intake (W)

Intake height Wm)

Terrain elevation difference (m)

-5 10.0 45.7 2.7 1253.0

=

.00

=

.00

=

.00 045

=

90

=

000 -

090 80.4

=

11.4

.0

00 co 4-0 (0

N I,-

02)

LU 0~

0 w

o C:

0 0

_J Z

Calculation No.: BM-03-0018-001, Version 3 Attachment D Sheet 24 of 36 Unit 2 CR X/Q Estimates for Unit 2 Hatch Door Release (4.5 years of meteorological data)

Output file names C2H245s.out C2H245s.jfd Minimum Wind Speed (m/s)

.5 Surface roughness length (m)

.20 Sector averaging constant

=

4.3 Initial value of sigma y

.00 Initial value of sigma z

.00 Expanded output for code testing not selected Total number of hours of data processed =

39528 Hours of missing data

=

406 Hours direction in window

=

10025 Hours elevated plume w/ dir. in window

=

0 Hours of calm winds

=

1874 Hours direction not in window or calm

=

27223 DISTRIBUTION

SUMMARY

DATA BY AVERAGING INTERVAL AVER.

PER.

1 2

4 8

UPPER LIM.

1.OOE-02 1.OOE-02 1.00E-02 1.OOE-02 LOW LIM.

1.00E-06 1.OOE-06 1.OOE-06 1.00E-06 ABOVE RANGE

0.

IN RANGE 11899.

14857.

18538.

22801.

BELOW RANGE

0.

ZERO 27223.

24186.

20396.

15955.

TOTAL X/Qs 39122.

39043.

38934.

38756.

% NON ZERO 30.42 38.05 47.61 58.83 95th PERCENTILE X/Q VALUES 8.18E-04 7.87E-04 7.54E-04 7.12E-04 95% X/Q for standard averaging intervals 0 to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 8.18E-04 2 to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 6.77E-04 8 to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 3.32E-04 1 to 4 days 2.49E-04 4 to 30 days 1.89E-04 12 1.00E-03 1.OOE-07 0.

25574.

0.

13256.

38830.

65.86 24 1.00E-03 1.00E-07 0.

30132.

0.

8521.

38653.

77.96 96 1.00E-03 1.OOE-07 0.

38147.

0.

415.

38562.

98.92 168 1.00E-03 1.00E-07 0.

38218.

0.

71.

38289.

99.81 360 1.OOE-03

1. OOE-07 0.

38121.

0.

38121.

100.00 720 1.00E-03 1.00E-07 0.

38753.

0.

38753.

100.00 5.96E-04 4.59E-04 3.01E-04 2.57E-04 2.13E-04 2.04E-04 HOURLY VALUE RANGE MAX X/Q 1.25E-03 7.26E-04 CENTERLINE SECTOR-AVERAGE NORMAL PROGRAM COMPLETION MIN X/Q 1.97E-04 1.15E-04

00

'4-0 Calculation No.: BM-03-0018-001, Version 3 (V

TSC X/Q Estim utj Program

Title:

ARCON96.

Attachment D Sheet 25 of 36 ates for Unit 1 Vent Release (4.5 years of meteorological data) 0)

L..

O C:)

0 rw K

Co z-z Developed For:

Date:

NRC Contacts:

U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation Division of Reactor Program Management June 25, 1997 11:00 a.m.

J.

Y. Lee Phone: (301) 415 1080 e-mail: jyll@nrc.gov J.

J.

Hayes Phone: (301) 415 3167 e-mail: jjh@nrc.gov L. A Brown Phone: (301) 415 1232 e-mail: lab2*nrc.gov Code Developer: J.

V. Ramsdell Phone:

(509) 372 6316 e-mail: j ramsdellepnl.gov Code Documentation:

NUREG/CR-6331 Rev.

I The program was prepared for an agency of the United States Government. Neither the United States Government nor any agency thereof, nor any of their employees, makes any warranty, expressed or implied, or assumes any legal liability or responsibilities for any third party's use, or the results of such use, of any portion of this program or represents that its use by such third party would not infringe privately owned rights.

Program Run 9/ 7/2004 at 17:29:40 ARCON INPUT Number of Meteorological Data Files Meteorological Data File Names Farley00.met Farley0l.met Farley02.met FarleyO3.met southern.met

=

5 Height of lower wind instrument (m)

=

Height of upper wind instrument (ma

)

Wind speeds entered as meters/second Ground-level release Release height (m)

Building Area (mW2)

=

Effluent vertical velocity (m/s)

=

Vent or stack flow (m^3/s)

=

Vent or stack radius (m)

=

Direction..

intake to source (deg)

=

Wind direction sector width (deg)

=

Wind direction window (deg)

=

Distance to intake (m)

Intake height (m)

Terrain elevation difference (m)

=

10.0 45.7 44.3

.0

.00

.00 141 90 096 -

186 70.2 6.9

.0

00 0o 4-0 b*

Calculation No.: BM-03-0018-001, Version 3 Attachment D Sheet 26 of 36 a)

MTSC X/Q Estimates for Unit 1 Vent Release (4.5 years of meteorological data)

Cl-a, Output file names T2V145s.out T2Vl4Ss.jfd Ci) 0 Minimum Wind Speed (m/s)

=

.5 Surface roughness length (m)

.20 W

Sector averaging constant

=

4.3 Pl7 D

Initial value of sigma y

.00 CD Initial value of sigma z

=

.00 Expanded output for code testing not selected Total number of hours of data processed =

39528 Z

Hours of missing data

=

406 Hours direction in window

=

7528 Hours elevated plume w/ dir. in window

=

0 Hours of calm winds

=

2112 Hours direction not in window or calm

=

29482 DISTRIBUTION

SUMMARY

DATA BY AVERAGING INTERVAL AVER.

PER.

1 2

4 a

12 24 96 168 360 720 UPPER LIM.

I.OOE-02 1.OOE-02 1.OOE-02 1.OOE-02 1.00E-02 1.OOE-02 1.00E-02 1.00E-02 1.OOE-02 1.OOE-02 LOW LIM.

1.OOE-06 l.OE-06 1.OOE-06 1.OOE-06 1.00E-06 1.OOE-06 1.00E-06 1.00E-06 1.00E-06 1.00E-06 ABOVE RANGE

0.

IN RANGE 9640.

12370.

16280.

20726.

23591.

28129.

36679.

37863.

38121.

38753.

BELOW RANGE

0.

ZERO 29482.

26673.

22654.

18030.

15239.

10524.

1883.

426.

0.

TOTAL X/Qs 39122.

39043, 38934.

38756.

38830.

386S3.

38562.

38289.

38121.

38753.

% NON ZERO 24.64 31.68 41.81 53.48 60.75 72.77 95.12 98.89 100.00 100.00 95th PERCENTILE X/Q VALUES 1.23E-03 1.16E-03 1.06E-03 9.59E-04 7.81E-04 5.63E-04 3.03E-04 2.88E-04 2.68E-04 2.64E-04 95% X/Q for standard averaging intervals 0 to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 1.23E-03 2 to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 8.68E-04 8 to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 3.65E-04 1 to 4 days 2.17E-04 4 to 30 days 2.57E-04 HOURLY VALUE RANGE MAX X/1 MIN X/Q CENTERLINE 2.33E-03 2.39E-04 SECTOR-AVERAGE 1.36E-03 1.40E-04 NORMAL PROGRAM COMPLETION

00

.4-0 0) 0)

C:

rL.

C) z Calculation No.: BM-03-0018-001, Version 3 Attachment D Sheet 27 of 36 TSC X/Q Estimates for Unit 1 Reactor Release (4.5 years of meteorological data)

Program

Title:

ARCON96.

Developed For:

Date:

U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation Division of Reactor Program Management June 25, 1997 11:00 a.m.

NRC Contacts:

J.

Y. Lee Phone:

(301) 415 1080 e-mail: jyll@nrc.gov J.

J. Hayes Phone:

(301) 415 3167 e-mail: jjh@nrc.gov L.

A Brown Phone:

(301) 415 1232 e-mail: lab2@nrc.gov Code Developer: J.

V. Ramsdell Phone: (509) 372 6316 e-mail: jramsdell@pnl.gov Code Documentation:

NUREG/CR-6331 Rev. 1 The program was prepared for an agency of the United States Government. Neither the United States Government nor any agency thereof, nor any of their employees, makes any warranty, expressed or implied, or assumes any legal liability or responsibilities for any third party's use, or the results of such use, of any portion of this program or represents that its use by such third party would not infringe privately owned rights.

Program Run 9/

7/2004 at 17:29:40

....... ARCON INPUT Number of Meteorological Data Files

=

5 Meteorological Data File Names FarleyOO.met Farley0l.met FarleyO2.met FarleyO3.met southern.met Height of lower wind instrument (m)

=

Height of upper wind instrument (m)

=

Wind speeds entered as meters/second Ground-level release Release height (m)

Building Area (Wn2)

Effluent vertical velocity (m/s)

=

Vent or stack flow (m^3/s)

Vent or stack radius (m)

Direction..

intake to source (deg)

=

Wind direction sector width (deg)

=

Wind direction window (deg)

Distance to intake (m)

Intake height (m)

Terrain elevation difference (m) 10.0 45.7 6.9 1078.0

.00

.00 157 90 112 - 202 57.6 6.9

.0

co co

.4-0 00 Lo ci) 0o (U

IL 09 C:w I-z Calculation No.: BM-03-0018-001, Version 3 Attachment D Sheet 28 of 36 TSC X/Q Estimates for Unit 1 Reactor Release (4.5 years of meteorological data)

Output file names T2R145s.out T2RI45s.jfd Minimum Wind Speed (m/s)

Surface roughness length (m)

Sector averaging constant Initial value of sigma y Initial value of sigma z

.5

.20 4.3 6.97 4.30 Expanded output for code testing not selected Total number of hours of data processed =

39528 Hours of missing data

=

406 Hours direction in window

=

6714 Hours elevated plume w/ dir. in window

=

0 Hours of calm winds

=

1874 Hours direction not in window or calm

=

30534 DISTRIBUTION

SUMMARY

DATA BY AVERAGING INTERVAL AVER.

PER.

UPPER LIM.

LOW LIM.

ABOVE RANGE IN RANGE BELOW RANGE ZERO TOTAL X/Qs

% NON ZERO 1

1.OOE-02 1.OOE-06 0.

8588.

0.

30534.

39122.

21.95 2

1.OOE-02 1.00E-06 0'.

11248.

0.

27795.

39043.

28.81 4

1.00E-02 1.OOE-06 0.

14858.

0.

24076.

38934.

38.16 8

1.OOE-02

1. OOE-06 0.

19074.

0.

19682.

38756.

49.22 12 1.00E-03 1.00E-07 0.

21865.

0.

16965.

38830.

56.31 24 1.OOE-03 1.00E-07 0.

26737.

0.

11916.

38653.

69.17 96

1. 00E-03 1.00E-07 0.

36119.

0.

2443.

38562.

93.66 168 1.00E-03

1. OOE-07 0.

37713.

0.

576.

38289.

98.50 360 1.OOE-03 1.OOE-07 0.

38121.

0.

38121.

100.00 720 1.00E-03 1.00E-07 0.

38753.

0.

38753.

100.00 95th PERCENTILE X/Q VALUES 8.74E-04 8.18E-04 7.26E-04 6.66E-04 5.45E-04 4.25E-04 3.O01E-04 2,91E-04 2.89E-04

2. 88E-04 95% X/Q for standard averaging intervals 0 to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 2 to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 8 to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 1 to 4 days 4 to 30 days 8.74E-04 5.97E-04 3.04E-04 2.59E-04 2.86E-04 HOURLY VALUE RANGE MAX X/Q MIN X/O 1.15E-03 2.26E-04 6.73E-04 1.32E-04 CENTERLINE SECTOR-AVERAGE NORMAL PROGRAM COMPLETION

~0 4-0T_

OD Calculation No.: BM-03-0018-001, Version 3 Attachment D Sheet 29 of 36 (1)

0)

TSC X/Q Estimates for Unit 1 Hatch Door Release (4.5 years of meteorological data) 0L.

Program

Title:

ARCON96.

Q)

Developed For:

U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation (1)

Division of Reactor Program Management 0

O Date:

June 25, 1997 11:00 a.m.

NRC Contacts:

J. Y. Lee Phone:

(301) 415 1080 1_:

e-mail: jyll@nrc.gov (03 J.J.

Hayes Phone:

(301) 415 3167 o

e-mail: jjh@nrc.gov L.

A Brown Phone:

(301) 415 1232 O

e-mail: lab2@nrc.gov

_J Z

Code Developer: J. V. Ramsdell Phone: (509) 372 6316 e-mail: j ramsdell@pnl.gov Code Documentation:

NUREG/CR-6331 Rev. 1 The program was prepared for an agency of the United States Government. Neither the United States Government nor any agency thereof, nor any of their employees, makes any warranty, expressed or implied, or assumes any legal liability or responsibilities for any third party's use, or the results of such use, of any portion of this program or represents that its use by such third party would not infringe privately owned rights.

Program Run 9/ 7/2004 at 17:29:40

- - ~***

ARCON INPUT **********

Number of Meteorological Data Files

=

5 Meteorological Data File Names Farley00.met Farley0l.met Farley02.met Parley03.met southern.met Height of lower wind instrument (m)

=

10.0 Height of upper wind instrument (m)

=

45.7 wind speeds entered as meters/second Ground-level release Release height (m)

=

2.7 Building Area (m2) 1253.0 Effluent vertical velocity (m/s)

=

.00 Vent or stack flow (m^3/s)

=

.00 Vent or stack radius (m)

=

.00 Direction.. intake to source (deg) 157 Wind direction sector width (deg)

=

90 Wind direction window (deg)

=

112 - 202 Distance to intake (m)

=

100.7 Intake height (m)

=

6.9 Terrain elevation difference (m)

=

.0

DO 00 (Ni 03 04 00 (D

0)

M (0

0 0

I'L z

Calculation No.: BM-03-0018--001, Version 3 Attachment D Sheet 30 of 36 TSC X/Q Estimates for Unit 1 Hatch Door Release (4.5 years of meteorological data)

Output file names T2H145s.out T2H145s.jfd Minimum Wind Speed (m/s)

=

.5 Surface roughness length (m)

=

.20 Sector averaging constant

=

4.3 Initial value of sigma y

=

.00 Initial value of sigma z

.00 Expanded output for code testing not selected Total number of hours of data processed 39528 Hours of missing data

=

406 Hours direction in window

=

6714 Hours elevated plume w/ dir, in window

=

0 Hours of calm winds

=

1874 Hours direction not in window or calm

=

30534 DISTRIBUTION

SUMMARY

DATA BY AVERAGING INTERVAL AVER.

PER.

1 2

4 8

UPPER LIM.

1.00E-03 1.00E-03 1.00E-03 1.00E-03 LOW LIM.

1.OOE-07 1.00E-0?

1.00E-07 1.0OER-07 ABOVE RANGE

0.

IN RANGE 8588.

11248.

14858.

19074.

BELOW RANGE

0.

-0.

0.

ZERO 30534.

27795.

24076.

19682.

TOTAL X/Qs 39122.

39043.

38934.

38756.

% NON ZERO 21.95 28.81 38.16 49.22 95th PERCENTILE X/Q VALUES 5,94E-04 5.69E-04 5.26E-04 4.80E-04 12 1.00E-03 1.00E-07 0.

21865.

0.

16965.

38830.

56.31 24 1.00E-03 1.00E-07 0.

26737.

0.

11916.

38653.

69.17 96 1.00E-03 1.00E-07 0.

36119.

0.

2443.

38562.

93.66 168 1.00E-03 1.00E-07 0.

37713.

0.

576.

38289.

98.50 360 1.00E-03 1.00E-07 0.

38121.

0.

38121.

100.00 720 1.00E-03 1.00E-07 0.

38753.

0.

38753.

100.00 3.88E-04 3.06E-04 2.06E-04 2.018-04 1.99E-04 1.97E-04 95% X/Q for standard averaging intervals 0 to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 5.94E-04 2 to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 4.43E-04 8 to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 2.18E-04 I to 4 days 1.73E-04 4 to 30 days 1.96E-04 HOURLY VALUE RANGE MAX X/Q 8.04E-04 4.68E-04 CENTERLINE SECTOR-AVERAGE NORMAL PROGRAM COMPLETION MIN X/Q 1.46E-04 8.53E-05

cO cO O4-0 0O Calculation No.: BM-03-0018-O01, Version 3

0)

TSC X/Q Estim Program

Title:

ARCON96.

L6)

()

Developed For:

U.S. Nuclear Regulatory Commissi Office of Nuclear Reactor Regulz U)

Division of Reactor Program Mane 0

C Date:

June 25, 1997 11:00 a.m.

C W

NRC Contacts:

J.

Y. Lee Phone:

(301) 41l N-e-mail: jyllanr CO J.

J.

Hayes Phone:

(301) 41!

0 e-mail: jjhanrc C)

L. A Brown Phone:

(301) 41!

a" e-mail: lab2@nr:

Code Developer: J.

V. Ramsdell Phone:

(509) 37:

Attachment D Sheet 31 of 36 ates for Unit 2 Vent Release (4.5 years of meteorological data) ion ation agement 5 1080 c.gov 5 3167

.gov 5 1232 c.gov 2 6316 e-mail: jramsuel+/-@pnl.gov Code Documentation:

NUREG/CR-6331 Rev. 1 The program was prepared for an agency of the United States Government. Neither the United States Government nor any agency thereof, nor any of their employees, makes any warranty, expressed or implied, or assumes any legal liability or responsibilities for any third party's use, or the results of such use, of any portion of this program or represents that its use by such third party would not infringe privately owned rights.

Program Run 9/

7/2004 at 17:29:41 ARCON INPUT Number of Meteorological Data Files

=

5 Meteorological Data File Names FarleyQO.met Farley0l.met Farley02.met Farley03.met southern.met Height of lower wind instrument (m) 10.0 Height of upper wind instrument (m)

=

45.7 Wind speeds entered as meters/second Ground-level release Release height (m)

=

44.3 Building Area (m2)

.0 Effluent vertical velocity (m/s)

=

.00 Vent or stack flow (mi3/s)

=

.00 Vent or stack radius (m)

=

.00 Direction..

intake to source (deg)

=

096 Wind direction sector width (deg)

=

90 Wind direction window (deg)

=

051 - 141 Distance to intake (m)

=

46.8 Intake height (m)

=

6.9 Terrain elevation difference (m)

=

.0

00 cO 00 cO It00 C:)

C)

Co

-J z

Calculation No.; BM-03-0018-001, Version 3 Attachment D Sheet 32 of 36 TSC X/Q Estimates for Unit 2 Vent Release (4.5 years of meteorological data)

Output file names T2V245s.out T2V245s.jfd Minimum Wind Speed (m/s)

Surface roughness length (m)

Sector averaging constant Initial value of sigma y Initial value of sigma z

=

.5

.20 4.3

.00

.00 Expanded output for code testing not selected Total number of hours of data processed =

39528 Hours of missing data

=

406 Hours direction in window

=

9019 Hours elevated plume w/ dir. in window

=

0 Hours of calm winds 2112 Hours direction not in window or calm 27991 DISTRIBUTION

SUMMARY

DATA BY AVERAGING INTERVAL AVER.

PER.

UPPER LIM.

LOW LIM.

ABOVE RANGE IN RANGE BELOW RANGE ZERO TOTAL X/Qs

% NON ZERO 1

1.00E-02 1.OGE-06 0.

11131.

0.

27991.

39122.

28.45 2

1.OOE-02 1.O0E-06 0.

13685.

0, 25358.

39043.

35.05 4

1.00E-02 1.00E-06 0.

17135.

0.

21799.

38934.

44.01 a

1.00E-02 1.00E-06 0.

21376.

0.

17380.

38756.

55.16 12 1.OOE-02 1.OOE-06 0.

24315.

0.

14515.

38830.

62.62 24 1.00E-02 1.00E-06 0.

29881.

0.

8772.

38653.

77.31 96 1.00E-02

1. 00E-06 0.

37853.

0.

709.

38562.

98.16 168 1.00E-02 1.00E-06 0.

38186.

0.

103.

38289.

99.73 360 1.OOE-02 1.00E-06 0.

38121.

0.

38121.

100.00 720 1.00E-02 1.00E-06 0.

38753.

0.

38753.

100.00 95th PERCENTILE X/Q VALUES 2.07E-03 1.96E-03 1.87E-03 1.77E-03 1.46E-03 1.12E-03 6.64E-04 5.78E-04 4.60E-04 3.95E-04 95% X/Q for standard averaging intervals 0 to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 2 to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 8 to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 1 to 4 days 4 to 30 days 2.07E-03 1.67E-03 7.90E-04 5.14E-04 3.54E-04 HOURLY VALUE RANGE MAX X/Q MIN X/Q 3.85E-03 2.99E-04 2.24E-03 1.74E-04 CENTERLINE SECTOR-AVERAGE NORMAL PROGRAM COMPLETION

00 00 00 Calculation No.: BM-03-0018-001, Version 3 Attachment D Sheet 33 of 36 0)

M*

TSC X/Q Estimates for Unit 2 Reactor Release (4.5 years of meteorological data) tL6 Program

Title:

ARCON96.

0)

Developed For:

U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation U)

Division of Reactor Program Management C

Date:

June 25, 1997 11:00 a.m.

LU NRC Contacts:

J.

Y. Lee Phone: (301) 415 1080 r-:

e-mail: jyll@nrc.gov (D

J.

Hayes Phone:

(301) 415 3167 O

e-mail: jjh@nrc.gov r-L L.

A Brown Phone:

(301) 415 1232 O

e-mail; lab2@nrc.gov IJ z

Code Developer: J.

V. Ramsdell Phone:

(509) 372 6316 e-mail: jramsdell@pnl.gov code Documentation:

NUREG/CR-6331 Rev. 1 The program was prepared for an agency of the United States Government. Neither the United States Government nor any agency thereof, nor any of their employees, makes any warranty, expressed or implied, or assumes any legal liability or responsibilities for any third party's use, or the results of such use, of any portion of this program or represents that its use by such third party would not infringe privately owned rights.

Program Run 9/

7/2004 at 17:29:41

.***** ARCON INPUT *******

Number of Meteorological Data Files 5

Meteorological Data File Names FarleyO0.met Farley0l.met Farley02.met FarleyO3.met southern.met Height of lower wind instrument (m)

=

10.0 Height of upper wind instrument (m) 45.7 Wind speeds entered as meters/second Ground-level release Release height (m)

=

6.9 Building Area (m^2)

=

1078.0 Effluent vertical velocity (m/s)

=

.00 Vent or stack flow (m53/s)

.00 Vent or stack radius (m)

=

.00 Direction..

intake to source (deg) 066 wind direction sector width (deg)

=

90 Wind direction window (deg)

=

021 - 111 Distance to intake (m)

=

13.6 Intake height (m)

=

6.9 Terrain elevation difference (m)

=

.0

0O 4 -

0 (0

OD Calculation No.: BM-03-0018-001, Version 3 Attachment D Sheet 34 of 36 a,

DTSC X/Q Estimates for Unit 2 Reactor Release (4.5 years of meteorological data) a-ui 0

C:

0 I'-

C:

I.z output file names T2R245s.out T2R245s.jfd Minimum Wind Speed (m/s)

Surface roughness length (m)

Sector averaging constant Initial value of sigma y Initial value of sigma z

=

.5

=

.20 4.4 6.97 4.30 Expanded output for code testing not selected Total number of hours of data processed 39528 Hours of missing data

=

406 Hours direction in window

=

8655 Hours elevated plume w/ dir. in window

=

0 Hours of calm winds 1874 Hours direction not in window or calm

=

28593 DISTRIBUTION

SUMMARY

DATA BY AVERAGING INTERVAL AVER.

PER.

1 2

4 8

UPPER LIM.

1.00E-02 1.00E-02 1,00E-02 1.00E-02 LOW LIM.

1.00E-06 1.00E-06 1.00E-06 1.00E-06 ABOVE RANGE

0.

IN RANGE 10529.

13220.

16722.

20982.

BELOW RANGE

0.

ZERO 28593.

25823.

22212.

17774.

TOTAL X/Qs 39122.

39043.

38934.

38756.

% NON ZERO 26.91 33.86 42.95 54.14 95th PERCENTILE X/Q VALUES 5.03E-03 4.79E-03 3.96E-03 3.57E-03 95% X/Q for standard averaging intervals 0 to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 5.03E-03 2 to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 3.08E-03 8 to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 1.43E-03 1 to 4 days 1.29E-03 4 to 30 days 9.95E-04 12 1.00E-02 1.00E-06 0.

23838.

0.

14992.

38830.

61.39 24

1. OOE-02 1.00E-06 0.

28682.

0.

9971.

38653.

74.20 96

1. 00E-02 1.00E-06 0.

38105.

0.

457.

38562.

98.81 168 1.00E-02 1.OOE-06 0.

38232.

0.

57.

38289.

99.85 360 1.00E-02 1.OOE-06 0.

38121.

0.

38121.

100.00 720 1.OOE-02 1.00E-06 0.

38753.

0.

38753.

100.00 2.88E-03 2.15E-03 1.50E-03 1.33E-03 1.17E-03 1.06E-03 HOURLY VALUE RANGE MAX X/Q 5.54E-03 3.15E-03 CENTERLINE SECTOR-AVERAGE NORMAL PROGRAM COMPLETION MIN X/Q 8.41E-04 4.79E-04

cO 4-0 C'-.

0O Calculation No.: BM-03-0018-001, Version 3 Attachment D Sheet 35 of 36 TSC X/Q Estimates for Unit 2 Hatch Release (4.5 years of meteorological data) 0_

Program

Title:

ARCON96.

()

Developed For:

U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation (I)

Division of Reactor Program Management 0

O Date:

June 25, 1997 11:00 a.m.

NRC Contacts:

J.

Y. Lee Phone:

(301) 415 1080 r-e-mail: jyllenrc.gov (0

J.

J. Hayes Phone:

(301) 415 3167 O

e-mail; jjh@nrc.gov C

L.

A Brown Phone: (301) 415 1232 oi e-mail: lab2gnrc.gov C

Code Developer: J.

V. Ramsdell Phone; (509) 372 6316 e-mail: j-ramsdellOpnl.gov Code Documentation:

NUREG/CR-6331 Rev.

1 The program was prepared for an agency of the United States Government. Neither the United States Government nor any agency thereof, nor any of their employees, makes any warranty, expressed or implied, or assumes any legal liability or responsibilities for any third party's use, or the results of such use, of any portion of this program or represents that its use by such third party would not infringe privately owned rights.

Program Run 9/

8/2004 at 14:09:42

- - - - ARCON INPUT**********

Number of Meteorological Data Files

=

5 Meteorological Data File Names FarleyOO.met Farley0l.met FarleyO2.met Farley03.met southern.met Height of lower wind instrument (m)

=

10.0 Height of upper wind instrument (m)

=

45.7 Wind speeds entered as meters/second Ground-level release Release height (m)

=

2.7 Building Area (m^2)

=

1253.0 Effluent vertical velocity (m/s)

=

.00 Vent or stack flow (m^3/s)

=

.00 Vent or stack radius (m)

=

.00 Direction.. intake to source (deg) 056 Wind direction sector width (deg)

=

90 Wind direction window (deg)

=

011 -

101 Distance to intake (m) 53.9 Intake height (m)

=

6.9 Terrain elevation difference (m)

=

.0

0o 00 co 0) 0)

00 C/)

0 (0

C)

C) r-J z:

Calculation No.: BM-03-0018-001, Version 3 Attachment D Sheet 36 of 36 TSC X/Q Estimates for Unit 2 Hatch Release (4.5 years of meteorological data)

Output file names T2H245s.out T2H245s.jfd Minimum Wind Speed (m/s)

=

.5 Surface roughness length (m)

=

.20 Sector averaging constant

=

4.3 Initial value of sigma y

.00 Initial value of sigma z

.00 Expanded output for code testing not selected Total number of hours of data processed 39528 Hours of missing data

=

406 Hours direction in window 9254 Hours elevated plume w/ dir. in window 0

Hours of calm winds

=

1874 Hours direction not in window or calm

=

27994 DISTRIBUTION

SUMMARY

DATA BY AVERAGING INTERVAL AVER.

PER.

1 2

4 8

UPPER LIM.

1.00E-02 1.00E-02 1.00E-02 1.OOE-02 LOW LIM.

1.00E-06 1.OOE-06 1.00E-06 1.OOE-06 ABOVE RANGE

0.

0-

0.

IN RANGE 11128.

13908.

17483.

21818.

BELOW RANGE

0.

ZERO 27994.

25135.

21451.

16938.

TOTAL X/Qs 39122.

39043.

38934.

38756.

% NON ZERO 28.44 35.62 44.90 56.30 95th PERCENTILE X/Q VALUES 1.75E-03 1.69E-03 1.62E-03 1.53E-03 95% X/Q for standard averaging intervals 0 to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 1.75E-03 2 to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 1.46E-03 8 to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 7.01E-04 1 to 4 days 4.98E-04 4 to 30 days 3.87E-04 12 1.OOE-02 1.OOE-06 0.

24740.

0.

14090.

38830.

63.71 24 1.00-02 1.OOE-06 0.

29531.

0.

9122.

38653.

76.40 96

1. OOE-02 1.OOE-06 0.

38165.

0.

397.

38562.

98.97 168 1.OOE-02 1.0OE-06 0.

38220.

0.

69.

38289.

99.82 360 1.OOE-02 1.00E-06 0.

38121.

0.

38121.

100.00 720

1. 00E-02 1.OOE-06 0.

38753.

0.

38753.

100.00 1.27E-03 9.77E-04 6.17E-04 5.32E-04 4.34E-04 4.17E-04 HOURLY VALUE RANGE MAX X/Q MIN X1Q 2.70E-03 4.15E-04 1.57E-03 2.42E-04 CENTERLINE SECTOR-AVERAGE NORMAL PROGRAM COMPLETION

Joseph M. Farley Nuclear Plant, Units 1 and 2 License Amendment Request to Technical Specification 3.9.3, Containment Penetrations, Personnel Air Locks Doors Open During Fuel Movement Excerpt from SNC Design Bases Fuel Handling Accident Calculation

NL-07-0067, Enclosure 6, Page 1 of 27 Southern Nuclear Design Calculations MC-F-04-0058, Version 3

[ Project Calculation Number Farley Nuclear Plant SM-96-1064-001 Subject/Title Sheet Fuel Handling Accident Doses 2 of 252 Conclusions :

The fuel handling accident in the containment with the equipment hatch open, analyzed in accordance with RG 1.195 results in offsite doses which meet the acceptance criteria:

Thyroid dose (Rem)

Skin dose (Rem)

Whole body dose (Rem)

Site Boundary Low Population Zone Control Room 68.5 25.3 17.0*

0.5 0.2 0.5*

0.2 0.1

<0.1 The realistic estimate of the thyroid dose to the individual(s) designated to close the equipment hatch, with no credit for a respirator or other protective gear is 24.1 Rem (30 minutes), 48.2 Rem (1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months ) which meet the RG 1.195 limit for operators. The maximum exposure time to remain within the 50 Rem limit would be about 62 minutes.

Control room doses with the personnel airlocks open shown in Attachment 2, page 43 are more limiting, but continue to meet the acceptance criteria.

From Attachment 5, the FHA in the auxiliary building meets the acceptance criteria assuming credit for the operation of the penetration area filtration system with 0.5% bypass flow.

Thyroid dose (Rem)

Skin dose (Rem)

Whole body dose (Rem)

Site Boundary Low Population Zone 21.6 7.9 10.3 0.8 0.3 0.4 0.1 Control Room From Attachment 6, the FHA in the auxiliary building without PRF filters meets the "minimal increase" acceptance criteria IF the most recently discharged fuel in the spent fuel pool has decayed at least 676 hours0.00782 days 0.188 hours 0.00112 weeks 2.57218e-4 months since discharge from the reactor.

Site Boundary Low Population Zone Control Room Thyroid Dose (Rem) 25.7 9.5 12.3

NL-07-0067, Enclosure 6, Page 2 of 27 MC-F-04-0058, Version 2 Southern Nuclear Design Calculations Project Calculation Number Farley Nuclear Plant SM-96-1064-001 Subject/Title Sheet Fuel Handling Accident Doses 3 of 252 From Attachment 7, the FHA in the containment with the doors closed and no credit for the purge filter meets the acceptance criteria:

Site Boundary Low Population Zone Control Room Thyroid Dose (Rem) 12.1 4.5 28.6 Whole Body Dose (Rem) 0.4 0.1 0.8 Major Equations: The calculations are prepared using the TACT V computer code, installed on a NEC P90 computer. Proper operation of the program was verified by running the test problems and comparing the output to that provided in reference 10. A listing of the TACT 5 directory is included in the Attachment 1.

The equations for the numerical solutions for activity transport, and the equations for the resultant doses are described and discussed in reference 10. Equations used to derive input values for the computer analyses are explained in the body of the calculation.

Assumptions : Offsite dose analyses are prepared using dose conversion factors from ICRP 30 as provided in the users manual for TACT5 (reference 10). This is consistent with the NRC SERs for similar recent analyses (references 13 and 14).

Assumptions for the FHA in the auxiliary building or containment with the equipment hatch closed are consistent with the guidelines of Regulatory Guide 1.25, except as noted above, and are shown in Table 1. It is assumed that the air space above the spent fuel pool is maintained at a negative pressure relative to adjacent areas so that radioactive releases are processed through the penetration area filtration system--this implies that the doors and hatches into the area are closed or allow flow only into the spent fuel handling area, thereby preventing bypass of the filtration system.

Assumptions for the FHA in the containment with the equipment hatch open are consistent with the applicable portions of RG 1.195. The major differences are in the activity transport assumptions, which are dependent on the functioning of the various HVAC systems in the containment and the volume served.

In addition, control room HVAC is modeled to bound the Generic Letter 2003-001 test results. The normal intake rate is increased to 3000 (or the reftrence 29 value of 2340 cfm for Attachment 2) which includes the as-tested unfiltered inleakage. The isolation mode unfiltered inleakage is assumed to be 600 cfm plus 10 cfm for ingress/egress, and the pressurization flow is assumed to be 450 cfm with 450 cfm unfiltered inleakage plus 10 cfm for ingress/egress. Finally, the margin included in source terms is increased to bound the U2C17 core reload design as described in Al 200420003.

NL-07-0067, Enclosure 6, Page 3 of 27 Southern Nuclear Desian Calculations MC-F-04-0058. Version 2 E

Project Calculation Number Farley Nuclear Plant SM-96-1064-001 Subject/Title Sheet Fuel Handling Accident Doses 6 of 252

25. USNRC Regulatory Guide 1.195, "Methods and Assumptions for Evaluating Radiological Consequences of Design Basis Accidents at Light-Water Nuclear Power Reactors," 5/03.

26. Calculation BM-03-0018-001, "Control Room and Technical Support Center Air Intake X/Q Estimates," Version 1, 4/9/04 and Version 3, 9/15/04.

27. Technical Specifications, Section 5.5.11

28. "Fuel Handling Accident (FHA) Drop Height Reference Information," Westinghouse Letter ALA-0 1-057, June 12, 2001.

29. FNP TS Amendment 174/167, Section 5.5.18.

30. Emailfrom Michael J. Smith to Steve Berrvhill, 10/24/06, "10/23/06 CR DP" (copy attached as pp 44-46)

31. DOEJ-SM-C060059901-001, "Descriptions of Operation for Adjacent HVAC Systems to the Common Control Room Containment Personnel Air Locks Open During Refleling Outage," Version 1.

32. FNP Drawings a) D-206039, V22; D-1 76039, V20 b) D-206006, V38; D-206007, V1 7 c) D-206516, V20

33. DOEJ-SM-C060059901-002, "Design Air Flow Rates/for Unit 2 Auxiliaiy Building For Areas Between the Control Room and Containment Personnel Air Locks Open During Refjieling Outage, " Version 1.

34. U518468, Version 1, NUCON Test Report 12FARL Y1032, "Control Room Habitabiliot Tracer Gas Leak Test of the Farley Nuclear Plant, " 7/15/04

NL-07-0067, Enclosure 6, Page 4 of 27 Southern Nuclear Desian Calculations MC-F-04-0058, Version 2 L

Project Calculation Number Farley Nuclear Plant SM-96-1064-001 Subject/Title Sheet Fuel Handling Accident Doses 7 of 252 Body of Calculation:

A. Source Terms Source terms for the FHA are developed from Westinghouse input of core activities for the uprated power of 2775 MWt (reference 8), plus 2% for calorimetric errors (2831 MWt, reference 9). Parametric analyses were performed by Westinghouse to determine how these source terms might vary with different fuel designs (enrichment, blankets, burnup) to define a "bounding" factor to be applied to the base core inventory.

Because the resultant thyroid, whole body and skin doses are determined by iodines and noble gases, the parametric analyses for these isotopes were examined and the base core inventory is multiplied by 1.02 to bound the FHA source terms.

For the FHA in the containment with the equipment hatch open, the core inventory margin is increased to 1.03 (except KrO 8 is 1.15 and Xe 133 is 1.05) and the gas gap inventory is 5 % of the core inventory except Kr85 is 10% and 1131 is 8 % consistent with reference 25. The TACT V code uses Ci/MWt as input, so the core inventory is then divided by 2775 and input without decay to file FNPGAP30:

Inventory x 1.03 x 0.05 (ex Kr85 = 0.1 and 1131 = 0.08) + 2775 = Ci/MWt.

The remaining data in FNPGP195 is from file MLWRICRP.30 from reference 10a, Appendix E 2

3 3

0 WHOLEBDY SKIN THYROID HALOGENS NOBLES ELEM.

ORG.

PART.

I 131

9. 963996E-07

5. 590000E-02 0.0 0.0 1

1 I

132 8.269001E-05

3. 550000E-01 0.0 0.0 2

1 I

133

9. 219000E-06 9.110000E-02 0.0 0.0 3

1 I

134 2.228000E-04 4.1100OOE-01 0.0 0.0 4

1 0

1 1

1.100000E 06 0.0 0.0 0

0 0

2.227000E+03 3.070000E-02 0.0 0.0 0

0 2.041000E+03 1.100000E-01 0.0 0.0 0

0 2.969000E+03 8.900000E-02 0.0 0.0 0

0 3.155000E+03

1. 420000E-01 0.0 0.0 0

0 03 6.300000E 0.0 0.0 0

0 1.800000E 0.0 0.0 0

0 I

135 2.864000E-05 2.490000E-01 0.0 0.0 5

1 KR 83M 1.035000E-04 1.270000E-05 0.0 0.0 6

2 KR 85M 4.385000E-05 2.310000E-02 0.0 0.0 7

2 KR 85 2.042000E-09 3.310000E-04 0.0 0.0 8

2 0

2.784000E+03 7.860000E-02 3.100000E 04 0.0 0.0 0.0 0.0 0

0 0

1.800000E+02 0.OOOOOOE+00 0.0 0.0 0.0 0.0 0.0 0

0 0

3.897000E+02 4.970000E-02 0.0 0.0 0.0 0.0 0.0 0

0 0

2.984000E+01 4.840000E-02 0.0 0.0 0.0 0.0 0.0 0

0 0

05 0

0 0

1.100000E 03 0.0 0.0 0

0 0

NL-07-0067, Enclosure 6, Page 5 of 27 Souithern Nurlear DAsian CIlculationnq Mr-F-o44lo~R V~r.~inn 9 EProject Calculation Number Farley Nuclear Plant SM-96-1064-001 SubjecttTitle Sheet Fuel Handling Accident Doses 8 of 252 wm IV KR 87 1.519000E-04 1.330000E-01 0.0 0.0 9

2 KR 88 6.875000E-05 3.380000E-01 0.0 0.0 10 2

XE 131M 6.680000E-07 1.250000E-03 0.0 0.0 11 2

XE 133M 3.490000E-06 4.290000E-03 0.0 0.0 12 2

7. 423000E+02 3.360000E-01 0.0 0.0 0.0 0.0 0.0 0

0 0

1.058000E+03 7.760000E-02 0.0 0.0 0.0 0.0 0.0 0

0 0

1. 559000E+01 1.330000E-02 0.0 0.0 0.0 0.0 0.0 0

0 0

8. 908000E+01 2.960000E-02 0.0 0.0 0.0 0.0 0.0 0

0 0

XE 133

1. 522000E-06

4. 960000E-03 0.0 0.0 13 2

XE 135M

7. 400000E-04

6. 370000E-02 0.0 0.0 14 2

XE 135

2. 091999E-05 3.590000E-02 0.0 0.0 15 2

XE 137

2. 961000E-03

2. 830000E-02 0.0 0.0 16 2

XE 138 6.796001E-04 1.870000E-01 0.0 0.0 17 2

2. 838000E+03

9. 670000E-03 0.0 0.0 0

0 5.568000E+02

2. 140000E-02 0.0 0.0 0.0 0.0 0.0 0

0.0 0.0 0.0 0

0 0

6.495000E+02 6.320000E-02 0.0 0.0 0

0 0.0 0.0 0.0 0

2.598000E+03 4.590000E-01 0.0 0.0 0.0 0

0.0 0.0 0

0 2.413000E+03 1.470000E-01 0.0 0.0 0.0 0

0 0

0 0.0 0

0.0 0

0

NL-07-0067, Enclosure 6, Page 6 of 27 MC-F-04-0058, Version 2 Southern Nuclear Design Calculations Project Calculation Number FarSuclear Plant SM-96-1064-001 Subject/Title S

1eet Fuel Handling Accident Doses 12 of 252 C. Containment Building Model: Regulatory Guide 1.195 FHA in Containment (with Equipment Hatch Open)

For an FHA in the containment, the source terms are modeled as above in section A. The building space used for the dilution of the activity released is evaluated for the space served by the containment purge system.

During refueling, the containment may be purged using the main purge, drawing air from elevations 130 and 155 of the containment (references 7d, e, f). [NOTE that the pool sweep is typically not used, and has been disconnected from the purge exhaust system. If it were to be used, it would exhaust to elevation 139 below the operating deck.] Assuming mixing above the pool up to the elevation of the containment cooling fan header (but no mixing outside this envelope which might be induced by operation of the fans), and 90% free space, this represents a volume of approximately:

7r x 652 x 55 x 0.9;z 6.6 x 105 ft3.

An FHA in the refueling canal will release the gap activity from one assembly (reference 11, 12) with the highest power, i.e. with a peaking factor of 1.7 (reference 8, which exceeds the reference 1 value), to the pool water which will scrub the iodines as they evolve into the building air. Release factors from reference 25 are 8% 1"', 10% Kr85, and 5% for other iodines and noble gases. The original partition between elemental and organic species of iodine and the cleanup by the pool water as described in reference 25 are combined to result in 0.5% of the iodine instantaneously released to the building, partitioned as 50% organic and 50%

elemental.

Releases are driven from the containment by the purge system flow. Although the initial release concentration in the containment:

895.9 Ci Kr85 x 1 E6 lLCi/Ci - 6.6 E5 ft3 28317 cc/ft3 = 5 x 10-2 ltCi/cc is well in excess of the containment purge isolation radiation monitor setpoint of 1 x 10-4 [tCi/cc, no credit for stopping the purge flow or purge filter is modeled. Then the main purge flow plus 10% exhausts 48,500 x 1.1 cfm x 120 min / 6.6 x 105 ft3 = 9.7 times the containment volume, which exceeds the requirements of reference 25. Releases are modeled as ground level, with x/Q values taken from reference 26. The annotated input file (FHAHTCH6.in) and offsite dose results are shown below, where control room parameters are taken from section D below.

NL-07-0067, Enclosure 6, Page 7 of 27 Southern Nuclear Desian Calculations MC.P'-A-I4-Ofl*R Ver!£,nn ")

T[Project Calculation Number Farley Nuclear Plant SM-96-1064-001 Subject/Title Sheet Fuel Handling Accident Doses 13 of 252 oov FHAHTCH6.in

'c:fnpgpl95'

'c:LTAPE

','c:MTAPE

','c:NTAPE Farley FHA in the Containment with 0,

0, 1, 1, 0 2,

2

'Contnmnt','CtrlRoom' 25 1.7675E+1,

-1.00E+02 8.500E-03, 1.700E-00 0.000E+00, 0.000E+00 5.000E-01, 5.000E-01, 0.000E+00 1.OOOE+00, 0.000E+00, 0.000E+00 RG 1.195 source terms from section A above.

Equip Hatch Open (RG 1.195) 1 2775 MWt/157 assys, 100 hr decay 1.7/400 iodine and 1.7 noble gas release fractions 50% elemental, 50% organic iodine 100% elemental noble gases Containment refueling area and 6.600E+05, 1.140E+05

'TIME INTERVAL

',0,0,0,0,2,

'INITIAL FRACTION',0,0,0,0,2,

'FILTER EFF

'TRANSFER CFM

'DOSE PARAMS

'TIME INTERVAL

'FILTER EFF

'TRANSFER CFM

'DOSE PARAMS

'TIME INTERVAL

'FILTER EFF

'TRANSFER CFM

'DOSE PARAMS

'TIME INTERVAL

'TIME INTERVAL 1,1,0,1,3, 1,2,0,1,3,

,1, 3, 0, 1, 3, 0,0,0,1,3,

,0, 0, 0, 2, 3,

,0, 0, 0, 0, 5,

,0, 0, 0, 0, 2,

', 0, 0, 0, 0,2, 1, 1, 0, 1, 3, 1, 2, 0, 1, 3,

'1,3,0, 1,3,

,0, 0, 0, 1, 3,

,0, 0, 0, 2, 3,

,0, 0, 0, 0, 5,

,0, 0, 0, 0, 2,

,0,0,0,0,2,

',0,0,0,0,2,

,1,1,0,1,3,

,1,2,0,1,3,

',1,3,0,1,3,

',1,1,0,2,3,

,1,2,0,2,3,

',1,3,0,2,3,

',0,0,0,1,3,

',0,0,0,2,3,

, 0,0,0,0,5,

',0,0,0,0,2,

,0,0,0,0,2,

0. OOOE+00,

1. OOOE+00,

0. OOOE+00,

0. OOOE+00, 5.350E+04, 1 360E+03, 7.600E-04,

2. 778E-06,

2. 778E-04,

8. 333E-03, 1.250E-02,

0. OOOE+00,

5. 350E+04,

6. 100E+02, 7 600E-04, 2 500E-02, 4 167E-02,

8. 333E-02,

1. 667E-01,

0. OOOE+00,

5. 350E+04,

9. 100E+02, 7.600E-04,

3. 333E-01,

6. 666E-01,

7. 500E-01,

1. OOOE+00, 1.250E+00, 1.500E+00,

1. 750E+00, 2 7778E-06 0 OOOE+00

0. OOOE+00,

3. 470E-04,

2. 778E-04

8. 333E-03 1.250E-02

2. 500E-02

0. OOOE+00,

0. OOOE+00, 0 OOOE+00,

0. OOOE+00,

3. 470E-04,

4. 167E-02

8. 333E-02

1. 667E-01

3. 333E-01

0. OOOE+00,

0. OOOE+00, 3 470E-04,

6. 666E-01

7. 50DE-01

1. OOOE+00

1. 250E+00

1. 500E+00 1. 750E+00

2. OOOE+00 control room volumes 0.OOOE+00 0.OOOE+00 0.OOOE+00 6.434E+01 HVAC flow rates 1.000E+00 Normal CR exhaust flow 2.800E-04, 3.470E-04, 0.OOOE+00 45 sec, CR auto-isolation 0.OOOE+00 0.OOOE+00 0.OOOE+00 18.08E+00 Unpressurized CR intake 1.OOOE+00 flow and exhaust 2.800E-04, 3.470E-04, 0.OOOE+00 10 min, manual pressurization 4.871E+01 Equivalent pressurization 4.871E+01 filter efficiency 0.OOOE+00 9.450E+01 Recirculation filter 9.450E+01 efficiency 0.OOOE+00 19.52E+00 Pressurized CR intake 2.700E+03 flow and exhaust 2.800E-04, 3.470E-04, 0.OOOE+00

NL-07-0067, Enclosure 6, Page 8 of 27 Southern Nuclear Desian Calculations MC-F-O4-AO~ V~r~inn 2

[Proj ect Calculation Number Farley Nuclear Plant SM-96-1064-001 Subject/Title Sheet Fuel Handling Accident Doses 14 of 252

.. w

'TIME INTERVAL

'TRANSFER CFM

'DOSE PARAMS

'TIME INTERVAL

'TRANSFER CFM

'DOSE PARAMS

'END 0,0,0,0,2, 0,0,0,1,3, 0,0,0,0,5, 0,0,0,0,2, 0,0,0,0,2, 0,0,0,0,2,

,0, 0, 0, 0, 2,

,0,0,0,0,2, 0,0,0,0,2,

,0,0,0,0,2,

,0,0,0,1,3, 1,0,0,0,0,5, 1,0,0,0,0,0, 2 OOOE+00, 5 350E+04, 2 900E-04,

2. 250E+00, 2. 500E+00,

3. OOOE+00,

4. OOOE+00,

5. OOOE+00,

6. OOOE+00,

7. OOOE+00,

8. OOOE+00,

5. 350E+04,

3. 300E-05,

0. OOOE+00, 2 250E+00 0 OOOE+00, 3 470E-04, 2 500E+00 3 OOOE+00 4 OOOE+00

5. OOOE+00 6 OOOE+00 7 OOOE+00 8 OOOE+00 24. OOE+00

0. OOOE+00, 3.470E-04,

0. OOOE+00, 14.96E+00 1.100E-04,

7. 379E+00

1. OOOE-05,

0. OOOE+00 Change CR X/Q and flow 3.470E-04, 0.OOOE+00 Change CR X/Q and flow 3.470E-04, 0.OOOE+00

SUMMARY

OF OFF-SITE DOSES Farley FHA in the Containment with Equip Hatch Open (RG 1.195)

CALCULATION FOR WHOLEBDY DOSE (REMS)

MULTI NODE EXCLUSION RADIUS START TIME (HRS)

0. OOOE+00

2. 778E-06

2. 778E-04 8.333E-03 1. 250E-02

2. 500E-02 4.167E-02

8. 333E-02

1. 667E-01

3. 333E-01 6 666E-01

7. 500E-01

1. OOOE+00 1. 250E+00

1. 500E+00 1. 750E+00

2. OOOE+00 2. 250E+00

2. 500E+00

3. OOOE+00 4. OOOE+00

5. OOOE+00

6. OOOE+00

7. OOOE+00 8. OOOE+00 CONTAINMENT WITH ESF LOW POPULATION ZONE EACH ACCUM.

STEP EACH STEP ACCUM.

2.705E-06 2.705E-06 9.967E-07 9.967E-07 2.677E-04 2.704E-04 9.861E-05 9.961E-05 7.683E-03 7.953E-03 2.830E-03 2.930E-03 3.857E-03 1.181E-02 1.421E-03 4.351E-03 1.111E-02 2.292E-02 4.094E-03 8.445E-03 1.381E-02 3.673E-02 5.086E-03 1.353E-02 2.997E-02 6.670E-02 1.104E-02 2.457E-02 4.447E-02 1.112E-01 1.638E-02 4.096E-02 4.934E-02 1.605E-01 1.818E-02 5.914E-02 3.166E-02 1.922E-01 1.166E-02 7.080E-02 2.598E-03 1.948E-01 9.571E-04 7.176E-02 3.653E-03 1.984E-01 1.346E-03 7.310E-02 1.084E-03 1.995E-01 3.995E-04 7.350E-02 3.239E-04 1.998E-01 1.193E-04 7.362E-02 9.857E-05 1.999E-01 3.632E-05 7.366E-02 3.158E-05 2.OOOE-01 1.164E-05 7.367E-02 4.382E-06 2.OOOE-01 1.662E-06 7.367E-02 2.019E-06 2.OOOE-01 7.657E-07 7.367E-02 2.168E-06 2.OOOE-01 8.225E-07 7.367E-02 2.567E-06 2.OOOE-01 9.739E-07 7.368E-02 1.560E-06 2.OOOE-01 5.917E-07 7.368E-02 9.606E-07 2.OOOE-01 3.643E-07 7.368E-02 5.916E-07 2.OOOE-01 2.244E-07 7.368E-02 3.644E-07 2.OOOE-01 1.382E-07 7.368E-02 6.649E-08 2.OOOE-01 2.015E-08 7.368E-02 TOTAL 2.OOOE-01 TOTAL 7.368E-02

NL-07-0067, Enclosure 6, Page 9 of 27 Southern Nuclear Desion Calculation*

MC-F-04-0058 Version 2

[ Project Calculation Number Farley Nuclear Plant SM-96-1064-001 Subject/Title Sheet Fuel Handling Accident Doses 15 of 252 CALCULATION F MULTI NODE EXCLUSION RADIUS OR SKIN DOSE (REMS)

START TIME (HRS)

0. OOOE+00

2. 778E-06

2. 778E-04

8. 333E-03

1. 250E-02

2. 500E-02

4. 167E-02

8. 333E-02

1. 667E-01

3. 333E-01

6. 666E-01

7. 500E-01 1.OOOE+00 1.250E+00

1. 500E+00

1. 750E+00

2. OOOE+00 2. 250E+00

2. 500E+00

3. OOOE+00

4. OOOE+00

5. OOOE+00

6. OOOE+00

7. OOOE+00

8. OOOE+00 CONTAINMENT WITH ESF LOW POPULATION ZONE EACH ACCUM.

STEP EACH STEP ACCUM.

5.717E-06 5.717E-06 2.106E-06 2.106E-06 5.657E-04 5.714E-04 2.084E-04 2.10SE-04 1.624E-02 1.681E-02 5.982E-03 6.192E-03 8.152E-03 2.496E-02 3.004E-03 9.196E-03 2.349E-02 4.845E-02 8.653E-03 1.785E-02 2.918E-02 7.762E-02 1.075E-02 2.860E-02 6.335E-02 1.410E-01 2.334E-02 5.194E-02 9.398E-02 2.350E-01 3.463E-02 8.656E-02 1.043E-01 3.392E-01 3.842E-02 1.250E-01 6.692E-02 4.062E-01 2.466E-02 1.496E-01 5.492E-03 4.117E-01 2.023E-03 1.517E-01 7.722E-03 4.194E-01 2.845E-03 1.545E-01 2.293E-03 4.217E-01 8.446E-04 1.554E-01 6.852E-04 4.224E-01 2.524E-04 1.556E-01 2.087E-04 4.226E-01 7.690E-05 1.557E-01 6.707E-05 4.226E-01 2.471E-05 1.557E-01 9.367E-06 4.226E-01 3.553E-06 1.557E-01 4.361E-06 4.226E-01 1.654E-06 1.557E-01 4.742E-06 4.227E-01 1.799E-06 1.557E-01 5.654E-06 4.227E-01 2.145E-06 1.557E-01 3.439E-06 4.227E-01 1.304E-06 1.557E-01 2.119E-06 4.227E-01 8.036E-07 1.557E-01 1.305E-06 4.227E-01 4.951E-07 1.557E-01 8.042E-07 4.227E-01 3.051E-07 1.557E-01 1.468E-07 4.227E-01 4.450E-08 1.557E-01 TOTAL 4.227E-01 TOTAL 1.557E-01

NL-07-0067, Enclosure 6, Page 10 of 27 Southern Nuclear Desian Calculations MC-F-04-0058. Version 2 Project Calculation Number Farley Nuclear Plant SM-96-1064-001 Subject/Title Sheet Fuel Handling Accident Doses 16 of 252 w

CALCULATION FOR THYROID DOSE (REMS)

MULTI NODE CONTAINMENT WITH ESF START TIME (HRS) 0. OOOE+00 2.778E-06 2.778E-04

8. 333E-03 1.250E-02 2. 500E-02 4.167E-02

8. 333E-02

1. 667E-01

3. 333E-01

6. 666E-01 7.500E-01

1. OOOE+00 1.250E+00

1. 500E+00

1. 750E+00

2. OOE+00 2. 250E+00

2. 500E+00 3.OOOE+00

4. OOOE+00

5. OOE+00

6. OOE+00

7. OOOE+00

8. OOOE+00 EXCLUSION RADIUS EACH STEP ACCUM.

LOW POPULATION ZONE EACH ACCUM.

STEP 9.264E-04 9.264E-04 3.413E-04 3.413E-04 9.166E-02 9.258E-02 3.377E-02 3.411E-02 2.631E+00 2.723E+00 9.692E-01 1.003E+00 1.321E+00 4.044E+00 4.867E-01 1.490E+00 3.805E+00 7.850E+00 1.402E+00 2.892E+00 4.728E+00 1.258E+01 1.742E+00 4.634E+00 1.027E+01 2.284E+01 3.782E+00 8.416E+00 1.523E+01 3.807E+01 5.611E+00 1.403E+01 1.690E+01 5.498E+01 6.228E+00 2.025E+01 1.085E+01 6.583E+01 3.997E+00 2.425E+01 8.903E-01 6.672E+01 3.280E-01 2.458E+01 1.251E+00 6.797E+01 4.610E-01 2.504E+01 3.707E-01 6.834E+01 1.366E-01 2.518E+01 1.099E-01 6.845E+01 4.048E-02 2.522E+01 3.263E-02 6.848E+01 1.202E-02 2.523E+01 9.720E-03 6.849E+01 3.581E-03 2.523E+01 1.113E-03 6.849E+01 4.222E-04 2.523E+01 3.390E-04 6.849E+01 1.286E-04 2.523E+01 1.417E-04 6.849E+01 5.374E-05 2.523E+01 2.212E-05 6.849E+01 8.391E-06 2.523E+01 1.761E-06 6.849E+01 6.681E-07 2.523E+01 2.699E-07 6.849E+01 1.024E-07 2.523E+01 4.338E-08 6.849E+01 1.645E-08 2.523E+01 6.988E-09 6.849E+01 2.650E-09 2.523E+01 1.527E-10 6.849E+01 4.627E-11 2.523E+01 TOTAL 6.849E+01 TOTAL 2.523E+01 D. Control Room Model (with Equipment Hatch Open)

The control room normally operates with an air supply of 1350 cfm (ref. 17) coming from the computer room A/C. Approximately 35% of this is outside air; however for conservatism, a total air flow of 3000 cfm is assumed to be outside air. TACT5 will not allow transfer from the environment back into a node and allows only one transfer path for both filtered intake and unfiltered inleakage, so an equivalent direct transfer from the containment via the environment to the control room is modeled as:

Release Rate x X/Q x Unit Conversion Factors x Intake Rate.

NL-07-0067, Enclosure 6, Page 11 of 27 Southern Nuclear Desian Calculations MC-F-04-0058 Version 2 Project Calculation Number Farley Nuclear Plant SM-96-1064-001 Subject/Title Sheet Fuel Handling Accident Doses 17 of 252 The effective filter efficiency is modeled as:

1 - [(1-efficiency) x Filtered Flow + Unfiltered Flow] / [Filtered Flow +Unfiltered Flow].

The control room intake radiation monitor will isolate the control room HVAC, but manual action is required to start the pressurization and recirculation mode of operation. The pressurization and recirculation flows are assumed to be 450 cfm and the Technical Specification (ref. 27) value of 3000 +/- 10%. The control room filter efficiency modeled includes a reduction of 0.5% allowed for bypass testing yielding:

CR volume = 114,000 ft3 (ref. 16)

Pressurization flow = 450 cfm, efficiency = 98.5%

Recirculation flow = 2700 cfm, efficiency =24.5%

Normal flow = 3000 cfm, efficiency = 0%

Pressurized unfiltered inleakage = 450 cfm, efficiency = 0% (assumed)

For the period between automatic isolation and manual pressurization, the control room unfiltered inleakage is assumed to be 600 cfm unfiltered inleakage plus 10 cfm for ingress/egress = 600 + 10 = 610 cfm. Then for a containment purge flow rate of 53,500 cfm, the direct transfer rates and filter efficiencies are:

Release flow (cfm)

X/Q Unit Conversion Factors (s/m 3)

(min/s) (m3/ft 3)

Intake flow (cfm) 53,500 8.49E-4 1l 6.51 E-4 3.21 E-4 1/60I 1/35.3 3000 610 910 Direct transfer (cfm) 64.34 13.08 19.52 14.96 7.379 Effective Filter Efficiency 0

0 1-[450 (1-0.985) + 450 +10]/(910) = 48.71%

-1 (Checker's note: The X/Q values used are slightly different from reference 26, but give conservative results.) (Original initialed by DJC 5-2 7-04)

The switch from normal HVAC intake (3000 cfm) to unpressurized unfiltered inleakage (600 cfm) occurs when the intake radiation monitor responds to the released activity. The 1 E-4 iiCi/ml monitor setpoint (ref.

15) is exceeded within about 1 second of the activity reaching the monitor:

Released activity = 1.21 0E6 ýtCi X/Q = 8.49E-4 (s/mi3)

(1.210 E6 ýLCi / 1 sec) x (8.49E-4 s/m3) / (1E6 mIl/m 3) Z 1 E-3 ýtCi/mnl Based on a review of reference 20, this should result in an isolation signal within about 15-20 seconds, so use 45 seconds to conservatively bound the AOV (HV3622-3629) stroke time.

NL-07-0067, Enclosure 6, Page 12 of 27 Southern Nuclear Desian Calculations MC-F-04-0058, Version 2 Project Calculation Number Farley Nuclear Plant SM-96-1064-001 Subject/Title Sheet Fuel Handling Accident Doses 18 of 252

.. v The resultant control room activities are entered into an EXCEL spreadsheet to calculate the control room thyroid, skin and whole body doses. Thyroid doses are calculated for each time step from (Average Ci-hr) x (3600 s/hr) x (35.3 ft 3/m 3) x (3.47E-4 m3 inhaled/s) x DCF (REM/Ci)/1 14,000 ft3 and whole body and skin doses are calculated from (I/Geometry factor) x (Average Ci-hr) x (35.3 ft3/m 3) x DCF (REM-m 3/Ci-hr)/1 14,000 ft 3 where the geometry factor is applied only to the whole body dose and is 1173/ 1140000.338 = 22.91 (ref 25).

FHAHTCH6.out Control Room (Curies)

Inventory ISOTOPE 1 131 1131 1 132 1132 1 133 1133 1 134 1134 1 135 1135 KR 83M KR 85M KR 85 KR 87 KR 88 XE 131M XE 133M XE 133 XE 135 Time (hr) 2.78E-06

1. 90E-06

2. 94E-19

1. 31E-07

1. 31E-07 0. OOE+00 0. OOE+00 1 13E-10

1. 13E-10

5. 78E-21

2. 65E-11 1.4 6E-05

6. 46E-28

9. 21E-15

5. 98E-06

1. 24E-05

8. 01E-04 1. 70E-07 8.33E-03 1.25E-02 2.50E-02 4.17E-02 8.33E-02

5. 57E-03 5. 57E-03 8.59E-16

8. 59E-16

3. 84E-04

0. OOE+00

3. 31E-07

1. 69E-17 7 76E-08

4. 27E-02

1. 89E-24

2. 69E-11

1. 75E-02 3 63E-02

2. 35E+00 4.98E-04 8.25E-03 8.25E-03 1.27E-15 1.27E-15 5.70E-04 5.70E-04

0. OOE+00

4. 91E-07

2. 50E-17

1. 15E-07

6. 33E-02

2. 79E-24

3. 99E-11

2. 60E-02

5. 38E-02

3. 48E+00 7.38E-04

9. 80E-03

1. 51E-15

6. 77E-04

6. 77E-04 0. OOE+00

0. OOE+00

5. 82E-07

2. 96E-17

1. 36E-07

7. 52E-02

3. 29E-24 4.73E-11

3. 09E-02

6. 39E-02

4. 13E+00

8. 76E-04 1 17E-02 1 17E-02 1 79E-15 1 79E-15 8 08E-04 8 08E-04 o.OOE+00 o.OOE+00

6. 95E-07

3. 51E-17

1. 63E-07

8. 98E-02

3. 90E-24

5. 62E-11

3. 69E-02

7. 63E-02

4. 94E+00

1. 05E-03 1.58E-02 1. 58E-02 2.39E-15 2.39E-15

1. 09E-03

1. 09E-03 0. OOE+00

0. OOE+00

9. 34E-07

4. 67E-17

2. 18E-07 1.21E-01

5. 14E-24

7. 51E-II 4.98E-02 1.03E-01

6. 67E+00

1. 41E-03

MC-F-04-0058, Version 2 Plant:

Farley Nuclear Plant

Title:

Fuel Handling Accident Doses I

Southern Nuclear Design Calculations Unit:

Calculation Number:

01 E32 21&

&2 SM-96-1064-001 n

Sheet:

19 Control Room Inventory (Curies) 1.67E-01 3.33E-01 6.67E-01 7.50E-01 1.OOE+00 1.25 1.5 1.75 2

2.25 2.5 3

2. 16E-02

3. 19E-15

3. 19E-15 1 49E-03 1 49E-03 0.OE+00

0. OOE+00 1.27E-06 1.27E-06

6. 20E-17

2. 95E-07

1. 66E-01

6. 72E-24 1

1OE-10 6.82E-02

1. 41E-01

9. 13E+00

1. 91E-03 2.05E-02 2.05E-02 2.88E-15

2. 88E-15 1. 40E-03

1. 40E-03

0. OOE+00 0. OOE+00 1.18E-06 1.18E-06 8.08E-17

3. 98E-07

2. 30E-01

8. 52E-24

1. 34E-10 9.46E-02 1. 95E-01

1. 27E+01

2. 62E-03

1. 36E-02

1. 73E-15

9. 18E-04 9 18E-04 0. OOE+00 0. OOE+00

7. 56E-07

7. 54E-17 3. 99E-07 2 43E-01

7. 50E-24

1. 31E-10

9. 98E-02 2. 05E-01

1. 33E+01

2. 70E-03

1. 19E-02

1. 48E-15

8. 04E-04 8.04E-04 0. OOE+00 0. OOE+00

6. 58E-07

6. 58E-07 7.15E-17

3. 85E-07 2.38E-01

7. 01E-24 1.25E-10

9. 76E-02

2. OOE-01

1. 30E+01

2. 63E-03

7. 85E-03

9. 05E-16

5. 26E-04

5. 26E-04 0. OOE+00 0. OOE+00 4.23E-07 4.23E-07

5. 93E-17

3. 37E-07

2. 17E-01

5. 56E-24 1.07E-10 8.88E-02 1.82E-01

1. 19E+01

2. 35E-03 5.06E-03 5.06E-03

5. 43E-16 5.43E-16

3. 37E-04

3. 37E-04 0. OOE+00 0. OOE+00 2. 66E-07 2.66E-07 4.83E-17 2.90E-07

1. 94E-01

4. 34E-24

9. OOE-11 7.94E-02 1. 62E-01

1. 06E+01 2. 06E-03

3. 24E-03 3.24E-03 3.22E-16 3.22E-16 2.14E-04

2. 14E-04 0. OOE+00

0. OOE+00

1. 66E-07

3. 92E-17 2.48E-07 1.73E-01

3. 37E-24

7. 53E-11 7. 06E-02 1.44E-01

9. 41E+00 1.80E-03

2. 06E-03

1. 90E-16 1.90E-16 1. 35E-04

1. 35E-04 0. OOE+00 0. OOE+00 1. 03E-07

1. 03E-07

3. 17E-17 2.12E-07 1. 53E-01 2 61E-24

6. 28E-11

6. 27E-02 1.27E-01

8. 35E+00 1. 57E-03

1. 31E-03

1. 12E-16

8. 50E-05

8. 50E-05 0. OOE+00 0. OOE+00

6. 38E-08

2. 56E-17

1. 81E-07

1. 36E-01

2. 02E-24

5. 24E-11 5.56E-02 1.13E-01 7.40E+00 1.37E-03 8.28E-04 8.28E-04

6. 61E-17

5. 35E-05

5. 35E-05 0. OOE+00 0. OOE+00 3. 94E-08 3.94E-08

2. 07E-17 1.54E-07 1.21E-01

1. 56E-24

4. 37E-11

4. 93E-02

9. 97E-02

6. 55E+00

1. 19E-03 5.25E-04 5.25E-04

3. 89E-17

3. 89E-17 3.37E-05 3.37E-05 0. OOE+00 0. OOE+00 2.44E-08 2.44E-08

1. 67E-17 1.31E-07

1. 07E-01 1.21E-24

3. 65E-11

4. 37E-02

8. 82E-02

5. 81E+00

1. 04E-03

2. 11E-04

1. 35E-17 1.35E-17 1.33E-05

1. 33E-05

0. OOE+00

9. 31E-09

1. 09E-17

9. 55E-08 8.42E-02

7. 24E-25

2. 54E-11

3. 44E-02

6. 90E-02 4.56E+00 7.84E-04 MC-CCN Calculation Cover Sheet, Rev. 0 11/19/03

MC-F-04-0058, Version 2 SPlant:

Farley Nuclear Plant

Title:

Fuel Handling Accident Doses Southern Nuclear Design Calculations Unit:

Calculation Number:

0 1 012 Z I & 2 SM-96-1064-001 I

Sheet:

20 (0~

U,0 Cw Control Room Inventory (Curies)

Exposure (Ci - hr)

Time (hr) 4 5

6 7

8 2.78E-06 8.33E-03

3. 40E-05

1. 62E-18

1. 62E-18 2.08E-06

2. 08E-06 0 OOE+00 0 OOE+00 1 36E-09

1. 36E-09

4. 67E-18

5. 05E-08

5. 22E-02

2. 60E-25

1. 23E-11

2. 12E-02 4.22E-02 2.81E+00 4.51E-04

5. 47E-06 5.47E-06

1. 95E-19

3. 26E-07 3.26E-07

0. OOE+00 0. OOE+00

1. 98E-10'

1. 98E-10

1. 99E-18 2. 67E-08 3. 23E-02

9. 31E-26

5. 93E-12

1. 31E-02

2. 58E-02 1.73E+00

2. 59E-04

8. 82E-07

8. 82E-07 2.34E-20

2. 34E-20

5. 1OE-08

0. OOE+00

0. OOE+00 2.89E-11 2.89E-11 8.50E-19

1. 41E-08

2. OOE-02

3. 34E-26 2. 87E-12 8 11E-03 1. 58E-02 1. 07E+00

1. 49E-04 1.42E-07 1. 42E-07

2. 81E-21

2. 81E-21 7. 98E-09 7.98E-09 0. OOE+00 0. OOE+00

4. 22E-12 4.22E-12

3. 63E-19 7 48E-09

1. 24E-02 1.20E-26

1. 39E-12

5. O1E-03

9. 65E-03

6. 56E-01 8. 54E-05

2. 29E-08 2.29E-08

3. 37E-22

1. 25E-09 1.25E-09 0. OOE+00 0. OOE+00

6. 15E-13

1. 55E-19

3. 96E-09

7. 68E-03 4.29E-27

6. 71E-13

3. 1OE-03

5. 91E-03 4.04E-01

4. 91E-05 I

I I

131 132 133 134 135 KR 83M KR 85M KR 85 KR 87 KR 88 XE131M XE133M XE 133 XE 135 5.28E-12 8.16E-25 3.64E-1 3 O.OOE+00 3.14E-16 8.03E-27 3.68E-17 2.02E-1 1 8.97E-34 1.28E-20 8.31E-12 1.72E-11 1.11E-09 2.36E-13 4.64E-05 7.16E-18 3.20E-06 O.OOE+00 2.76E-09 7.04E-20 3.23E-1 0 1.78E-04 7.85E-27 1.12E-13 7.31 E-05 1.51 E-04 9.78E-03 2.07E-06 1.25E-02 5.76E-05 8.88E-18 3.98E-06 O.OOE+00 3.43E-09 8.73E-20 4.01 E-10 2.21 E-04 9.74E-27 1.39E-13 9.07E-05 1.88E-04 1.21 E-02 2.58E-06 MC-CCN Calculation Cover Sheet, Rev. 0 11/19/03

MC-F-04-0058, Version 2 Southern Nuclear Design Calculations Plant:

Unit:

Calculation Number:

Farley Nuclear Plant 0 1 03 2 RI 1 & 2 SM-96-1064-001

Title:

Fuel Handling Accident Doses Sheet:

21 Exposure (Ci - hr) 2.50E-02 4.17E-02 8.33E-02 1.67E-01 3.33E-01 6.67E-01 7.50E-01 1.OOE+00 1.25E+00 1.50E+00 1.75E+00 2.OOE+00 2.26E-04 3.59E-04 1.15E-03 3.12E-03 7.02E-03 1.14E-02 2.12E-03 4.94E-03 3.23E-03 2.07E-03 1.32E-03 8.41E-04 3.47E-17 5.50E-17 1.74E-16 4.65E-16 1.01E-15 1.53E-15 2.67E-16 5.96E-16 3.62E-16 2.16E-16 1.28E-16 7.56E-17 1.56E-05 2.47E-05 7.90E-05 2.15E-04 4.81E-04 7.74E-04 1.44E-04 3.33E-04 2.16E-04 1.38E-04 8.72E-05 5.50E-05 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 O.0OE+00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 O.OOE+00 O.OQE+00 O.QOE+00 1.34E-08 2.13E-08 6.78E-08 1.84E-07 4.08E-07 6.46E-07 1.18E-07 2.70E-07 1.72E-07 1.08E-07 6.73E-08 4.17E-08 3.41E-19 5.39E-19 1.70E-18 4.53E-18 1.19E-17 2.60E-17 6.12E-18 1.63E-17 1.35E-17 1.09E-17 8.86E-18 7.16E-18 1.57E-09 2.49E-09 7.92E-09 2.14E-08 5.77E-08 1.33E-07 3.27E-08 9.03E-08 7.84E-08 6.72E-08 5.74E-08 4.90E-08 8.65E-04 1.37E-03 4.40E-03 1.20E-02 3.30E-02 7.89E-02 2.01E-02 5.68E-02 5.13E-02 4.58E-02 4.07E-02 3.61E-02 3.80E-26 5.99E-26 1.88E-25 4.94E-25 1.27E-24 2.67E-24 6.05E-25 1.57E-24 1.24E-24 9.64E-25 7.48E-25 5.79E-25 5.45E-13 8.63E-13 2.74E-12 7.33E-12 1.96E-11 4.41E-11 1.07E-11 2.90E-11 2.46E-11 2.07E-11 1.73E-11 1.44E-11 3.56E-04 5.65E-04 1.81E-03 4.92E-03 1.36E-02 3.24E-02 8.23E-03 2.33E-02 2.10E-02 1.88E-02 1.67E-02 1.48E-02 7.36E-04 1.17E-03 3.74E-03 1.02E-02 2.80E-02 6.67E-02 1.69E-02 4.78E-02 4.30E-02 3.83E-02 3.39E-02 3.OOE-02 4.76E-02 7.56E-02 2.42E-01 6.58E-01 1.81E+00 4.33E+00 1.10E+00 3.11E+00 2.81E+00 2.50E+00 2.22E+00 1.97E+00 1.01E-05 1.60E-05 5.11E-05 1.38E-04 3.78E-04 8.87E-04 2.22E-04 6.22E-04 5.51E-04 4.82E-04 4.21E-04 3.67E-04 MC-CCN Calculation Cover Sheet, Rev. 0 11/19/03

MC-F-04-0058, Version 2 Southern Nuclear Design Calculations Plant:

Unit:

Calculation Number:

Farley Nuclear Plant E0 1 0 2 01 & 2 SM-96-1064-001

Title:

Fuel Handling Accident Doses Sheet:

CoU)

CL) 0 Co

-Jz Sum 0-2 hr (Ci inhaled) 1.46E-05 1.91E-18 9.94E-07 0.00E+00 8.22E-10 (Ci-seC/m3) 1.20E-16 6.68E-07 4.26E-01 1.16E-23 2.14E-10 1.75E-01 3.58E-01 2.33E+01 4.63E-03 Exposure (Ci - hr) 2.25E+00 2.50E+00 3.OOE+00 4.00E+00 5.00E+00 6.OOE+00 7.OOE+00 8.OOE+00 5.34E-04 4.46E-1 7 3.46E-05 0.00E+00 2.58E-08 5.79E-18 4.18E-08 3.21 E-02 4.48E-25 1.20E-11 1.31 E-02 2.65E-02 1.74E+00 3.19E-04 3.38E-04 2.63E-17 2.18E-05 0.OOE+00 1.60E-08 4.68E-1 8 3.57E-08 2.85E-02 3.47E-25 1.00E-1 1 1.16E-02 2.35E-02 1.55E+00 2.78E-04 3.68E-04 2.62E-17 2.35E-05 0.OOE+00 1.68E-08 6.92E-18 5.67E-08 4.78E-02 4.84E-25 1.55E-11 1.95E-02 3.93E-02 2.59E+00 4.55E-04 2.45E-04 1.51 E-17 1.54E-05 0.00E+00 1.07E-08 7.80E-18 7.30E-08 6.82E-02 4.92E-25 1.88E-11 2.78E-02 5.56E-02 3.68E+00 6.17E-04 3.94E-05 1.82E-18 2.41 E-06 0.OOE+00 1.56E-09 3.33E-18 3.86E-08 4.22E-02 1.76E-25 9.10E-12 1.72E-02 3.40E-02 2.27E+00 3.55E-04 6.35E-06 2.18E-19 3.77E-07 0.OOE+00 2.27E-10 1.42E-18 2.04E-08 2.62E-02 6.32E-26 4.40E-12 1.06E-02 2.08E-02 1.40E+00 2.04E-04 1.02E-06 2.62E-20 5.90E-08 0.OOE+00 3.31 E-11 6.06E-19 1.08E-08 1.62E-02 2.27E-26 2.13E-12 6.56E-03 1.27E-02 8.61 E-01 1.17E-04 1.65E-07 3.14E-21 9.23E-09 0.00E+00 4.83E-12 2.59E-19 5.72E-09 1.00E-02 8.13E-27 1.03E-12 4.05E-03 7.78E-03 5.30E-01 6.73E-05 Sum 2-8 hr (Ci inhaled) 5.92E-07 4.42E-20 3.80E-08 0.OOE+00 2.75E-1 1 (Ci-sec/m 3) 3.43E-1 7 3.15E-07 3.02E-01 2.27E-24 8.13E-11 1.23E-01 2.45E-01 1.63E+01 2.69E-03 MC-CCN Calculation Cover Sheet, Rev. 0 11/19/03

MC-F-04-0058, Version 2 Southern Nuclear Design Calculations Plant:

Unit:

Calculation Number:

Farley Nuclear Plant 0 1 0 2 11 1 & 2 SM-96-1064-001

Title:

Fuel Handling Accident Doses Sheet:

23 co 0

C C) 9 I-9 Dose Results 131 132 133 134 135 Thyroid DCF (Rem/Ci) 1.10E+06 6.30E+03 1.80E+05 1.10E+03 3.10E+04 Thyroid Dose (REM) 0 - 2 hr 1.61 E+01 1.20E-14 1.79E-01 O.OOE+00 2.55E-05 Thyroid Dose (REM) 2 - 8 hr 6.51 E-01 2.78E-16 6.84E-03 0.00E+00 8.52E-07 Exposure 0-720 hrs (Ci-Sec/M3) 4.39E-02 5.63E-1 5 2.97E-03 0.OOE+00 2.45E-06 Skin DCF (REM-M3/

Sec-Ci) 3.07E-02 1.10E-01 8.90E-02 1.42E-01 7.86E-02 Skin Dose (REM) 1.35E-03 6.19E-16 2.65E-04 0.OOE+00 1.92E-07 Body DCF (REM-M3/

Sec-Ci) 5.59E-02 3.55E-01 9.11 E-02 4.11E-01 2.49E-01 Body Dose (REM) 1.07E-04 8.72E-1 7 1.18E-05 0.OOE+00 2.66E-08 TOTAL 1.63E+01 6.58E-01 TOTAL 1.61E-03 TOTAL 1.19E-04 KR 83M KR 85M KR 85 KR 87 KR 88 XE131M XE133M XE133 XE135M Skin DCF (REM-M3/

Sec-Ci) 0.OOE+00 4.97E-02 4.84E-02 3.36E-01 7.76E-02 1.33E-02 2.96E-02 9.67E-03 2.14E-02 TOTAL NG Skin Dose (REM) 0.OOE+00 4.89E-08 3.52E-02 4.68E-24 2.29E-1 1 3.96E-03 1.78E-02 3.83E-01 1.57E-04 4.40E-01 (REM-M3/

Sec-Ci) 1.27E-05 2.31 E-02 3.31 E-04 1.33E-01 3.38E-01 1.25E-03 4.29E-03 4.96E-03 6.37E-02 TOTAL 8.58E-23 9.92E-10 1.05E-05 8.08E-26 4.36E-12 1.62E-05 1.13E-04 8.57E-03 2.03E-05 8.73E-03 Body DCF NG Body Dose (REM) 4.42E-01 (w/ iodine) 8.85E-03 (w/ iodine)

MC-CCN Calculation Cover Shect, Rev. 0 11/19/03

NL-07-0067, Enclosure 6, Page 18 of 27 MC-F-04-0058, Version 3 Southern Nuclear Design Calculations Plant:

Unit:

Calculation Number:

Farley Nuclear Plant 01 0 2 01 1 & 2 SM-96-1064-001

Title:

Fuel Handling Accident Doses Sheet:

24 E. Realistic Dose Estimate for the Individual Designated to Close the Equipment Hatch As noted in reference 23 and supporting documentation, the thyroid dose to the individual(s) designated to close the equipment hatch may be excessively high if he were to remain in the containment with design basis FHA parameters. Thus a realistic estimate of the thyroid (and whole body and skin) dose to this individual is made based on the following parameters.

Power - reference 28 indicates that fuel drops from up to 20 feet are not expected to cause damage to a fuel assembly that result in radioactive releases. The FSAR (ref. 1 la) indicates only the outer row of fuel pins might be assumed to fail, so the power is taken to be 2775 MWt / 157 assy / 264 rods x 17 rods = 1.138 MWt.

Decay Time - historically FNP has started fuel movement at about 150 hours0.00174 days 0.0417 hours 2.480159e-4 weeks 5.7075e-5 months due to limitations in spent fuel pool heat removal capability. To bound future improvements, the accident is assumed at 120 hrs.

Release Fraction - the average peaking factor for the core is 1.0. The highest power assemblies are expected to have a peaking factor of approximately 1.3. Per ref 23 the pool DF for elemental iodine is expected to be 500. Ref 11 e indicates the expected gap fraction for 1131 (by far the limiting isotope) is 2.2% as compared to 8% used in input file FNPGAP3 1 (earlier version of FNPGP 195 with a constant core inventory margin of 1.02). With a gap activity 99.75% elemental and.25% organic the expected iodine release fraction to the containment air will be (.9975/500 +.0025) x 2.2/8 = 1.236 x 10-3.

Although reference 11 e indicates the expected gap fraction for noble gases (except Kr85) is less than the noble gas release modeled, it is unchanged for this evaluation.

Containment Volume - Operation of the containment cooling system will quickly dilute the released activity throughout the free volume of the containment, so use 2 x 106 ft3. For the whole body dose, the individual is assumed exposed to the free volume above the operating deck (2 x 106 ft3) x (0.822),

which results in a Geometry Factor of 1173 / [(2 x 106)(0.822)](0.338) = 9.3.

Purge Rate - a slower purge will maintain the released activity inside the containment longer. Use of the slow speed exhaust rate will maximize the operator dose. Note however, the containment purge radiation monitor will still rapidly reach the isolation setpoint, thus purge will be rapidly terminated and the operator dose will be calculated based on the initial activity released into the containment.

ACTIVITY RELEASED TO ENVIRONMENT AND IN EACH NODE AT END OF...

2.778E-06 (HRS)

ISO NAM ENV.

Contnmnt CtrlRoom I

131 1.887E-06 8.235E-01 9.950E-10 1131 is clearly the limiting I

131 2.363E-06 1.031E+00 1.246E-09 isotope.

The total 1T"'

I 132 8.056E-22 3.515E-16 4.248E-25 inventory released into the I

132 1.009E-21 4.402E-16 5.319E-25 containment is 1.9 Ci.

I 133 6.967E-08 3.040E-02 3.673E-11 I

133 8.724E-08 3.807E-02 4.600E-11 (Note power used here was I

135 1.480E-11 6.456E-06 7.801E-15 1.04 vs 1.138 above and is I

135 1.853E-11 8.085E-06 9.769E-15 corrected on page 25)

MC-CCN Calculation Cover Sheet, Rev. 0 11/19/03

NL-07-0067, Enclosure 6, Page 19 of 27 MC-F-04-0058, Version 3 Southern Nuclear Design Calculations

[Plant:

Unit:

Calculation Number:

Farley Nuclear Plant 0 1 0 2 Z] I & 2 SM-96-1064-001

Title:

Fuel Handling Accident Doses ISheet:

I.

25 KR KR KR KR KR XE XE XE XE 83M 85M 85 87 88 131M 133M 133 135

1. 629E-23 5.409E-12

6. 327E-05

5. 578E-32

3. 131E-16

2. 761E-05 4 697E-05 3.402E-03 1.810E-07 7.107E-18 2.360E-06 2.761E+01 2.434E-26

1. 366E-10

1. 205E+01 2.049E+01 1.485E+03 7.900E-02 8.588E-27 2.852E-15 3.336E-08 0.000E+00 1.651E-19 1.456E-08 2.476E-08

1. 794E-06 9.546E-1i 27.6 E6 pCi/2 E6 ft 3/28317 ml/ft 3 yields 4.9 E-4 pCi/ml which is greater than the containment purge radiation monitor setpoint; thus purge would be expected to automatically isolate.NOTE, use of Ref. lle would increase the activity resulting in a faster response time.

Then the operator thyroid dose would be (1.138/1.04)1.855 Ci / (2 x 106/35.3) m3 x 3.47 x 10-4 m3/second x 1.08 x 106 REM/Ci inhaled

= 1.34 x 10-2 REM per second of exposure.

Then the 30 minute dose would be 1.34 x 10-2 REM/secondx 1800 seconds = 24.1 REM (48.2 Rem for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months ), and for a 50 REM limit the maximum stay time would be 3700 seconds or 62 minutes. If a high peaking factor assembly is dropped, the one hour dose could exceed 50 REM thyroid and compensatory measures to protect the closure crew would be implemented as directed by HP.

Skin and whole body doses (calculated similarly for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , shown below) clearly are not limiting.

ISOTOPE Activity Conc.

Skin DCF Skin Dose WB DCF WB Dose (Ci)

(Ci/rm3) 1(R-m3/s-Ci)

(R)

(R-m3/s-Ci)

(R) 1-131 1.90E+00 {3.35E-05 3.07E-02 3.70E-03 I 5.59E-02

.7-5-0-4 Kr-85 2.76E+01 4.87E-04 4.84E-02 8.48E-02 3.31E-04 6.67E-05 Xe-131m 1.21E+01 2.13E-04 1.33E-02 1.02E-02 1.25E-03 1.10E-04 Xe-133m 2.05E+01 3.62E-04 2.96E-02 3.85E-02 4.29E-03 6.42E-04 Xe-133 1.49E+03 2.62E-02 9.67E-03 9.12E-01 4.96E-03 5.38E-02 TOTAL 1.05E+00 5.53E-02 Assuming all activity is released during the transit time and an average transit X/O equal to the bounding control room value (1.81 x 10-3, Ref 26), the transit doses are also not limiting. NOTE that the operator cannot be exposed to 100 % of both the inside containment and outside containment activities since each is 100% of the release.

(Ci )

(R-m3/s-Ci)

(R)

(R-m3/s-Ci)

(R)

(R) 1-131 Kr-85 Xe-131m Xe-133m Xe-1 33 1.90E+00 2.76E+01 1.21E+01 2.05E+01 1.49E+03 3.07E-02 4.84E-02 1.33E-02 2.96E-02 9.67E-03 TOTAL 0.0001056 0.0024179 0.0002913 0.0010983 0.0259915 0.0299045 5.59E-02 3.31 E-04 1.25E-03 4.29E-03 4.96E-03 0.0001922 1.312666 1.654E-05 2.738E-05 0.0001592 0.0133317 0.0137271 1.312666 MC-CCN Calculation Cover Sheet, Rev. 0 11/19/03

MC-F-04-0058, Version 2 Plant:

Farley Nuclear Plant

Title:

Fuel Handling Accident Doses Unit:

I-1 5-2 1i-1&

NL-07-0067, Enclosure 6, Page 20 of 27 Southern Nuclear Design Calculations Calculation Number:

2 SM-96-1064-001 Sheet:

29 ATTACHMENT 2 FHA in the Containment with Personnel Airlocks Locks Open With the personnel airlock door(s) open the potential exists for activity released in case of an FHA in the containment to be exhausted through the auxiliary building. This represents a different release pathway than examined in part D above, but has minimal impact on offsite doses since the distance from the equipment hatch to the site boundary or LPZ is approximately the same as the distance from the vent stack or auxiliary building to these offsite locations. To support a technical specification change to allow the airlock door(s) to remain open during fuel movement, two control room dose cases are examined:

1. Activity release through the auxiliary building to the environment without auxiliary building leakage directly into the control room.

2. Activity release through the auxiliary building to the environment coupled with unfiltered leakage from the auxiliary building directly into the control room.

It should be noted that walkdowns of the (accessible) rooms and areas directly outside the airlocks which also bound the control room found no penetrations directly connecting these areas to the control room, and the walls separating them from the control room are concrete sealed and coated on both sides. The control room is also maintained at a positive pressure with respect to adjacent areas for both normal and emergency operations (Ref. 30), so any leakage which may occur would be out of the control room. Thus no inleakage from these areas is expected and case 2 is prepared to demonstrate the robustness of the design and defense-in-depth.

Case 1 - No direct leakage from the auxiliary building into the control room This case is nearly identical to that examined in part D of the calculation (pp 16-23), except the release point is from the containment or the plant vent and the normal air supply is limited to the TS value of 2340 cfm (Ref. 29). Thus only the X/Q and equivalent direct transfer to the control room change. The X/Qs used are a bounding composite of the containment ("reactor" in reference 26) and vent taken from reference 26 (version 3):

Release flow (cfm) 53,500I X/Q Unit Conversion Intake Direct Factors flow transfer (s/m 3)

(min/s)

(m 3/ft 3)

(cfm)

(cfm) 5.06E-3 1/60 1/35.3 2350 300.4 1.66E-3 610 25.58 I910 38.16 1.38E-3 31.72 7.20E-4 v

I, 16.55 Effective Filter Efficiency 0

0 1-[450 (1-0.985) + 450 +10]/(910) = 48.71%

4, MC-CCN Calculation Cover Sheet, Rev. 0 11/19/03

NL-07-0067, Enclosure 6, Page 21 of 27 Southern Nuclear Design Calculations MC-F-04-0058, Version 2 Plant:

Farley Nuclear Plant

Title:

Fuel Handling Accident Doses Unit:

j Calculation Number:

0l1 012 0]1&2 SM-96-1064-001 et:S h eet:

30 The control room radiation monitor will detect the released activity within about 1 second:

ACTIVITY RELEASED TO ENVIRONMENT AND IN EACH NODE AT END OF...

2.778E-04 (HRS)

ISO I

I I

I KR KR KR KR KR XE XE XE XE NAM 131 131 132 132 133 133 135 135 83M 85M 85 87 88 131M 133M 133 135 ENV.

1. 578E-01

1. 578E-01 2.442E-14

2. 442E-14

1. 090E-02

9. 399E-06

9. 399E-06 4.804E-16 2.204E-06 1.210E+00

5. 368E-23 7.654E-10 4.973E-01

1. 029E+00

6. 657E+01

1. 413E-02 Contnmnt 1.167E+02 1.167E+02 1.806E-11 1.806E-11 8.061E+00

8. 061E+00

6. 951E-03

3. 553E-13

1. 630E-03

8. 947E+02

3. 970E-20

5. 661E-07

3. 678E+02

7. 609E+02 4.923E+04 1.045E+01 CtrlRoom 8.858E-04

8. 858E-04 1.371E-16 1.371E-16

6. 119E-05

5. 276E-08

2. 697E-18 1.237E-08

6. 791E-03

3. 013E-25 4.297E-12

2. 792E-03

5. 776E-03

3. 737E-01

7. 930E-05 The noble gases released to the environment times X/Q yield a concentration of 68 Ci/sec x 5.06 x 10-3 sec/m 3 = 0.3 /Ci/cc, which will result in about a 1 second detection time (Ref. 20). With an assumed damper stroke time of 29 seconds, the isolation of the normal system will be complete at 30 seconds (8.33 E-3 hr).

Then the TACT 5 input file PERLOK1T.in (with FHAHTCH6.in annotated changes) is shown below:

MC-CCN Calculation Cover Sheet, Rev. 0 11/19/03

MC-F-04-0058, Version 2 NL-07-0067, Enclosure 6, Page 22 of 27 Southern Nuclear Design Calculations Calculation Number:

2 SM-96-1064-001 Plant:

Farley Nuclear Plant

Title:

Fuel Handling Accident Doses Unit:

101 02 -ll&;

Sheet:

31 PERLOK1T.in

'c:fnpgpl95

'c:LTAPE

','c:MTAPE

','c:NTAPE Farley FHA in Containment with Personnel Airlock(s) Open (RG 1.195)

I 0,

0, 1, 1, 0 2,

2

'Contnmnt','CtrlRoom' 25 1.7675E+l,

-1.OOE+02 8.500E-03, 1.700E-00 0.OOOE+00, 0.000E+00 5.000E-01, 5.OOOE-01, 0.OOOE+00 1.OOOE+00, 0.OOOE+00, 0.OOOE+00 6.600E+05, 1.

'TIME INTERVAL

'INITIAL FRACT:

'FILTER EFF

'TRANSFER CFM

'DOSE PARAMS 0.OOOE+00

'TIME INTERVAL

'FILTER EFF

'TRANSFER CFM

'DOSE PARAMS 0.OOOE+00

'TIME INTERVAL

'FILTER EFF

'TRANSFER CFM

'DOSE PARAMS

0. OOOE+00

'TIME INTERVAL

'TIME INTERVAL 140E+05

',0,0,0,0,2, ION',0,0,0,0,2,

',1,1,0,1,3, 1,,2,0,1,3,

',1,3,0,1,3,

', 0, 0, 0, 1, 3',

',0,0,0,2,3,

,0,0,0,0,5,

',0,0,0,0,2,

',1,1,0,1,3,

,1, 2, 0, 1, 3,

,1,3,0,1,3,

,0,0,0, 1,3,

',0,0,0,2,3,

,0,0,0,0,5,

,0,0,0,0,2,

',0,0,0,0,2,

',1,1,0,1,3,

',1,2,0,1,3,

,1,3,0,1,3,

,1,1,0,2,3,

', 1, 2, 0, 2, 3,

,1,3,0,2,3,

',0,0,0,1,3,

,0,0,0,2,3,

',0,0,0,0,5,

',0,0,0,0,2,

',0,00,0,02,

',0,0,0,0,2,

',0,0,0,0,2, 0.OOOE+00,

1. OOOE+00,

0. OOOE+00, 0 OOOE+00, 0 OOOE+00, 5 350E+04,

2. 350E+03,

7. 600E-04, 2.778E-06, 2.778E-04, 8.333E-03, 0.OOOE+00,

0. OOOE+00,

5. 350E+04,

6. 100E+02,

7. 600E-04, 1.250E-02, 2.500E-02, 4 167E-02,

8. 333E-02,

1. 667E-01,

0. OOOE+00,

0. OOOE+00, 0 OOOE+00,

0. OOOE+00,

5. 350E+04,

9. 100E+02,

7. 600E-04,

3. 333E-01,

6. 666E-01, 7.500E-01,

1. OOOE+00,

2. 7778E-06 0. OOOE+00

0. OOOE+00,

3. 470E-04, 2.778E-04 8.333E-03 1.250E-02

0. OOOE+00,

0. OOOE+00, 3.470E-04,

2. 500E-02

4. 167E-02

8. 333E-02

1. 667E-01

3. 333E-01

0. OOOE+00,

3. 470E-04,

6. 666E-01

7. 500E-01

1. OOOE+00 1.250E+00

0. OOOE+00 0 OOOE+00 0 OOOE+00

3. 004E+02 1. OOOE+00

2. 800E-04,

0. OOOE+00

0. OOOE+00 25.58E+00

1. OOOE+00 2.800E-04, 4.871E+01

4. 871E+01

0. OOOE+00

9. 450E+01

0. OOOE+00

38. 16E+00

2. 700E+03

2. 800E-04, Change X/Q and transfer flow 3.470E-04, Normal HVAC isolation at 30 seconds Change X/Q and transfer flow 3.470E-04, Change X/Q and transfer flow 3.470E-04, MC-CCN Calculation Cover Sheet, Rev. 0 11/19/03

NL-07-0067, Enclosure 6, Page 23 of 27 MC-F-04-0058, Version 2 Southern Nuclear Design Calculations Plant:

Unit:

Calculation Number:

Farley Nuclear Plant 0 1 0 2 01 I & 2 SM-96-1064-001

Title:

Fuel Handling Accident Doses Sheet:

I 32

'TIME INTERVAL

',0,0,0,0,2,

'TIME INTERVAL

',0,0,0,0,2,

'TIME INTERVAL

',0,0,0,0,2,

'TIME INTERVAL

',0,0,0,0,2,

'TRANSFER CFM

',0,0,0,1,3,

'DOSE PARAMS

',0,0,0,0,5, 3.470E-04, 0.OOOE+00

'TIME INTERVAL

'TRANSFER CFM

'DOSE PARAMS 0, 0, 0, 0, 2, 0,0,0, 0,2,

,0,0, 0,0,2, 0,0,0,0,2, 0,0,0,0,2, 0, 0, 0, 0,2, 0,0,0,0,2, 0, 0, 0, 0, 2, 0,0,0,1,3, 0, 0, 0, 0, 5,

1. 250E+00,

1. 500E+00, 1.750E+00,

2. OOOE+00, 5.350E+04, 2.900E-04,

2. 250E+00, 2.500E+00,

3. OOOE+00,

4. OOOE+00,

5. OOOE+00,

6. OOOE+00,

7. OOOE+00,

8. OOOE+00, 5.350E+04,

3. 300E-05,

1. 500E+00

1. 750E+00

2. OOOE+00

2. 250E+00

0. OOOE+00, 3.470E-04,

2. 500E+00

3. OOOE+00

4. OOOE+00

5. OOOE+00

6. OOOE+00

7. OOOE+00

8. OOOE+00

24. OOE+00 0.OOOE+00,

3. 470E-04, 31.72E+00

1. 100E-04,

16. 55E+00

1. OOOE-05, Change X/Q and transfer flow Change X/Q and transfer flow 3.470E-04, 0.OOOE+00

'END

',0,0,0,0,0, 0.OOOE+00, 0.OOOE+00, 0.OOOE+00 As with the equipment hatch analysis, the containment activity at the end of two hours is essentially zero:

ACTIVITY RELEASED TO ENVIRONMENT AND IN EACH NODE AT END OF...

2.OOOE+00 (HRS)

ISO I

I I

I KR KR KR KR KR XE XE XE XE NAM 131 131 132 132 133 133 135 135 83M 85M 85 87 88 131M 133M 133 135 ENV.

1.167E+02 1.167E+02

1. 703E-11

1. 703E-11 8. 012E+00

8. 012E+00

6. 812E-03

3. 304E-13

1. 580E-03

8. 956E+02

3. 572E-20

5. 392E-07

3. 680E+02 7.597E+02

4. 923E+04

1. 030E+01 Contnmnt 6.871E-03 6.871E-03 5.905E-16 5.905E-16

4. 474E-04 4.474E-04 3.354E-07 3.354E-07 1.OOOE-17 7.049E-08 5.306E-02 7.887E-25 2.046E-11 2.171E-02 4.401E-02 2.888E+00 5.330E-04 CtrlRoom

2. 915E-03 2.915E-03

2. 505E-16 2.505E-16

1. 898E-04 1.898E-04 1.423E-07 1. 423E-07

5. 626E-17

3. 965E-07

2. 984E-01 4.436E-24

1. 151E-10 1.221E-01 2.475E-01

1. 624E+01 2.998E-03 Note that the offsite doses remain essentially unchanged (except for the small impact of increased control room flow, filtration and holdup) compared to pp 14-16:

MC-CCN Calculation Cover Sheet, Rev. 0 11/19/03

NL-07-0067, Enclosure 6, Page 24 of 27 MC-F-04-0058, Version 2 Southern Nuclear Design Calculations Plant:

Unit:

Calculation Number:

Farley Nuclear Plant El 1 E0 2 l] I & 2 SM-96-1064-001

Title:

Fuel Handling Accident Doses Sheet:

33 CALCULATION MULTI NODE EXCLUSION RADIUS TOTAL 2.O0OE-01 FOR WHOLEBDY DOSE (REMS)

CONTAINMENT WITH ESF LOW POPULATION ZONE TOTAL 7.367E-02 CALCULATION FOR SKIN DOSE (REMS)

MULTI NODE CONTAINMENT WITH ESF EXCLUSION RADIUS LOW POPULATION ZONE TOTAL 4.226E-01 TOTAL 1.557E-01 CALCULATION FOR THYROID DOSE (REMS)

MULTI NODE CONTAINMENT WITH ESF EXCLUSION RADIUS LOW POPULATION ZONE TOTAL 6.847E+01 TOTAL 2.523E+01 Then the control room doses are calculated in the EXCEL spreadsheet used on pp 18-23:

MC-CCN Calculation Cover Sheet, Rev. 0 11/19/03

MC-F-04-0058, Version 2 Southern Nuclear Design Calculations I Plant:

Unit:

Calculation Number:

Farley Nuclear Plant 1l 1 0 2 2 1 & 2 SM-96-1064-001

Title:

Fuel Handling Accident Doses Sheet:

1 34 Case 1 - Control Room Dose Results Thyroid Thyroid Thyroid DCF Dose Dose (Rem/Ci)

(REM)

I-

-Jz 131 132 133 135 1.1OE+06 6.30E+03 1.80E+05 3.1 OE+04 0 - 2 hr 37.69949 2.83E-14 0.418738 5.97E-05 2 - 8 hr 1.453947 6.21 E-16 0.015257 1.9E-06 Exposure 0-720 hrs (Ci-Sec/M3) 0.102576 1.32E-14 0.006948 5.73E-06 Iodine Skin DCF (REM-M3/

Sec-Ci) 3.07E-02 1.10E-01 8.90E-02 7.86E-02 Iodine Skin Dose (REM) 0.003149 1.45E-15 0.000618 4.5E-07 Iodine Body DCF (REM-M3/

Sec-Ci) 5.59E-02 3.55E-01 9.11 E-02 2.49E-01 Iodine Body Dose (REM) 0.000288 2.36E-16 3.18E-05 7.17E-08 0.00032 TOTAL 38.11829 1.469206 TOTAL 0.003768 TOTAL KR 83M KR 85M KR 85 KR 87 KR 88 XE 131M XE 133M XE 133 XE 135M Skin DCF (REM-M3/

Sec-Ci) 0.OOE+00 4.97E-02 4.84E-02 3.36E-01 7.76E-02 1.33E-02 2.96E-02 9.67E-03 2.14E-02 TOTAL NG Skin Dose (REM) 0 1.09E-07 0.078308 1.05E-23 5.13E-11 0.008802 0.039692 0.851235 0.000349 0.978386 Body DCF (REM-M3/

Sec-Ci) 1.27E-05 2.31 E-02 3.31 E-04 1.33E-01 3.38E-01 1.25E-03 4.29E-03 4.96E-03 6.37E-02 TOTAL NG Body Dose (REM) 2.2162E-22 2.5481 E-09 2.6911E-05 2.0949E-25 1.122E-11 4.1571E-05 0.00028908 0.02194075 5.2146E-05 0.02235046 0.982154 (w/ iodine) 0.02267 (w/ iodine)

These results meet the acceptance criteria from RG 1.195 discussed on sheet 1.

MC-CCN Calculation Cover Sheet, Rev. 0 11/19/03

MC-F-04-0058, Version 2 Plant:

Farley Nuclear Plant

Title:

Fuel Handling Accident Doses NL-07-0067, Enclosure 6, Page 26 of 27 Southern Nuclear Design Calculations Calculation Number:

2 SM-96-1064-001 Unit:

El1 2 211&;

I Shee:

The cloud surrounding the control room has two components, the activity in the auxiliary building and that released from the plant, dispersed in the environment and surrounding the auxiliary building.

Review of the initial activity distribution below indicates the later is significantly less than the concentration in the auxiliary building:

(101.5 + 15.4) Ci / (2 x 3600) sec x 1.66 x 10-3 s/m 3 < 15.4 Ci / (1 x 105 /35.3) m 3,

thus assuming the auxiliary building contribution on six sides of the control room is conservative. The initial corridor activity ACTIVITY RELEASED TO ENVIRONMENT AND IN EACH NODE AT END OF...

2.778E-06 (HRS)

ISO I

I I

I KR KR KR KR KR XE XE XE XE NAM 131 131 132 132 133 133 135 135 83M 85M 85 87 88 131M 133M 133 135 ENV.

1. 371E-03 1.371E-03

2. 122E-16 2.122E-16

9. 471E-05

9. 471E-05 8.167E-08 8.167E-08 4.175E-18

1. 915E-08

1. 051E-02

4. 665E-25

6. 651E-12

4. 321E-03

8. 940E-03

5. 784E-01 1.227E-04 Contnmnt 1.015E+02 1.015E+02 1.571E-11 1.571E-11 7.010E+00 7.010E+00 6.045E-03 6.045E-03 3.090E-13 1.417E-03 7.780E+02 3.453E-20 4.923E-07 3.198E+02 6.617E+02 4.281E+04 9.085E+00 CtrlRoom 1.764E-06

1. 764E-06

2. 729E-19 2.729E-19

1. 218E-07

1. 050E-10

5. 370E-21

2. 463E-11

1. 352E-05

6. 000E-28

8. 555E-15

5. 558E-06

1. 150E-05 7 440E-04

1. 579E-07 Corridor

1. 538E+01

2. 380E-12

1. 062E+00

9. 161E-04

4. 683E-14

2. 148E-04

1. 179E+02 5.232E-21

7. 460E-08 4.847E+01

1. 003E+02

6. 488E+03 1.377E+00 is put into a MICROSHIELD rectangular block 23' x 23' x 190' with a (Ref. 32c) 2' concrete shield (assumed density of 2.35 g/cc) yielding a control room dose rate of 0.004 mR/hr per wall/floor cieling:

MicroShield 4.21 -

Serial #4.21-00817 Licensed to Southern Company Services CONVERSION OF CALCULATED EXPOSURE IN AIR TO DOSE FILE: FHASHINE.MS4 Case

Title:

FHA Shine This case was run on Sunday, December 12, 1999 at 9:23 a.m.

Results (Summed over energies)

Units Photon Fluence Rate (flux)

Photons/cm*/sec Photon Energy Fluence Rate MeV/cmV/sec Without Buildup 8.540e-002 7.511e-002 1.354e-004 1.182e-006 1.182e-004 With Buildup 3.311e+000 2.054e+000 3.865e-003 3.374e-005 3.374e-003 Exposure Rate in Air Absorbed Dose Rate in Air 11 mR/hr mGy/hr mrad/hr Then the shine contribution to the control room is 0.004 mR/hr x 6 x 8 hr = 0.2 mR.

MC-CCN Calculation Cover Sheet, Rev. 0 11/19/03

NL-07-0067, Enclosure 6, Page 27 of 27 MC-F-04-0058, Version 2 Southern Nuclear Design Calculations Plant:

Unit:

Calculation Number:

Farley Nuclear Plant 0l1 02 2

1 1 & 2 SM-96-1064-001

Title:

Fuel Handling Accident Doses Sheet:

43 (airlocks open) Results and

Conclusions:

Case 1 (no direct auxiliary building inleakage)

Thyroid dose Whole Body dose Skin dose 39.6 REM

<0.1 REM 1.0 REM meet the acceptance criteria of Regulatory Guide 1.195.

MC-CCN Calculation Cover Sheet, Rev. 0 11/19/03

NL-07-0067, License Amendment Request to Technical Specification 3.9.3, Containment Penetrations, Personnel Air Locks Doors Open During Fuel Movement

Text

Enclosures:

SUMMARY

7.0 REFERENCES

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