ML041590204

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Response to Request for Information and Supplement I to Proposed Change Number (PCN) 534 Containment Penetrations
ML041590204
Person / Time
Site: San Onofre  Southern California Edison icon.png
Issue date: 06/03/2004
From: Nunn D
Southern California Edison Co
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML041590204 (46)
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Text

iSOUTHERN CALIFORNIA J EDISON An EDISON INTERNATIONAL6 Company Dwight E. Nunn Vice President June 3, 2004 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20055

Subject:

Docket Nos. 50-361 and 50-362 Response to Request for Information and Supplement I to Proposed Change Number (PCN) 534 Containment Penetrations San Onofre Nuclear Generating Station Units 2 and 3

References:

1. Letter dated August 4, 2003, from Dwight E. Nunn (SCE) to Document Control Desk (NRC),

Subject:

Proposed Change Number (PCN) 534, Containment Penetrations"

2. Letter dated December 24, 2003, from A. E. Scherer (SCE) to Document Control Desk (NRC),

Subject:

Response to Request for Additional Information (RAI) regarding Containment Structure Equipment Hatch Shield Doors This letter provides additional information in response to NRC questions on Proposed Change Number (PCN) 534. Items 1 through 4 in Enclosure 2 provide additional information to supplement our submittal of December 24, 2003 (Reference 2). Items 5 and 6 in Enclosure 2 are responses to additional questions received from the NRC staff.

Additionally, as noted in the response to item 4 in Enclosure 2, this letter provides revised proposed Technical Specification changes. These changes provide additional restrictions to opening of the Containment equipment hatch during core alterations and movement of irradiated fuel. The revised Technical Specifications are provided in Enclosure 3. Revised Bases for Unit 2 are also provided in Enclosure 3 for information.

SCE has evaluated the information in the Enclosures and concludes that there is no change to the previous finding of "no significant hazards consideration."

P.O. Box 128 San Clemente. CA 92674-0128 949-368-1480 Fax 949-368-1490 Die)

Document Control Desk June 3, 2004 SCE requests approval of the proposed amendment by August 2004 to support the San Onofre Unit 3 Cycle 13 refueling outage. Once approved, the amendment shall be implemented within 60 days.

If you have any questions or require additional information, please contact Mr. Jack Rainsberry at 949-368-7420.

Sincerely,

Enclosures:

1. Notarized Affidavits
2. RAI Responses
3. Revised Proposed Technical Specification Pages:

A. Proposed Technical Specification pages, Redline and Strikeout, Unit 2 B. Proposed Technical Specification pages, Redline and Strikeout, Unit 3 C. Proposed Technical Specifications pages, Unit 2 D. Proposed Technical Specifications pages, Unit 3 E. Proposed TS Bases pages, Redline and Strikeout, Unit 2 (typical for both Units - for information) cc:

B. S. Mallett, Regional Administrator, NRC Region IV B. M. Pham, NRC Project Manager, San Onofre Units 2 and 3 C. C. Osterholtz, NRC Senior Resident Inspector, San Onofre Units 2 and 3 S. Y. Hsu, Department of Health Services, Radiologic Health Branch

ENCLOSURE I NOTARIZED AFFIDAVITS

UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION Application of SOUTHERN CALIFORNIA EDISON COMPANY, ETAL. for a Class 103 License to Acquire, Possess, and Use Utilization Facility as Part of Unit No. 2 of the San Onofre Nuclear Generating Station

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Docket No. 50-361 Supplement I to Amendment Application No. 220 SOUTHERN CALIFORNIA EDISON COMPANY, et a]. pursuant to 10CFR50.90, hereby submit Supplement 1 to Amendment Application No. 220. This amendment application consists of Supplement 1 to proposed Change No. 534 to Facility Operating License No.

NPF-10. Proposed Change No. 534 is a request to revise Technical Specification (TS) 3.9.3, "Containment Penetrations." This change will permit the containment equipment hatch to remain open during core alterations and movement of irradiated fuel.

State of California County of San Diego Subscribed and sworn to (or affirmed) before me this

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day of 0

, 200qL.

Vice Presid t

FRANCES M. THURBER i

Commission if 1295266 1

Notary Public - Califomia 7 WCSan Diego County r

i ZnE N rry Public

UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION Application of SOUTHERN CALIFORNIA EDISON COMPANY, ET AL. for a Class 103 License to Acquire, Possess, and Use A Utilization Facility as Part of Unit No. 3 of the San Onofre Nuclear Generating Station

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Docket No. 50-362 Supplement 1 to Amendment Application No. 205 SOUTHERN CALIFORNIA EDISON COMPANY, et al. pursuant to 10CFR50.90, hereby submit Supplement I to Amendment Application No. 205. This amendment application consists of Supplement I to proposed Change No. 534 to Facility Operating License No.

NPF-10. Proposed Change No. 534 is a request to revise Technical Specification (TS) 3.9.3, "Containment Penetrations." This change will permit the containment equipment hatch to remain open during core alterations and movement of irradiated fuel.

State of California County of San Diego Subscribed and sworn to (or affirmed) before me this 3 day of 20_dq.

-By:~

Dwight E. N Ln Vice Presidens FRANCES M. THURBER L Commission i1 1295266 1 Notary Public - California 2 San Diego County

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ENCLOSURE 2 RAI RESPONSES

ITEM 1 - RAI # 4 A value of 1000 cfm is assumed for the value of unfiltered inleakage into the control room.

Because this value is not based upon a measurement, sufficient justification should be provided to explain why this number is appropriate. Provide sufficient details regarding your control room, design, maintenance and assessments to justify the use of and your plans to verify this number.

SCE to NRC December 24, 2003 Response:

San Onofre Units 2 and 3 Technical Specifications (TS) Limiting Condition for Operation (LCO) 3.7.11 requires the control room boundary to be tested every 24 months to demonstrate that the control room boundary has at least 0.125 inches water gauge positive pressure with respect to the atmosphere. The TS 3.7.11 Bases state that this pressurization prevents unfiltered inleakage. Pressurization tests over the past 10 years show that the lowest positive pressure for a single operating CREACUS train was 0.56 inches of water gauge with respect to the atmosphere.

The San Onofre Units 2 and 3 control room design has numerous features to minimize and prevent control room inleakage. These features include a design in which the Control Room Emergency Air Cleanup System (CREACUS) units are wholly contained within the Control Room Envelope (CRE), effective boundary maintenance as evidenced by the high pressure gradient across the control room boundary during pressurization tests, and the existence of procedures requiring periodic control room boundary integrity inspections, control room damper inspections, and control of CREACUS breaches during routine maintenance activities. Based on the above information, a value of 1000 cfm assumed for the value of unfiltered inleakage is considered conservative.

San Onofre has committed to perform Control Room Envelope inleakage testing in accordance with NRC Generic Letter 2003-01, "Control Room Habitability," to verify actual inleakage. This testing will be completed prior to the Unit 3 Cycle 13 outage that is currently scheduled to begin in September 2004.

Additional Information:

The following additional information is provided to support the conclusion that the assumed Control Room inleakage value of 1000 cfm is considered conservative. The areas of potential vulnerabilities for unfiltered inleakage into the Control Room boundary are discussed with supporting information to demonstrate the SONGS design and boundary maintenance practices have addressed these issues. In addition, preliminary results of the control room inleakage testing are discussed.

Ductinq and Housings All CREACUS ducting is constructed of welded steel plate with gasketed flange joints.

Compared to other types of ducting such as commercial, pocket lock, non-seal welded, or non-bolted, the SONGS ducting design has a minimal number of potential leakage paths and is considered to be superior to these other designs.

Page 1 of 21

Most of the ducting is located within the Control Room boundary. The sections of ducting located outside the Control Room boundary, which could be vulnerable to inleakage are not insulated and can be visually inspected. These inspections have not revealed any deficiencies.

There are two other HVAC system ducts, which pass through the boundary and can operate at a higher pressure relative to the control room. These ducts have the same design as the CREACUS ducting described above. These ducts have been leak tested and there are no leaks.

Positive pressure sections of CREACUS ducting inside the boundary, but upstream of the filters, have the same flanged design. Inspections show no deficiencies.

The CREACUS filter units are all located within the boundary. Any inleakage into these housings would be filtered air and therefore is not a vulnerability.

There are six other fan coil units, which provide cooling to the cabinet area of the Control Room. Four of these units are safety related and two of them are non safety related. These units are all located outside the boundary and present a potential vulnerability for outside air inleakage. The emergency units are welded plate and bolted connections with low vulnerability. The non safety related units are sheet metal with screwed connections and present a higher vulnerability to inleakage. The housings, flexible connectors, penetrations, and access doors for all six units have been inspected, tested, and sealed as necessary to prevent inleakage.

Boundary and Boundary Penetrations The Control Room boundary consists of a poured concrete slab over metal decking for the floor and ceiling. The adjacent walls are primarily concrete. There are some sections of the walls, which are lath and plaster. Included in the boundary are a concrete walled pipe shaft and a concrete walled elevator/stairwell shaft. All accessible walls, floors, ceilings, joints, and HVAC ducting penetrations are routinely inspected and repaired as necessary.

There are a large number of penetrations through the control room boundary. Many of these penetrations are cables and conduits. All accessible penetrations including the cables and conduits are routinely inspected for pressure boundary integrity and repaired as necessary. Also, construction specifications require sealing and testing all new or modified control room boundary penetrations. Floor and equipment drains are routinely inspected and filled with water.

BoundarV Doors There are 10 control room boundary doors. All doors are routinely inspected for structural integrity, sealing, latching, and fastener integrity. Also, each door is routinely smoke tested. Any deficiencies are repaired.

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Isolation Dampers There are 5 sets of redundant control room isolation dampers. Three sets of these dampers isolate ducting, which is at a relatively low pressure compared to the outside.

All dampers are a bubble tight design with seals on each damper blade. All dampers are routinely inspected, maintained and tested to ensure they seal properly and stroke properly.

Control Room Pressure Another indication of an effectively maintained control room boundary is the Control Room positive pressure measured during Technical Specification Surveillance Testing.

The testing is performed every 18 months. In the last 10 years of testing, the lowest developed positive pressure was 0.56 inches of water. This is significantly greater than the required 0.125 inches of water per Technical Specifications.

Control Room Breaches A program is in place to provide administrative controls for pre-planned breaches of the control room boundary to support maintenance activities. The controls include restrictions on the delivery of bulk chemicals, heightened awareness of alarms, continuous communication, pre-planned methods to secure the breach, and continuous manning.

Control Room Envelope Inleakage Testing The Control Room Envelope inleakage testing to verify actual inleakage was conducted from May 18, 2004 to May 25, 2004. The preliminary review of the test data indicates that the control room envelope inleakage rate is well below 1000 cfm. Industry experience indicates that the final results are not expected to change significantly.

Formal analysis of the test data and determination of the actual inleakage rates is being performed. As described in SCE's 60-day response to Generic Letter 2003-01 (Reference 1), following the completion of inleakage testing SCE will submit a letter describing how and when the analyses, tests, and measurements were performed and the final results of the testing.

References:

1.

SCE to NRC letter dated August 5, 2003, Response to Generic Letter 2003-01 "Control Room Habitability," San Onofre Nuclear Generating Station Units 2 and 3.

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ITEM 2 - RAI # 10 What criteria will be used to determine if closure of the Containment is necessary in the event that environmental conditions could impact fuel handling? Has the impact of wind on fuel handling been evaluated (for example, reduced pool visibility due to pool surface disruption)? What steps would be taken in the event of severe weather to minimize the impact of flying debris?

SCE to NRC December 24. 2003 Response:

The Equipment Hatch is at ground elevation (30 feet above sea level) and the refueling deck is at 63 feet. Accordingly, there is a tortuous path between the outside air and the surface of the refueling pool. The procedure for fuel movement is S023-X-7. This procedure includes general guidance in addition to specific procedure steps, including a verification scan of the top of all core locations to check for debris (step 3.16). This could not be satisfactorily accomplished with disturbed water or water that contains debris. If there were an unacceptable impact on pool visibility, action would be taken to either secure the Containment Hatch or Shield Doors or Core Alteration/Movement of irradiated Fuel would be secured.

Additional Information:

In addition to the guidelines given in the fuel movement procedure, Abnormal Operating Instruction S023-13-8 (Severe Weather), requires in steps 2.1 and 2.5 (for Units 2 and 3, respectively) to verify "Missile Barrier doors are closed," including the Containment Structure Equipment Hatch Shield Doors. The severe weather conditions that require entry to this procedure include tornado warning, hurricane watch, flash flood watch or warning, or tsunami warning.

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ITEM3-RAI#s17andl8 The response to RAI #17 states that the current SONGS licensing basis atmospheric dispersion factors (X/Q values) used to evaluate releases from the containment to the control room HVAC intake are applicable for all potential containment release points.

This statement is inaccurate. The current licensing basis X/Q values can be used for newly identified release scenarios only if they are appropriate for the application in which they are being used. The use of the current licensing basis X/Q value of 3.1 E-3 sec/M3 does not appear appropriate for this radiological analysis. As stated in SONGS UFSAR Section 2.3.4.2, the control room (CR) X/Q value of 3.1 E-3 sec/m3 is based on the Murphy & Campe diffuse source-point receptor algorithm. This algorithm is applicable when activity is assumed to leak from many points on the surface of the containment in conjunction with a single point receptor (i.e., CR air intake); that is, the activity is assumed to be homogeneously distributed throughout the containment and the release rate is assumed to be reasonably constant over the surface of the building.

This is not the situation in this accident scenario where the release is assumed to occur through the open containment equipment hatch shield doors. Releases from open containment equipment hatch shield doors to the CR air intake are typically modeled using the Murphy & Campe point source-point receptor algorithm in lieu of the Murphy &

Campe diffuse source-point receptor algorithm.

In addition, the response to RAI #18 states that the CR X/Q value of 3.1 E-3 sec/m3 is based on a distance of 180 feet between the containment surface and the midpoint of the two CR emergency HVAC intakes. This same X/Q value is also being used to model releases to the CR normal HVAC intake during the first three minutes of the accident, prior to CR isolation. However, the distance between the Unit 2 containment surface and the normal HVAC is approximately 118 ft, not the 180 ft assumed in generating the CR X/O value of 3.1 E-3 sec/M3.

Please either justify quantitatively the continued use of 3.1 E-3 sec/M3 as a bounding CR X/Q value for this radiological analysis or provide a new set of CR X/Q values appropriate for the proposed change. An acceptable method for determining a new set of CR X/Q values is provided in Regulatory Guide 1.194, "Atmospheric Relative Concentrations for Control Room Radiological Habitability Assessment at Nuclear Power Plants.

Additional Information:

As suggested in the RAI, a new set of control room (CR) X/Q values to define dispersion between a Unit 2 or 3 containment equipment hatch and the control room outside air normal and emergency ventilation intakes has been calculated using the guidance in Regulatory Guide (RG) 1.194, "Atmospheric Relative Concentrations for Control Room Radiological Habitability Assessment at Nuclear Power Plants." The analysis uses the ARCON96 computer code. A comparison of the 0 to 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> ARCON96 dispersion values with the 0 to 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> current licensing basis (CLB) CR X/Q value of 3.1 E-3 sec/M3 shows that it is conservative to use the CLB X/Q value in the dose analysis that supports the license amendment request (PCN-534).

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Meteorological Data The ARCON96 atmospheric dispersion analysis uses actual site hourly meteorological data spanning ten full years of 1993 through 2002. Full year meteorology is used to eliminate bias due to seasonal fluctuations. RG 1.184 Regulatory Position 3.1 states that 5 years of hourly observations are considered to be representative of long-term trends at most sites. The use of ten years of meteorological data satisfies this recommendation, while enhancing the statistical basis for the calculated control room atmospheric dispersion factors due to the expanded meteorological data set.

The input meteorological data identify invalid data by coding such data as either "999" or "9999". In each year, more than 99 percent of the lower level wind speed data are valid. Overall, about 99.8 percent of the lower level wind speed data are valid. Except for year 1994, more than 95 percent of each year's upper level wind speed data are valid. Overall, about 96.5 percent of the upper level wind speed data are valid.

Therefore, the meteorological input is representative.

The meteorological tower is located above the plant on the north bluff. The meteorological tower's lower wind instrument is at elevation 10 meters above the north bluff grade. The meteorological tower's upper wind instrument is at elevation 40 meters above the north bluff grade.

The meteorological data was converted to the ARCON96 format presented in NUREG/CR-6331 Section 4.4.2 and RG 1.194 Appendix A, Table A-1.

Consistent with RG 1.194 Regulatory Position 3.1, wind direction is expressed as the direction from which the wind is blowing (i.e., the upwind direction from the center of the site) referenced from true north. A north wind (wind from the north) is entered as 3600 and a south wind is entered as 1800.

Consistent with RG 1.194 Regulatory Position 3.1, atmospheric stability is entered as a number from 1 through 7. A stability class of 1 represents extremely unstable conditions, and a stability class of 7 represents extremely stable conditions.

Atmospheric stability classes are determined from the AT given in the meteorological data.

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Non-Meteorological Data Input RG 1.194 Appendix A Table A-2 discusses input parameters for ARCON96. Per the following table, the ARCON96 analysis complies with the regulatory guidance presented in Table A-2.

ARCON96 Input Parameters for Containment Equipment Hatch Release Parameter I

Acceptable Input I

Comments Lower Measurement Use the actual instrumentation height when Used actual measurement height, which is Height, meters known. Otherwise, assume 10 meters.

10 meters above bluff grade. The bluff grade is above the plant grade.

Upper Measurement Use the actual instrumentation height when Used actual measurement height of 40 Height, meters known. Otherwise, use the height of the meters above the bluff grade.

containment or the stack height, as appropriate. If wind speed measurements are available at more than two elevations, the instrumentation at the height closest to the release height should be used.

Wind Speed Units Use the wind speed units that correspond to he raw meteorological data expresses wind the units of the wind speeds in the speeds in miles per hour. However, these meteorological data file.

data are pre-processed to convert the wind speeds to meters per second in the resulting MET input files. The ARCON96 input files

(*.RSF) are set for wind speeds in units of meters per second. Thus, the units used for ind speeds in the analysis are applied consistently.

Release Height, Use the actual release heights whenever Used the actual release height. Equipment meters available. Plume rise from buoyancy and Hatch is 11.58 meters above plant grade.

mechanical jet effects may be considered in establishing the release height if the analyst can demonstrate with reasonable assurance that the vertical velocity of the release will be maintained during the course of the accident.

If actual release height is not available, set release height equal to intake height.

Building Area, Use the actual building vertical cross-Used the 2123.33 square meter cross-meters sectional area perpendicular to the wind ectional area of the containment.

direction. Use default of 2000 m2 if the area is not readily available. Do not enter zero.

Use 0.01 m if a zero entry is desired.

Note: This building area is for the building(s) hat has the largest impact on the building ake within the wind direction window. This is usually, but need not always be, the reactor containment. With regard to the diffuse area source option, the building area entered here may be different from that used to establish the diffuse source.

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l Parameter Acceptable Input l

Comments Vertical Velocity, Note: the vent release model should not be The vent release model is not used.

meters/second used for DBA accident calculations.

For stack release calculations only, use the actual vertical velocity if the licensee can demonstrate with reasonable assurance that the value will be maintained during the course of the accident (e.g., addressed by echnical specifications), otherwise, enter ero. If the vertical velocity is set to zero, RCON96 will reduce the stack height by 6 imes the stack radius for all wind speeds.

If this reduction is not desired, the stack radius should also be set to zero.

Stack Flow, Use actual flow if it can be demonstrated Stack flow is set to zero.

eters3ls with reasonable assurance that the value will be maintained during the course of the accident (e.g., addressed by technical specifications). Otherwise, enter zero. The low is used in both elevated and ground-level release modes to establish a maximum X/Q value. This value is significant only if the flow is large and the distance from the release point to the receptor is small.

tack Radius, Use the actual stack internal radius when Stack radius is set to zero.

eters both the stack radius and vertical velocity are available. If the stack flow is zero, the radius should be set to zero.

Distance to Use the actual, straight line, horizontal he Equipment Hatches are on the opposite Receptors, meters distance between the release point and the side of the Containment structure from the control room intake.

Control Room Intakes and are located at plant grade level. The top of the For ground-level releases, it may be Containment is 181 feet above plant grade appropriate to consider flow around an and the top of the equipment hatch is 17.5 intervening building if the building is feet above grade; therefore, it is unrealistic sufficiently tall that it is unrealistic to expect to expect flow from the Equipment Hatch to low from the release point to go over the go over the Containment building. The building.

Equipment Hatch to receptor distances are measured as the shortest path around the Note: If the distance to receptor is less than Containment ("taut string length"), as allowed about 10 meters, ARCON96 should not be by section 3.4 of Regulatory used to assess relative concentrations.

Guide 1.194.

The equipment hatch source-receptor distance is not less than 10 meters.

Intake Height, Use the actual intake height. If the intake The actual heights at the centerline of the meters height is not available for ground level control room intakes are used.

releases, assume the intake height is equal to the release height. For elevated leases, assume the height of the tallest site building.

Elevation Difference, Use zero unless it is known that the release The release and receptor heights are meters heights are reported relative to different reported with respect to the same grade grades or reference datum.

datum.

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Parameter l

Acceptable Input Comments Direction to Use the direction FROM the intake back SONGS' site arrangement drawings do Source, degrees TO the release point. (Wind directions have a "Plant North" designation that is are reported as the direction from which 57° west of "true north;" consequently, the wind is blowing. Thus, if the direction wind directions are corrected to true north from the intake to the release point is as the point of reference.

north, a north wind will carry the plume from the release point to the intake.)

For the scenario of an equipment hatch release, the X/Q is calculated assuming Note: some facilities have a "plant north" low both around and over (through) the shown on site arrangement drawings that containment building, and the higher of is different from "true north." The direction the X/Q values is used.

entered must have the same point of reference as the wind directions reported in the meteorological data.

For ground level releases, if the plume is assumed to flow around a building rather than over it, the direction may need to be modified to account for the redirected flow. In this case, the X/Q should be calculated assuming flow around and flow over (through) the building and the higher of the two X/O s should be used.

Surface Use a value of 0.2 in lieu of the default Used value of 0.2. SONGS is a seaside Roughness value of 0.1 for most sites. (Reasonable site with low surface vegetation.

Length, meters values range from 0.1 for sites with low surface vegetation to 0.5 for forest covered sites.)

Wind Direction Use the default window of 90 degrees (45 Used 90 degrees.

Window, degrees degrees on either side of line of sight from the source to the receptor).

Code Default Minimum Wind Use the default wind speed of 0.5 m/s Used the default wind speed of 0.5 m/s.

Speed, (regardless of the wind speed units The minimum SONGS site met tower meters/second entered earlier), unless there is some wind speed reported is 0.3 mph, or 0.13 indication that the anemometer threshold m/s. Thus, the anemometer threshold is Code Default is greater than 0.6 m/s.

less than 0.6 m/s.

Averaging Sector Although the default value is 4, a value of Used 4.3.

Width Constant 4.3 is preferred. (A future revision to ARCON96 will change the default to 4.3)

Code Default Initial Diffusion hese values will normally be set to zero. Used the diffuse source option.

Coefficients, If the diffuse source option is being used, meters see Regulatory Position 2.2.4.

Hours in Use the default values.

Used the default values.

Averages Minimum Number Use the default values.

Used the default values.

f Hours Page 9 of 21

Control Room HVAC Intakes Atmospheric dispersion factors have been calculated for the six combinations representing two activity release locations (Unit 2 and Unit 3 containment equipment hatches) and three control room HVAC intake locations (CR normal, Unit 2 emergency and Unit 3 emergency).

The center of the control room normal air intake is at plant elevation 35.50 feet (10.82 meters). The center of each control room emergency air intake is at plant elevation 43.00 feet (13.11 meters).

Containment Equipment Hatch Release Characteristics Atmospheric dispersion between the containment equipment hatch and the control room HVAC intakes is modeled as an area (diffuse) source, ground level release.

The Containment Equipment Hatch is a large circular opening through the containment wall. The equipment hatch meets the conditions for a diffuse source as set forth in RG 1.194 Section 3.2.4.8: (1) the release from the hatch will be essentially equally dispersed over the entire opening, and (2) assumptions of mixing, dilution and transport within Containment necessary to meet condition 1 are supported by the interior containment arrangement. Consistent with RG 1.194 Section 3.2.4.8, the initial horizontal and vertical diffusion coefficients (Oy.O and azO) are determined to be 0.97 meters, based on the clear 19-foot diameter of the hatch opening.

The Unit 2 & 3 Containment Equipment Hatches are on the opposite side of their respective Containment structures from the control room air intakes. The containment equipment hatch diffuse release is assumed to be from its mid-height at plant elevation 38.00 feet (11.58 meters). The top of the Containment is 181 feet above plant grade; therefore, it is unrealistic to expect flow from the Equipment Hatch to go over the Containment building. The Equipment Hatches to receptor distances are measured as the shortest path around the Containment ("taut string length"), as allowed by RG 1.194 Section 3.4. To determine the taut string length, a tangent is drawn from each intake to the side of the containment closest to the equipment hatch. That distance is added to the length of the arc around the containment from the tangent line intersection to the centerline of the hatch.

As requested by RG 1.194 Appendix A.2, since the plume is assumed to flow around the containment building rather than over it, the x/Q value is calculated assuming flow both around and over (through) the building, and the higher of the x/Q values is used.

For the containment equipment hatch release, only the cross-sectional area of the containment is used to determine the building wake area. All other intervening buildings, such as the auxiliary building, are conservatively ignored. The building wake area of 2123.33 square meters is the projected area of the containment cylindrical lower portion and the containment upper dome portion.

Page 10 of 21

The following table presents the separation distances and wind directions that characterize the releases from the two containment equipment hatch release point locations to the three control room HVAC intake locations:

CONTAINMENT EQUIPMENT HATCH TO CONTROL ROOM MODELING Release Point Control Room Receptor Separation Distance Wind Direction (meters)

Over/ Around Containment I_

(degrees, North = 0)

U2 Ctmt Equip. Hatch Normal Air Intake 98.1 353/11 U2 Ctmt Equip. Hatch U2 emergency air intake 96.8 355 / 15 U2 Ctmt Equip. Hatch U3 emergency air intake 126.9 336 /343 U3 Ctmt Equip. Hatch Normal Air Intake 124 89/82 l

U3 Ctmt Equip. Hatch U2 emergency air intake 126.9 90/ 83 U3 Ctmt Equip. Hatch U3 emergency air intake 96.8 71 /51 The results of the ARCON96 analysis show that the Unit 2 equipment hatch to Unit 2 emergency air intake release path modeling flow around the containment building has the more conservative atmospheric dispersion (i.e., the maximum atmospheric dispersion factor) during the first 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> of a release. The resultant 95th percentile control room atmospheric dispersion factors for this release path, without control room occupancy factors, are:

Time Interval Equipment Hatch Release Dispersion Factors l_____________

(sec/in3) 0 to 2 hrs 7.99E-04 2 to 8 hrs 6.30E-04 8 to 24 hrs 1.77E-04 1 to 4 days 2.23E-04 4 to 30 days I

2.03E-04 A comparison of the 0 to 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> dispersion values, with the current licensing basis control room X/Q value of 3.1 E-3 sec/M3, shows that the dose analysis that supports the license amendment request (PCN-534) is conservative.

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ITEM 4 - RAI #s 12 and 14 Additional Information:

The response to RAI #12 provided in the December 24, 2003 SCE to NRC letter discussed an additional restriction to opening the containment equipment hatch during core alterations and movement of irradiated fuel. In addition, based on discussions with the staff regarding RAI # 14, an additional restriction is being proposed. Revised Technical Specification pages to reflect the fourth and fifth restrictions are being provided with this response.

The RAI #12 response credited the continuous radiation monitoring system in the discharge of the containment purge whenever the purge is in service and discussed Technical Specification 3.9.3 and station procedures requiring containment purge to be in service whenever core alterations or fuel movement are performed and the equipment hatch is open. This will be in a fourth condition to enable the equipment hatch to be open.

The RAI #14 response discussed the proposed Technical Specification not providing a condition for open equipment hatch being limited to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> post shutdown for the Fuel Handling Accident (FHA). This condition is currently contained in the Licensee Controlled Specifications (LCS) 3.9.101. Because this condition is not in the Technical Specifications an additional restriction is being added to ensure the reactor has been subcritical for at least 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> before the equipment hatch is open during core alterations or fuel movement. This will be a fifth condition to enable the equipment hatch to be open.

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Item 5 - New RAI Describe the radiation protection job planning and job-site coverage, and the radiation surveys/personal protection and dose monitoring equipment that will be provided to the crew during this emergency response action. Describe the initial (and continuing) radiological training that will be provided, including whether the crew workers will be qualified and trained to use respiratory protection devices or other means to limit intake of radioactive materials.

Response

In the case of a fuel handling accident with the Containment Equipment Hatch open the following items will be in place to support closure of the missile shield doors:

  • A Health Physics work control plan with specific instructions to Health Physics Technicians regarding maximum calculated dose, stay times, and worker protective actions,
  • Four Full Face Negative Pressure respirators with GMR-l-P100 iodine canisters,
  • Dose rate meters capable of assessing the radiological conditions during a fuel handling accident,
  • Air sampling equipment.

Workers will receive initial and continuing radiation worker training and will be qualified and trained to use respiratory protection devices.

Page 13 of 21

Item 6 - New RAI The licensee has designated a crew of workers to manually close the containment structure equipment hatch shield doors in the event of an in-containment fuel handling accident during refueling. Provide an estimate of CEDE and DDE for a member of the crew, to ensure that the crew members' doses are consistent with the requirements and exposure guidelines of 10 CFR 50.47(b)(11). List all pertinent assumptions (e.g.,

airborne source term, stay time, external radiation levels, respiratory protection factors, etc.) taken to develop the dose estimates to these emergency workers.

Response

An analysis (described below) has been performed that quantifies the exposure to workers should an FHA occur within containment while the containment equipment hatch is open. This analysis modeled:

A release of airborne radioactivity within containment using the FHA inside containment Analysis of Record (AOR) source term; Dispersion of the released activity within the volume of the containment; Diffusion of the containment activity out the open equipment hatch over a 30 minute event duration; Dispersion of the diffused activity within the air volume trapped between the missile shield doors and the containment outer wall; and Calculation of doses to a worker who is present in this trapped air volume for 30 minutes.

The results of the analysis were a 30-minute thyroid inhalation dose of 44.4 Rem, a 30-minute whole body immersion dose of 0.3 Rem, and a 30-minute beta-skin dose of 0.4 Rem.

Using a thyroid weighting factor of 0.03 per Regulatory Guide 1.183 Footnote 7 (Ref.

5.2), the Total Effective Dose Equivalent (TEDE) to a crew member outside the equipment hatch was calculated to be 1.63 Rem.

Analysis FHA Activity Released Inside Containment The initial airborne radioactivity released inside containment due to an FHA is shown in the second column of Table 1. These data were taken from Table 8.3-1 of Calculation N-4072-003 (Ref. 5.3), which is the AOR for an FHA inside containment.

Page 14 of 21

The released activity was then dispersed within the containment by dividing by the containment net free volume, which is 1.422E+06 ft3 (= 4.027E+1 0 cc) per Section 4.3.1 of Ref. 5.3. The resulting containment activity concentrations are shown in the third column of Table 1.

Table 1 Initial Containment Activity Due to an FHA AOR Initial Initial Containment Isotope Containment Activity Activity Concentration (Ci)

(CLcc) 1-129 4.570E-05 1.135E-15 1-130 1.190E-02 2.955E-13 1-131 5.400E+02 1.341 E-08 1-132 3.300E-07 8.195E-18 1-133 1.260E+02 3.129E-09 1-135 6.920E-01 1.718E-1 1 Kr-83m 1.180E-08 2.930E-19 Kr-85 2.490E+03 6.183E-08 Kr-85m 2.620E-01 6.506E-12 Kr-87 3.300E-1 3 8.195E-24 Kr-88 1.350E-03 3.352E-1 4 Xe-131 m 6.150E+02 1.527E-08 Xe-1 33m 2.070E+03 5.140E-08 Xe-1 33 9.040E+04 2.245E-06 Xe-1 35m 8.840E+00 2.195E-1 0 Xe-135 1.190E+03 2.955E-08 FHA Activity Exiting Containment The amount of motive force for the movement of airborne radioactivity from inside containment to the outside environment during the time required to close the missile shield doors was judged to be very small and characterized as "no motive force".

Therefore, the only mechanism to facilitate the movement of airborne radioactivity from inside containment to the outside environment would be mass diffusion.

An evaluation was performed to characterize the radioisotope diffusion of airborne Iodine, Krypton, and Xenon from containment to the outside environment to allow for the estimation of the radiation exposure to a worker outside the equipment hatch. The results are shown in the second column of Table 2. Details of the diffusion rate analysis are given in Appendix A. Using a 30 minute event duration, the radioisotope activity exiting containment due to diffusion is shown in the third column of Table 2.

Page 15 of 21

Table 2 Radioisotope Diffusion frorm Containment Diffusion Rate Diffusion in 30 Minutes Element (Ci/sec per C/cc)

(Ci per Clcc)

Iodine 90 1.620E+05 Krypton 172 3.096E+05 Xenon 140 2.520E+05 The Table 2 values were then applied to the containment airborne activity concentration profile at the start of the FHA (Table 1). The results are shown in Table 3.

Table 3 Total Activity Exiting Containment Due to an FHA Initial Containment Diffusion Activity Exiting Isotope Activity Concentration Factor Containment in 30 Min.

(Clcc)

(Ci per Clcc)

(Ci) 1-129 1.135E-15 1.620E+05 1.838E-10 1-130 2.955E-1 3 1.620E+05 4.787E-08 1-131 1.341 E-08 1.620E+05 2.172E-03 1-132 8.195E-1 8 1.620E+05 1.328E-12 1-133 3.129E-09 1.620E+05 5.069E-04 1-135 1.718E-11 1.620E+05 2.784E-06 Kr-83m 2.930E-19 3.096E+05 9.072E-14 Kr-85 6.183E-08 3.096E+05 1.914E-02 Kr-85m 6.506E-1 2 3.096E+05 2.014E-06 Kr-87 8.195E-24 3.096E+05 2.537E-1 8 Kr-88 3.352E-1 4 3.096E+05 1.038E-08 Xe-1 31 m 1.527E-08 2.520E+05 3.849E-03 Xe-1 33m 5.140E-08 2.520E+05 1.295E-02 Xe-1 33 2.245E-06 2.520E+05 5.657E-01 Xe-1 35m 2.195E-1 0 2.520E+05 5.532E-05 Xe-1 35 2.955E-08 2.520E+05 7.447E-03 Page 16 of 21

FHA Activity Concentration Outside Containment To determine the activity concentration outside containment resulting from the FHA, the activity exiting containment in Table 3 was dispersed throughout a volume equal to the air volume trapped between the missile shield doors and the containment outer wall.

This volume was determined to be 34 cubic meters as shown in Appendix B. The resulting activity concentrations outside containment are shown in Table 4.

Table 4 Activity Concentration Outside Containment Due to an FHA Activity Concentration Isotope Outside Containment (C/m3) 1-129 5.407E-12 1-130 1.408E-09 1-131 6.389E-05 1-132 3.905E-14 1-133 1.491 E-05 1-135 8.188E-08 Kr-83m 2.668E-1 5 Kr-85 5.630E-04 Kr-85m 5.924E-08 Kr-87 7.462E-20 Kr-88 3.053E-1 0 Xe-131 m 1.132E-04 Xe-133m 3.81 OE-04 Xe-1 33 1.664E-02 Xe-1 35m 1.627E-06 Xe-1 35 2.190E-04 Page 17 of 21

Worker Dose Calculation The dose calculation for a worker located outside the containment hatch following an FHA within containment was performed using the following parameters:

Parameter Value Comment Exposure time 30 minutes Conservative estimate of the time needed to

(= 1800 sec) manually close the containment structure Iequipment hatch shield doors.

Breathing rate 3.47E-04 m3/sec 0-2 hour control room breathing rate from I__

I__

Table 3.10-1 of the FHA AOR (Ref. 5.3).

Thyroid inhalation, beta skin, and whole body gamma immersion doses were then calculated as follows:

Dthy, I = Ci * (DCFthy, d) BR - te Dskin, I =

C1 * (DCFskin, i) te DWbI

=

C; * (DCFwbI) *te where:

Dthy I is the thyroid inhalation dose due to isotope i (Rem)

Dskin, i is the beta skin dose due to isotope i (Rem)

Dwb, i is the whole body gamma dose due to isotope i (Rem)

CI is the concentration of isotope i outside containment (CUm 3)

DC Fthy, I is the thyroid inhalation dose conversion factor for isotope i (Rem/Ci)

DCFskin, I is the beta skin dose conversion factor for isotope i (Rem-m /Ci-sec)

DCFWb, I is the whole body gamma dose conversion factor for isotope i (Rem-m 3/Ci sec)

BR is the worker breathing rate (m3/sec) te is the exposure time (sec)

The dose conversion factors (DCFs) used in the analysis were taken from Table 4.8-1 of the FHA AOR (Ref. 5.3). The results are shown in Table 5, Table 6 and Table 7.

Page 18 of 21

Table 5 30 Minute Thyroid Inhalation Dose Outside Containment Activity Thyroid 30 Minute Isotope Concentration DCF Thyroid Dose Outside Containment

______iC/rn)

(Rem/Ci) 1(Rem) 1-129 5.407E-1 2 5.92E+06 2.00E-05 1-130 1.408E-09 7.40E+04 6.51 E-05 1-131 6.389E-05 1.07E+06 4.27E+01 1-132 3.905E-14 6.29E+03 1.53E-10 1-133 1.491 E-05 1.81 E+05 1.69E+O0 1-135 8.188E-08 3.15E+04 1.61 E-03 Kr-83m 2.668E-1 5 0

O.OOE+00 Kr-85 5.630E-04 0

0.OOE+00 Kr-85m 5.924E-08 0

O.OOE+O0 Kr-87 7.462E-20 0

O.OOE+00 Kr-88 3.053E-10 0

O.OOE+00 Xe-131 m 1.132E-04 0

O.OOE+00 Xe-1 33m 3.81 OE-04 0

O.OOE+00 Xe-1 33 1.664E-02 0

O.OOE+00 Xe-135m 1.627E-06 0

O.OOE+00 Xe-1 35 2.190E-04 0

O.OOE+00 Total Dose 44.4 Table 6 30 Minute Beta Skin Dose Outside Containment Activity Beta Skin 30 Minute Isotope Concentration DCF Beta Skin Dose Outside Containment

______(l/Mn)

(Rem-m3/Ci-sec)

(Rem) 1-129 5.407E-12 3.710E-04 3.61 E-12 1-130 1.408E-09 4.990E-02 1.26E-07 1-131 6.389E-05 3.170E-02 3.65E-03 I-132 3.905E-14 1.320E-01 9.28E-12 1-133 1.491 E-05 7.350E-02 1.97E-03 1-135 8.188E-08 1.290E-01 1.90E-05 Kr-83m 2.668E-15 O.OOOE+00 O.OOE+00 Kr-85 5.630E-04 4.246E-02 4.30E-02 Kr-85m 5.924E-08 4.626E-02 4.93E-06 Kr-87 7.462E-20 3.083E-01 4.14E-17 Kr-88 3.053E-10 7.51 OE-02 4.13E-08 Xe-131 m 1.132E-04 1.508E-02 3.07E-03 Xe-133m 3.81 OE-04 3.150E-02 2.16E-02 Xe-1 33 1.664E-02 9.697E-03 2.90E-01 Xe-135m 1.627E-06 2.253E-02 6.60E-05 Xe-1 35 2.190E-04 5.894E-02 2.32E-02 Total Dose 0.4 Page 19 of 21

Table 7 30 Minute Whole Body Gamma Dose Outside Containment Activity WWB Gamma 30 Minute Isotope Concentration DCF WB Gamma Dose Outside Containment Rm 31isc(e)

______iC/rn)

(Rem-m 3/Ci-sec)

(Rem) 1-129 5.407E-12 3.024E-03 2.94E-1 1 1-130 1.408E-09 4.980E-01 1.26E-06 1-131 6.389E-05 8.720E-02 1.OOE-02 1-132 3.905E-14 5.130E-01 3.61 E-11 1-133 1.491 E-05 1.550E-01 4.16E-03 1-135 8.188E-08 4.21 OE-01 6.20E-05 Kr-83m 2.668E-15 2.396E-06 1.15E-17 Kr-85 5.630E-04 5.102E-04 5.17E-04 Kr-85m 5.924E-08 3.708E-02 3.95E-06 Kr-87 7.462E-20 1.876E-01 2.52E-1 7 Kr-88 3.053E-10 4.658E-01 2.56E-07 Xe-131 m 1.1 32E-04 2.899E-03 5.91 E-04 Xe-1 33m 3.81 OE-04 7.954E-03 5.45E-03 Xe-1 33 1.664E-02 9.316E-03 2.79E-01 Xe-1 35m 1.627E-06 9.887E-02 2.90E-04 Xe-135 2.190E-04 5.736E-02 2.26E-02 Total Dose 0.3 In addition to the above, a Total Effective Dose Equivalent (TEDE) to a crew member outside the equipment hatch was calculated by:

TEDE = (thyroid dose)(thyroid weighting factor) + (whole body dose)

A thyroid weighting factor of 0.03 was used per Regulatory Guide 1.183 Footnote 7 (Ref. 5.2) and the resulting dose to a crew member outside the equipment hatch was then calculated to be:

TEDE = (44.4 Rem)(0.03) + (0.3 Rem) = 1.63 Rem.

RESULTS AND CONCLUSION Federal regulation 10 CFR 50.47(b)(1 1) (Ref. 5.4) requires that "means for controlling radiological exposures, in an emergency, are established for emergency workers. The means for controlling radiological exposures shall include exposure guidelines consistent with EPA Emergency Worker and Lifesaving Activity Protective Action Guides." Per the EPA 400-R-92-001 (Ref. 5.5) Table 2-2 "Guidance on Dose Limits for Workers Performing Emergency Services," the dose limit for all activities is 5 Rem TEDE. Per a Table 2-2 footnote, workers performing services during emergencies should limit the doses to the skin to ten times the listed value (i.e., 50 Rem).

Page 20 of 21

The calculated TEDE of 1.63 Rem and the beta-skin dose of 0.4 Rem meet the dose criteria discussed above. It is therefore concluded that, following an FHA inside containment, workers outside containment who are manually closing the containment structure equipment hatch shield doors will not receive radiation exposures in excess of limits.

REFERENCES 5.1 SCE letter to the USNRC dated August 4, 2003, San Onofre Nuclear Generating Station Units 2 and 3 Docket Nos. 50-361 and 50-362 Proposed Change Number (PCN) 534 Request to Revise Technical Specification 3.9.3, "Containment Penetrations" 5.2 USNRC Regulatory Guide 1.183, July 2000, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors 5.3 Calculation N-4072-003, Rev. 4, Fuel Handling Accident (FHA) Inside Containment - Control Room & Offsite Doses 5.4 Title 10 of the Code of Federal Regulations, Part 50, Domestic Licensing of Production and Utilization Facilities 5.5 EPA 400-R-92-001, Manual of Protective Action Guides and Protective Actions for Nuclear Incidents Page 21 of 21

Appendix A Diffusion Analysis This appendix characterizes the radioisotope diffusion of airborne krypton, xenon, and iodine radioisotopes from inside containment to the outside environment to allow estimation of the radiation exposure to a worker outside the containment equipment hatch.

Molecular Diffusion The mass flux of an isotope in air (N1) is given by Fick's law of diffusion for a gradient in mass concentration:

Ni =-DAB dC.j, in units of Curies/(cm 2_sec)

Equation A-1 dx (based on Bird, et al., Reference Al, Equation 16.2-3, p. 503) where DAB is the mass diffusivity of material A (radioisotope) in B (air), in units of cm2/sec is the concentration gradient of a specific isotope, in units of (Curies/cm 3)/cm.

dx The curie concentration is directly proportional to the mass concentration.

x is the distance, in cm. In this analysis, x is taken as the depth of the containment hatch opening, including the steel sleeve.

The flux will be maximized if it is assumed that the concentration approaches zero at the containment exterior, such as would occur if the air outside containment continually sweeps away the activity released through the hatch opening. Maximizing the flux N, maximizes the concentration gradient dC/dx therefore, the concentration gradient can be'approximated as:

,where Co is the concentration of isotope i inside containment.

dx x

Therefore, the maximized mass flux can be restated as:

NEX =-DAB ( C> =DAB Ci Equation A-2 x

x Mass Diffusivity The mass diffusivity DAB is calculated in accordance with the Chapman-Enskog kinetic theory as given in Equation 16.4-13 of Bird, et al. (Reference Al, p. 51 1):

DA =0.0018583 Equation A-3 P CAB 2D,AB Page A-1

where DAB

=

mass diffusivity of material A in B in units of cm2/sec T

= temperature in 0K MANB

=

molecular weight of substance A or B p

=

total pressure in atmospheres aAB

=

Lennard-Jones parameter, in Angstroms 2D, AB =

a dimensionless function of the temperature and of the intermolecular potential field for one molecule of A and one molecule of B.

Table B-2 of Bird, et al. (Reference Al, p. 746) provides values for QD, AB as a function of a term ICTIEAB-Table B-1 of Bird, et al. (Reference Al, pp. 744-745) provides values of the molecular weight and the Lennard-Jones parameters a and e/lK for various substances. The applicable values are shown in Table 8.

Table 8 - Intermolecular Force Parameters Substance Molecular Lennard-Jones Parameters Weight M

a (A)

E/K (0 K)

Kr 83.80 3.61 190 Xe 131.3 4.055 229 12 253.82 4.982 550 Air 28.97 3.617 97.0 Per Equation 16.4-15 of Bird, et al. (Reference Al, p. 51 1), GAB can be estimated by:

CAB 2 (CA + CB )

Equation A-4 Per Equation 16.4-16 of Bird, et al. (Reference Al, p. 51 1), EAB can be estimated by:

6 AB Equation A-5 The applicable values of the CAB and KT/EAB parameters can therefore be calculated using Table 8, Equation A-4, and Equation A-5. Then, assuming an ambient temperature of 20 0C (i.e.,

293 OK), the values for QD AB are determined by interpolating the lookup values from Table B-2 of Reference Al. The results are provided in Table 9.

Table 9 - Determination of Omega Diffusion Terms CAB E/1KAB l 1T/EAB lQD, AB System (A)

(OK) l l

Kr - Air 3.614 135.8 2.16 1.047 Xe - Air 3.836 149.0 1.97 1.080 12 - Air 4.300 231.0 1.27 1.287 Page A-2

The mass diffusivities of each isotope can now be calculated using Equation A-3.

DXr-Air = 0.0018583 1.047

= 0.147 cm2/sec 1I x 3.614 2 x 107 293' (

1

+

I DXe-Air= 0.0018583 (1 x3.8362 x1.08X) 0120 cm2/sec j2933 (23

+

I D12-Air =0.0018583

.532 1

97) = 0-0768 cm2/sec (I x4.32 x 1.287)

Maximized Radioisotope Diffusion Rates To convert the mass flux for each radioisotope to a release rate, the flux is multiplied by the clear area of the equipment hatch opening. Thus, n, = Nx A, where mj is the release rate and A is the clear hatch opening area, in cm2.

Therefore, based on Equation A-2,

_ DAB A Cin cm3 Ci Equation A-6 C in -.-

3 x

sec cm Area of Opening The containment equipment hatch penetration is a circular opening of radius of 9'-6", with the bottom 11/2 feet filled in with concrete to create a level floor at elevation 30'-0" (Dwgs. 23025, Ref. A2, and 23063, Ref. A4). Thus, the area of the opening is given by:

A° =iR 2 -[R2cos[ ( R ) - (R-h) 2 Rhlh2]

= r9.52

[9.52 cos-'(

9 5 J -

(9.5-1.5) F(2-9.5-1.5)-1.52]

= r90.25 - [90.25cos- (.-5) - (8)JI].25 = 273.1 ft2, or - 2.537 x 105 cm2 Hatch Opening Depth Per Dwg. 23054 (Ref. A3) section B, the depth of the containment equipment hatch opening from the edge of the flange inside containment to the containment shell outer face is 7'-11/h", or 217 cm.

Page A-3

Diffusion Rates Thus, using Equation A-6 and the mass diffusivities given above, the diffusion rates for each radioisotope are calculated as follows:

0.147*(2.537x1O5) Co =172COr Ci/sec Kr

~217Krr

  • 0.120-(2.537x1O0)

=

Xe 217 CKr =140CX¢ Cl/sec 0.0768 (2.537X1O5)C, =90CO C~sec 12 =217 Kr 9

12l/e Results Summary The diffusion rates of airborne krypton, xenon, and iodine radioisotopes from inside containment to the outside environment are summarized below.

Element Diffusion Rate (Ci/sec)

Krypton 172 CKrO Xenon 140 Cxe0 Iodine 90C,20 Using this information, the activity concentration outside containment can be estimated from the isotopic concentrations inside containment.

References Al. Bird, R. Byron, Stewart, Warren E., and Lightfoot, Edwin N., Transport Phenomena. John Wiley & Sons, Inc., New York, 1960.

A2. Units 2 & 3 Drawing 23025, Rev. 7, Containment Structure Reinforced Concrete Wall Sections & Details, Sht. 3 A3. Units 2 & 3 Drawing 23054, Rev. 5, Containment Structure Wall Liner & Inserts Sections &

Details Sh. 3 A4. Units 2 & 3 Drawing 23063, Rev. 9 (including DCNs 4 and 5), Containment Structure Hatches & Locks Page A-4

Appendix B Air Volume Between Shield Doors and Containment This appendix determines the volume of air between the shield doors and the containment outer wall. This volume is used to estimate the radionuclide concentrations to which a worker outside the equipment hatch would be exposed following a fuel handling accident inside containment.

The figure below is a plan view of the shield door/containment arrangement. This figure is not to scale.

The air volume to be calculated has a height of 19'5" and is defined by the six regions (circled numbers) as follows:

Volume = (Area 1 + Area 2 + Area 3 - Area 4 + Area 5 - Area 6) (19.42 ft)

By geometry:

Area 1 = (0.42 ft) (18.75 ft) = 7.9 ft2 Area 2 = (0.42 ft) (24.25 ft) = 10.2 ft2 Area 3 = (0.5) (18.75 ft) (2.66 ft - 0.42 ft) = 21.0 ft2 Area 5 = (0.5) (24.25 ft) (4.21 ft - 0.42 ft) = 46.0 ft2 Page B-1

To calculate Area 4 and Area 6, the angles a and P3 are needed. By geometry:

a = sin" (18.75 ft /79.33 ft) = 13.670 D = sin"' (24.25 ft /79.33 ft) = 17.800 The areas of the sectors enclosed by angles a and P are given by:

Sector a = (13.670 / 3600) (rr) (79.33 ft)2 = 750.8 ft2 Sector a = (17.80° / 3600) (Tr) (79.33 ft)2 = 977.6 ft2 The portion of Sector a that is not Area 4 is shown in the figure below (also not to scale).

Subtracting the area of this region from the area of Sector a will give Area 4.

By geometry:

sin (a/2) = L /79.33 ft With a = 13.670, La = 9.44 ft. By the Pythagorean Theorem:

Ha = [(79.33 ft)2 -(9.44 ft)2]05 = 78.77 ft The two halves of the region are equal, so the area of the whole a triangle shown above is then:

Area of a triangle = (2) (0.5) (9.44 ft) (78.77 ft) = 743.6 ft2 Subtracting this from the area of Sector a, gives Area 4:

Area 4 = 750.8 ft2 - 743.6 ft2 = 7.2 ft2 By a similar method, Lp = 12.27 ft, Hp = 78.38 ft, and the area of the associated p triangle is:

Area of P triangle = (2) (0.5) (12.27 ft) (78.38 ft) = 961.7 ft2 Subtracting this from the area of Sector f3, gives Area 6:

l Area 6 = 977.6 ft2 - 961.7 ft2 = 15.9 ft2 The air volume between the containment and the shield doors is:

Volume

(7.9 ft2 + 10.2 ft2 + 21.0 ft2 - 7.2 ft2 + 46.0 ft2 - 15.9 ft2) (19.42 ft)

1204 ft3 Using a conversion factor of 0.0283 m3/ft3, gives:

I Volume = 34 m 3 I

G:\\Work\\Plic\\Brough\\PCN534Sup1 Encl2.doc Page B-2

ENCLOSURE 3 Revised Proposed Technical Specification Pages

PCN 534 Attachment A (Proposed TS Page)

(Redline and Strikeout)

SONGS Unit 2

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 closed and held in place by four

-,~bolts:

  • d

.,?,

The equipment hatch may be open if all of the following conditions are met:

[1)

The Containment Structure Equipment Hatch Shield Doors are capable of being closed within 30 minutes,

2)

FThe plantJis in Mode 6'with-at least 23 feet-of water above the reactor ves'sel flanqe,

3) A desiqnated c'rew is available to close the Containment e

Structure Equipment'Hatch Shield Doors,

'4) Containment'purge is-in service, and

5) jThe reera'ctor has heen subcritical for at l1east 72 hoiurs'
b. One door in each air lock closed;

NOTE-----------------------------

Both doors of the containment personnel airlock may be open provided:

a.

one personnel airlock door is OPERABLE, and bl.

the plant is in MODE 6 with 23 feet of water above the fuel in the reactor vessel, or b2.

defueled configuration with fuel in containment (i.e., fuel in refueling machine or upender).

c. Each penetration providing direct access from the containment atmosphere to the outside atmosphere shall be 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 System.

APPLICABILITY:

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

SAN ONOFRE--UNIT 2 3.9-4 Amendment No. +i-T

PCN 534 Attachment B (Proposed TS Page)

(Redline and Strikeout)

SONGS Unit 3

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 closed and held in place by four Ubolts:-___

s..-r A

-- NOTE-------------_-__-_-__-_-_

The equipment- -hatch may be open if-;all of'the following conditions are 'met: _

1)

The Containment Structure Equipment Hatch Shield Doors are capable-of being closed within 30 minutes,---

2 The plant is in Mode 6 with at least 23 feet of-water Above the reactor vessel flange,

3)

A-designated crew is available toiclose the Containment Structure Equipment Hatch Shield.Doors, i) Containment purge is in service;' and_.

5)

The reactor' has been subcritical for-at least 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

b.

One door in each air lock closed;

NOTE-----------------------------

Both doors of the containment personnel airlock may be open provided:

a.

one personnel airlock door is OPERABLE, and bl. the plant is in MODE 6 with 23 feet of water above the fuel in the reactor vessel, or b2.

defueled configuration with fuel in containment (i.e., fuel in refueling machine or upender).

c. Each penetration providing direct access from the containment atmosphere to the outside atmosphere shall be 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 System.

APPLICABILITY:

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

SAN ONOFRE--UNIT 3 3.9-4 Amendment No. -16I l

PCN 534 Attachment C (Proposed TS Page)

SONGS Unit 2

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 closed and held in place by four bolts;

NOTE-----------------------------

The equipment hatch may be open if all of the following conditions are met:

1) The Containment Structure Equipment Hatch Shield Doors are capable of being closed within 30 minutes,
2) The plant is in Mode 6 with at least 23 feet of water above the reactor vessel flange,
3) A designated crew is available to close the Containment Structure Equipment Hatch Shield Doors,
4) Containment purge is in service, and
5) The reactor has been subcritical for at least 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
b. One door in each air lock closed;

NOTE--------------

Both doors of the containment personnel airlock may be open provided:

a. one personnel airlock door is OPERABLE, and bl.

the plant is in MODE 6 with 23 feet of water above the fuel in the reactor vessel, or b2.

defueled configuration with fuel in containment (i.e., fuel in refueling machine or upender).

c. Each penetration providing direct access from the containment atmosphere to the outside atmosphere shall be 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 System.

APPLICABILITY:

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

SAN ONOFRE--UNIT 2 3.9-4 Amendment No.

PCN 534 Attachment D (Proposed TS Page)

SONGS Unit 3

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 closed and held in place by four bolts;

NOTE-----------------------------

The equipment hatch may be open if all of the following conditions are met:

1) The Containment Structure Equipment Hatch Shield Doors are capable of being closed within 30 minutes,
2)

The plant is in Mode 6 with at least 23 feet of water above the reactor vessel flange,

3) A designated crew is available to close the Containment Structure Equipment Hatch Shield Doors, 43 Containment purge is in service, and
5) The reactor has been subcritical for at least 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
b. One door in each air lock closed;

NOTE-----------------------------

Both doors of the containment personnel airlock may be open provided:

a. one personnel airlock door is OPERABLE, and bl.

the plant is in MODE 6 with 23 feet of water above the fuel in the reactor vessel, or b2.

defueled configuration with fuel in containment (i.e., fuel in refueling machine or upender).

c. Each penetration providing direct access from the containment atmosphere to the outside atmosphere shall be 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 System.

APPLICABILITY:

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

SAN ONOFRE--UNIT 3 3.9-4 Amendment No.

PCN 534 Attachment E (Proposed TS Bases Pages)

(Redline and Strikeout)

SONGS Unit 2 (Typical for both Units)

Containment Penetrations B 3.9.3 BASES (continued)

LCO This LCO limits the consequencesof 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 and the containment personnel airlock.

For the containment personnel airlock, this LCO ensures that the airlock can be closed after containment evacuation in the event of a fuel handling accident. The requirement that the plant be in Mode 6 with 23 feet of water above the fuel in the reactor vessel or defueled configuration with fuel in the containment (i.e., fuel in the refueling machine or upender) ensures that there is sufficient time to close the personnel airlock following a loss of shutdown cooling before boiling occurs.

LCO part a. is modi fied by'a NOTE:

NOTE-------------

,The -equipment hatch 'may-be open if all:of the following

,conditions are.met:

1) tThe Containment Structure Equipment Hatch Shield Doors are capable of; being closed 'within 30 minutes,
2) The plant-is in-Mode 6 with at least 23 feet of water above the reactor vessel flange,,
3) A designated crew is available to close the _

Containment Structure Equipment Hatch Shield Doors,

4)

Containment purge is-in service,-and

5) The reactor'has been subcritical_ for at-least 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.'

These restrictions include the administrative-controls to tallow the opening of the-containment.equipment hatch-during CORE ALTERATIONS or. movement of-irradiated fuel in the

'containment provided that 1)The Containment.Structure:

,Equipment.Hatch Shield Doors capable-of being closed within 30 minutes, 2) The plantis.in Mode 6 with at least 23-feet bf water above the reactor.,vessel flange, 3) A:designated crew is available to close the Containment-:Structure Equipment Hatch Shield Doors, 4) Containment purge is in rservice, and 5) The reactorshall be subcritical-for-at least 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />..-The Containment Structure Equipment Hatch (continued)

SAN ONOFRE-UNIT 2 B 3.9-13 Amendment No. 127 10/I0/97 l

Containment Penetrations B 3.9.3 BASES (continued)

LCO (continued)

Shield Doors.:include flashing-on the top and'sides of the shield 'doors 'which act'toretard or-'restrict' a release of _,

post-accident fission products.-:.The capability.to close the containment shield doors.includes..requirements that the tdoors-are capable of'beinq closed and that-any cables or'

'hoses across-the opening have quick disconnects to ensure Xhe'doors are-capable'of being closed within 30 minutes..

~The 30 minute closure time for'the conta'inment'. shield doors is considered to start when the control room communicates the need to shut the:-Containment Structure Equipment'Hatch Shield Doors. :'This,30-min'ute requirement is-significantly Nless than the fuel handlinq accident analysis assumption that the'containment remains open to the outside environment for a two-hour'period subsequent to the accident. Placing containment.purge'f(either main purge or~mini purge)in service will ensuret.anyrelease from containment will be monitored.

rrhe-administrative controls.will-also include the responsibility to be able-to'communicate-with the control room,-and the-responsibility to ensure that the containment shield doors are capable of being closed in the event of a fuel handling -accident.

These administrative controls will ensure containment'closure would be'established in.the event of-a fuel handling accident inside containment.

This LCO p modified by a NOTEete which allows.4-ft keep both doo6rs of the containment 'pernnel airlock,to be open provided:

a. one personnel airlock door is OPERABLE, and b.1 the plant is in MODE 6 with 23 feet of water above the fuel in the reactor vessel, or b.2 defueled configuration with fuel in containment (i.e., fuel in refueling machine or upender).

(continued)

SAN ONOFRE-UNIT 2 B 3.9-13a Amendment No.

I

Containment Penetrations B 3.9.3 BASES (continued)

LCO (continued)

The OPERABILITY requirements ensure that the airlock door is capable of performing its function, and that a designated individual located outside of the affected area is available to close the door.

For the OPERABLE containment purge and exhaust penetrations, this LCO ensures that these LCO penetrations are isolable by the Containment Purge Isolation System.

The OPERABILITY requirements for this LCO ensure that the automatic purge and exhaust valve closure times specified in the UFSAR can be achieved and therefore meet the assumptions used in the safety analysis to ensure releases through the valves are terminated, such that the radiological doses are within the acceptance limit.

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, (continued)

SAN ONOFRE--UNIT 3 B 3.9-13b Amendment No.

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