ML040020042
| ML040020042 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 12/24/2003 |
| From: | Scherer A Southern California Edison Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| TAC MC0317, TAC MC0318 | |
| Download: ML040020042 (127) | |
Text
SOUTHERN CALIFORNIA A. Edward Scherer EDISONa Manager of Nuclear Regulatory Affairs An EDISON INTERNATIONALS Company December 24, 2003 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, D.C. 20555
Subject:
Docket Nos. 50-361 and 50462 Response to Request for Additional Information regarding Containment Equipment Hatch (TAC Nos. MC0317 and MC0318)
San Onofre Nuclear Generating Station Units 2 and 3
Dear Sir or Madam:
Enclosure I provides Southern California Edison's (SCE) response to a Request for Additional Information from the NRC staff preliminary review of proposed change number 534, dated November 7, 2003. In this amendment proposal, SCE requested to revise Technical Specification 3.9.3, Containment Penetrations." Specifically, the proposed changes would permit the Containment equipment hatch to be open during core alterations and movement of irradiated fuel in containment, with certain precautionary provisions in place.
If you have any question or require additional information, please contact Mr. Jack Rainsberry at 949-368-7420.
Sincerely, Enclosure 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 & 3 P.O. Box 128 San Clemente, CA 92674-0128 949-368-7501 Fax 949-368-7575
RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (RAI)
SOUTHERN CALIFORNIA EDISON COMPANY SAN ONOFRE NUCLEAR GENERATING STATION (SONGS), UNITS 2 AND 3 CONTAINMENT STRUCTURE EQUIPMENT HATCH SHIELD DOORS DOCKET NOS. 50-361 AND 50-362 By letter dated August 4, 2003, Southern California Edison (the licensee) submitted proposed amendments to Technical Specifications (TSs) for SONGS, Units 2 and 3. The proposed amendments requested to revise the TSs to permit the Containment equipment hatch to be open during core alterations and movement of irradiated fuel in containment.
To support this proposal, the licensee performed a re-analysis of the design basis Fuel Handling Accident (FHA) at SONGS, Units 2 and 3.
The Nuclear Regulatory Comnission (NRC) staff reviewed the information provided by the licensee to support the proposed TS changes. In order for the NRC staff to complete its evaluation, the following additional information was requested by letter dated November 7, 2003. The following includes SCE's response:
- 1.
What design bases parameters, assumptions or methodologies (other than those provided in the August 4, 2003 submittal), were changed in the radiological design basis accident analyses as a result of the proposed change? If there are many changes, it would be helpful to compare and contrast them in a table format. Also, please provide justification for any changes.
Response
The fuel handling accident inside containment (FHA-IC) dose analysis of record (AOR) is Calculation N-4072-003, Revision 4. For PCN-534, the only change to this AOR was an increase in the control room unfiltered inleakage rate from 10 to 1000 cfm total (i.e., 990 cfm inleakage plus 10 cfm for ingress and egress).
In addition to this change, it is noted that a minor difference exists between the FHA-IC dose AOR and the control room emergency air cleanup system (CREACUS) description provided in the UFSAR Appendix 15.10B Table 15.10B-1. UFSAR Table 15.101-1 states that two trains of CREACUS have a combined flow rate of 59,869 cflm.
Calculation N-4072-003 Revision 4 Design Input 4.4.3 models a flow rate of 59,870 cfm (representing twice the single train flow rate). All other design inputs, assumptions and methodology are consistent with the description provided in UFSAR Section 15.10.7.3.9 and its referenced Appendices 15B and 15.10B.
Page 1 of 13
RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (RAI)
CONTAINMENT STRUCTURE EQUIPMENT HATCH SHIELD DOORS
- 2.
Based on a preliminary review of the FHA for the proposed changes, the reviewer is unable to match the calculated doses. The staff requests that the licensee provide their design bases fuel handling calculations. (Note: Answers to other questions in this RAI may also reference these calculation once submitted.)
Response
The FHA-IC dose AOR is Calculation N-4072-003, Revision 4. A paper copy of this calculation is enclosed.
- 3.
What types of hoses and cables will be allowed to pass through the open equipment hatch? What provision will be made for the designated individual to separate these to close the air lock door, while reducing the hazards of these hoses and cables?
Response
Per Integrated Operating Instruction S023-5-1.8.1, the following requirements must be met for all hoses and cables allowed to pass through the open equipment hatch:
Configured with isolations and points of disconnect immediately adjacent to the hatch, Points of disconnect Caution Tagged and logged in the Tag Log Index identifying the service and Responsible Division, and Checked shiftly.
- 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.
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.
Page 2 of 13
RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (RAI)
CONTAINMENT STRUCTURE EQUIPMENT HATCH SHIELD DOORS 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.
- 5.
The proposed TSs changes specify that a "designated" crew is available to close the Containment Structure Equipment Hatch Shield Doors rather than a "dedicated" crew who would have no other duties. Specify what other duties the designated crew will have and where they will be stationed relative to the air lock doors.
Response
The "designated crew" will be part of the routine crew used to onload and offload equipment from containment. The requirements for this crew will be specified in administrative procedures. They may continue to perform equipment on-load and off-load through the equipment hatch which may require other workers. The crew will be "stationed" in the vicinity of the open equipment hatch. This may include the area as far away as the opposite unit equipment hatch where equipment is stored waiting to go in containment or place equipment being moved out of containment. These storage areas are within a few hundred feet of the open equipment hatch and the time it takes to respond from these locations will not significantly impact the time it takes to achieve closing of the missile shield doors.
- 6.
Provide a detailed account of the timing and flow rates, and filtration of the control room HVAC [heating, ventilation, and air conditioning] as it responds to an FHA. A schematic would be helpful.
Page 3 of 13
RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (RAI)
CONTAINMENT STRUCTURE EQUIPMENT HATCH SHIELD DOORS
Response
As indicated in the response to RAI #1, the only change to the FHA-IC dose AOR was an increase in the control room unfiltered inleakage rate from 10 to 1000 cfm total. The control room HVAC timing, flow rates and filtration model described in UFSAR Appendix 15B (Table 15B-5) and Appendix 15.10B (Table 15.101-1) are incorporated into FHA-IC Calculation N-4072-003, Revision 4. Section 8.6 of the calculation provides a time line showing the switchover from control room normal HVAC operation to control room emergency KVAC operation. Design Input 4.4 of the calculation presents the control room normal and emergency HVAC characteristics, including flow rates and filter efficiencies.
- 7.
Please provide engineering drawings of the proposed change. A photograph of the equipment hatch would also be helpful in the review of this proposed change.
Describe the steps taken to ensure that the proposed flashing will not interfere with closure of the shield doors. What is the acceptable design clearance between the flashing on the shield doors and the containment?
Response
Photographs and drawings of the missile shield doors and flashing are attached. The design of the flashing ensures there is no impediment to missile shield door operation. The flashing consists of stainless steel sheet metal anchored to the missile shield door with a 1 inch minimum gap to the containment wall. A neoprene rubber strip is attached to the stainless steel flashing and effectively fills the 1 inch gap. (Reference drawing 23070, Sheet 2). The curvature of the containment also allows the flashing to separate from the wall as the missile shield doors are rolled opened and will limit leakage when the doors are dosed, where the flashing is provided. The design does not provide a positive seal at the timber sections, the base of the doors, or at the joint where the doors meet. The largest gap is expected to be less than 1/2 inch at these locations and would therefore minimize leakage. As such, the acceptable design clearance is 1/2 inch between the missile shield doors and the containment structure.
- 8.
Provide the criterion used to decide if the Equipment Hatch Shield Doors are capable of being closed within 30-minutes.
Response
At the conclusion of the most recent Unit 3 refueling outage, on February 12, 2003, with this proposed License Amendment Request in mind, Engineering and Licensing personnel observed Maintenance Page 4 of 13
RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (RA)
CONTAINMENT STRUCTURE EQUIPMENT HATCH SHIELD DOORS personnel dose the missile shield doors. This action was completed within 30-minutes under extremely adverse conditions: during a rainstorm and without prestaging the manual chainfalls used to close the missile shield doors.
If more restrictive requirements are identified for the performance of the Shield Doors' closure, the closure time will be re-verified before use.
- 9.
Provide the Low Population Zone and Beta doses consistent with the information provided in current UFSAR [Updated Final Safety Analysis Report].
Response
The following table summarizes the control room and offsie dose consequences as currently documented in UFSAR Table 15.10.7.3.9-3 and as calculated in FHA-IC Calculation N-4072-003, Revision 4 (Section 2.1 Table 2.1-1). Per the table, the offsite doses remain unchanged. This is consistent with the response to RAI #1, which noted that the only change to the FHA-IC dose AOR was an increase in the control room unfiltered inleakage rate from 10 to 1000 cfm total.
DOSE CALCULATED CURRENT DOSE RECEPTOR CRITERIA DOSE UFSAR DOSE (REM)
(REM)
(REM)
Control Room (30-day dose)
Thyroid 30 25.4 14.8 Beta Skin 30 3.6 3.6 Whole Body Gamma 5
0.3 0.2 EAB (2-hour dose)
Thyroid 75 56.4 56.4 Beta Skin No dose criterion 0.3 0.3 Whole Body Gamma 6
0.3 0.3 LPZ (30-day dose)
Thyroid 75 1.6 1.6 Beta Skin No dose criterion
< 0.1
< 0.1 Whole Body Gamma 6
< 0.1
< 0.1 Page 5 of 13
RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (RAI)
CONTAINMENT STRUCTURE EQUIPMENT HATCH SHIELD DOORS
- 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?
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 Pe taken to either secure the Containment Hatch or Shield Doors or Core Alteration/Movement of irradiated Fuel would be secured.
- 11.
There appears to be inconsistencies between values used in the UFSAR and the values provided in Table I of the licensee's submittal, without a provided justification for the changes. Please verify the parameter provided in Table I against those in the analysis utilized to justify this amendment request, and provide a justification for the changes in values from those previously accepted.
Response
As indicated in the responses to RAI #1 and #6, the only change to the FHA-IC dose AOR was an increase in the assumed control room unfiltered inleakage rate from 10 to 1000 cfm total.
The control room HVAC filtration model described in UFSAR Appendix 15B (Table 15B-5) and Appendix 15.10B (Table 15.10B-1) is incorporated into FHA-IC Calculation N4072-003, Revision 4. Design Input 4.4 of the calculation presents the control room normal and emergency HVAC characteristics, including filter efficiencies. As shown in the following figure, in the high radiation isolation mode, the CR outside makeup air passes through intake filters SA1510MA206 and SAI 51 0MA207 of the emergency ventilation system (EVS) train and then through recirculation filters SAl 51 0ME418 and SAI 51 0ME419 of the emergency air conditioner (EAC) filter train.
Page 6 of 13
RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (RAI)
CONTAINMENT STRUCTURE EQUIPMENT HATCH SHIELD DOORS Credit is taken only for outside makeup and recirculated control room air iodine and particulate removal by the EAC filters. Credit is not taken for outside air iodine and particulate removal by the EVS filters.
OUTSIDE AIR
- old EVS FILTERS
[not credited]
- _10 EAC
- =0 FILTERS CONTROL
[credited
-=
ROOM
- 12.
General Design Criterion 64 of 10 CFR Part 50, Appendix A, states that means shall be provided for monitoring the reactor containment atmosphere, spaces containing components for recirculation of loss-of-coolant accident fluids, effluent discharge paths, and the plant environs for radioactivity that may be released from normal operations, including anticipated operational occurrences, and from postulated accidents. The proposed change should consider how Criterion 64 will be met in the event of a FHA with the equipment hatch open. Moreover, this information should be included as part of the Bases discussion.
Provide the bases for meeting Criterion 64 for the proposed change. Please confirm that your emergency planning dose assessment methodology includes the ability to assess this accident. For example, does your methodology include the capability to determine the source term, release rate out of containment, meteorology and consider feedback via field monitoring health physics survey teams? Have you evaluated the need for any special radiological monitoring or survey equipment (i.e.,
in-plant equipment or field team survey equipment) to evaluate the radiological conditions of this accident scenario? Will your emergency response personnel be trained to deal with this accident scenario?
Response
There is an existing continuous radiation monitoring system on the discharge for containment purge from unit 2 and a similar monitoring system on unit 3's containment purge whenever the purge system is in service. Proposed Change Number 534 will be supplemented so that Technical Specification 3.9.3 and station procedures require containment purge to be in-service whenever fuel alterations (movement) are performed and the equipment hatch is open (i.e., this will be a fourth condition to enable the equipment hatch to be open). In the unlikely event that a FHA occurs in containment with the equipment hatch in its proposed open position, the initial release will be detected by the radiation monitor on containment purge, allowing us to estimate the source term in containment. We recognize that the FHA-related release may exceed the setpoint on the containment purge radiation monitoring system and initiate a CPIS, terminating the purge.
Page 7 of 13
RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (RAI)
CONTAINMENT STRUCTURE EQUIPMENT HATCH SHIELD DOORS When containment purge is isolated as a result of the FHA, the remaining release paths are either (1) through the personnel hatch into the penetration building which is kept at negative pressure with respect to ambient conditions and vented via the Plant Vent Stack, a monitored release point, or, (2) through the open equipment hatch. Because of the size of the opening, it is not practical to monitor the equipment hatch using an installed continuous radiation monitor detection system.
However, the existing emergency preparedness procedures already include techniques to determine the source term and public dose impact from an unmonitored release point. Similarly, the existing emergency planning dose assessment methodology includes provisions to estimate the release rate from a structure, and, using site meteorology, detemine the public dose impact from the release. The same emergency planning procedures and dose assessment software also provides back-calculation methods for comparing source term estimates with field readings from the health physics survey teams. No additional training is necessary for emergency response personnel to respond to an unmonitored release through the open equipment hatch following a FHA in containment.
- 13.
The proposed change states:
"With the proposed TS 3.9.3 changes, the crew tasked with closing the containment shield doors as a means of providing [for] containment closure will be working
[performing this activity from] outside containment." "Since containment is unlikely to become pressurized during an in-containment fuel handling accident during refueling, there is no motive force for airborne radioactivity to be propelled through the opening. As a result, the resultant dose to the crew is anticipated to be minimal."
Provide justification for the statement that "there is no motive force for airborne radioactivity," considering the motive force that may be caused by (1) in containment heat sources, (2) the pressure from external sources such as wind, or interfaces with pressurized buildings, or (3) heating of the containment by the sun.
Response
The amount of motive force for airborne radioactivity from interior containment heat sources, thermal gradient of the containment structure, and from other external forces such as wind and pressurized buildings was judged to be very small and characterized as 'no motive force." The basis for the characterization is explained below. The containment purge system will be required by the technical specifications to be in operation during fuel movement. As such, prior to any fuel handling accident there would be a negative air flow Page 8 of 13
RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (RA)
CONTAINMENT STRUCTURE EQUIPMENT HATCH SHIELD DOORS through the equipment hatch opening. In the event of an accident, the personnel lock would be under the same 30-minute closure time requirement. Negative air flow would be sustained by the auxiliary building penetration area until the personnel lock was closed and the purge system isolated. If the personnel lock and purge system are conservatively taken to be closed at time zero, there is only 30-minutes available for heat generated motive forces to cause a positive air flow through the equipment hatch.
Positive air flow could be created by either a significant rise in temperature inside containment with respect to the outside temperature or by a low pressure condition outside of containment caused by winds.
In general the outside weather, including wind has minimal impact on refueling operations because of the containment configuration. The containment equipment hatch access is at the 30 feet elevation and the refueling deck is located at elevation 63 feet - 6 inches. The refueling deck is located well above the top of the equipment hatch area with minimal direct line of sight visibility from the operating deck that further limits air flow at the refueling deck. However, air flow out of containment could still be affected by climatic conditions.
Potential Air Flow from Heat Sources The containment has a very large free volume of 2,305,000 cubic feet, minimum. This volume would have to be heated a few degrees from the time of the accident to cause a positive air flow. There are 3 types of heat sources: 1) the irradiated fuel assemblies, which are actively cooled; 2) solar heating of the thick concrete containment structure; and 3) internally generated heat sources from people, lighting, welding machines, electric motors, pumps, resistors, and transformers.
Since the irradiated fuel assemblies are being water cooled, there is no change to the containment interior temperature from this source. Even with a loss of cooling there is a time lag between the heating of the water and the containment environment such that any change in the temperature of containment atmosphere over the 30-minute interval would be negligible.
The thick concrete containment structure slows solar heating of the interior environment. The temperature change over a 30-minute interval would be very small considering the insulating properties of concrete and the overall containment air volume.
Page 9 of 13
RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (RAI)
CONTAINMENT STRUCTURE EQUIPMENT HATCH SHIELD DOORS Other heat sources may include people, lighting, the refueling machine, cranes, hoists, pumps, testing equipment, monitors, welding machines, etc. Many of these sources cease or are significantly reduced in the event of an evacuation of containment personnel. Motive air forces from electrical sources and from the sun are very small over a 30-minute duration. The interior containment temperature is expected to remain essentially constant for the first 30-minutes of a fuel handling accident. As such, there is minimal potential for outward air flow from these sources.
Potential Air Flow from Weather Air flow could result from a low pressure condition near the equipment hatch. This could be caused by high winds or a sudden drop in temperature. As stated in the UFSAR the site has marine type weather conditions. During most of the year, daytime heating of the land surface makes it warm relative to the Pacific Ocean. This thermal difference produces an onshore wind (sea breeze) that normally begins shortly after sunrise and lasts until after sunset. At night, the land cools, reversing the thermal gradient, and an offshore wind (land breeze) develops. The presence of the Pacific Ocean has a moderating influence over the temperatures at the site. The equipment hatch faces an easterly direction and is somewhat protected by the raised switchyard area from offshore and strong onshore breezes. As such, the potential for a sudden low pressure condition around the equipment hatch is considered low.
Therefore, the characterization of "no motive force" was established during the time required to close the missile shield doors.
- 14.
10 CFR 50.36 states that:
A technical specification limiting condition for operation of a nuclear reactor must be established for each item meeting one or more of the following criteria:
(B)
Criterion 2. A process variable, design feature, or operating restriction that is an initial condition of a design basis accident or transient analysis that either assumes the failure of or presents a challenge to the integrity of a fission product barrier.
Page 10 of 13
RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (RAI)
CONTAINMENT STRUCTURE EQUIPMENT HATCH SHIELD DOORS The licensee's proposed analysis utilizes an initial condition of 72-hours of fuel decay for the FHA. The proposed TS does not provide a limiting condition of operation for this initial condition. Please justify why this decay time does not meet Criterion 2 of 10 CFR 50.36 or modify the TS to include the decay time.
Response
The requirement for 72-hours of fuel decay prior to movement of irradiated fuel in the reactor vessel was originally contained within the Technical Specifications and relocated to Licensee Controlled Specification (LCS) 3.9.101 as part of conversion to the Combustion Engineering Standard Technical Specifications, NUREG 1432, Revision 0 (Note: This is also consistent with NUREG 1432, Rev. 2).
This conversion and relocation was approved by an NRC Issuance of Amendment and accompanying Safety Evaluation, dated February 9, 1996. Existing procedures are based upon the 72-hour decay time requirement; any modification to the LCS and affected procedures would require satisfactory results of a 10 CFR 50.59 evaluation or NRC staff approval would be required.
- 15.
Will your Emergency Plan be updated to include an accident release through the equipment hatch? Will your Emergency Operating Procedures be updated to address the specific details needed to respond to this accident scenario?
Response
The existing Emergency Plan addresses a dropped fuel assembly in procedure S0123-VIII-1; this has been reviewed and it has been concluded no further changes are required. The Operations Division Abnormal Operating Instructions (AOI) will be modified to close the equipment hatch missile shield upon dropping a fuel assembly.
- 16.
Will the State Emergency Response personnel be informed about this accident scenario?
Response
SCE has reviewed the procedures used by the State offsite responders (in SONGS case, this is the Interjurisdictional Planning Committee, the Office of Emergency Services, etc.) and concluded no changes to their response or procedures are required by this accident scenario.
Nevertheless, we will notify the State offsite responders of this Technical Specification change prior to implementing the change.
Page 11 of 13
RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (RA)
CONTAINMENT STRUCTURE EQUIPMENT HATCH SHIELD DOORS
- 17.
The control room (CR) radiological analysis supporting this license amendment request is based on a FHA inside containment with the containment open to the outside environment. All airborne radioactivity reaching the containment atmosphere is assumed to be exhausted within 2-hours to the outside environment via the open containment equipment hatch shield doors. This analysis uses a CR atmospheric dispersion factor (X/Q value) of 3.1 E-3 see/rn3, as described in Section 2.3A.2 of the UFSAR.
UFSAR Section 2.3.4.2 states that the CR X/Q value of 3.1 E-3 seclm3 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); 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. As such, please justify the use of the Murphy & Campe diffuse source-point receptor algorithm in this analysis.
Response
As indicated In the response to RAI #1, the only change to the FHA-IC dose AOR was an increase in the control room unfiltered inleakage rate from 10 to 1000 cfm total. The current licensing basis applies the atmospheric dispersion factors (X/Qs) for the release from the containment to the control room HVAC intake for all potential release points. Consistent with the current licensing basis, these same X/Qs are used to define dispersion between the containment equipment hatch and the control room HVAC intake. The containment equipment hatches are separated from the control room HVAC intakes by the containment building. As such, the containment equipment hatch to control room X/Qs will provide more diffusion than the modeled containment to control room X/Qs.
- 18.
If the Murphy & Campe diffuse source-point receptor algorithm is to be used in this analysis, UFSAR Section 2.3.4.2 states that a value of 180 ft (54.9 m) was assumed for the distance s between the containment surface and receptor location (i.e., CR air intake). UFSAR Figure 6.4-3 shows the location of the two emergency CR air intakes with respect to the Unit 2 and Unit 3 containment structures and seems to indicate that the distance s between the closest containment surface and each air intake is more like 90 ft (27.4 m) rather than 180 ft (54.9 m). Please justify the continued use of 180 ft (54.9 m) for the value of s in this analysis.
Page 12 of 13
RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (RA)
CONTAINMENT STRUCTURE EQUIPMENT HATCH SHIELD DOORS
Response
As indicated in the response to RAI #1, the only change to the FHA-IC dose AOR was an increase in the control room unfiltered inleakage rate from 10 to 1000 cfm total. The current licensing basis atmospheric dispersion factors (X/Qs) that are used in the dose calculation are based on a distance of 180 feet between the containment surface and the midpoint of the two control room emergency HVAC intakes. Since the dose calculation assumes two trains of CREACUS in operation during the first 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> of the event, modeling the average distance is a reasonable approximation of atmospheric dispersion for determining the activity concentration introduced into the control room. As shown in the following table, the modeled separation distance of 180 feet is also conservative with respect to the true separation distance of 186 feet.
SEPARATION SOURCE RECEPTOR DISTANCE (feet)
Unit 2 Containment A206 CR Emergency Intake 118.12 feet Unit 2 Containment A207 CR Emergency Intake 254.00 feet Unit 2 Containment Midpoint of A206 and A207 186.06 feet Unit 3 Containment A207 CR Emergency Intake 118.12 feet Unit 3 Containment A206 CR Emergency Intake 254.00 feet Unit 3 Containment Midpoint of A206 and A207 186.06 feet Page 13 of 13
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CALCULATION TITLE PAGE ICCN NOJ l PRELIM. CCN NO.
I PAGE OF_
CCN CONVERSION:
- alc. No. N-4072-003 DCP/FIDCN/FCN/ECP No. & Rev. Rev. 04 CCN NO. CCN-Subject Fuel Handling Accident FHA) Inside Containment - Control Room & Offsite Doses Sheet I of 241 System Number/Primary Station System Designators 1501 & 1510 / XBI & GKA SONGS Unit 2 and 3 0-Class II Tech. Spec.JLCS Affecting? 0 NO 0 YES, Section No. see below Equipment Tag No. NA Site Programs/Procedure Impact? 0 NO QYES AR No.
10CFR50.59 Review CONTROLLED COMPUTER PROGRAM/DATABASE Is this calculation Revision Being PROGRAMIDATABASE NAME(S)
VERSION/RELEASE NO.(S)
Issued solely to Incorporate CCNs?
H PROGRAM
_______see below H NO 0 YES O DATABASE AR No.
020301633 According to SO123-XXIV-5.A E ALSO, LISTED BELOW RECORDS OF ISSUES Y~~~~~~~~~~~~~~~~~~~~~
I I
I REV.
DISC.
DESCRIPTION TOTAL SHTS.
LAST SHT.
PREPARED (Print name/sign/date)
APPROVED (Signature/date) 0 See CDM Microfiche Records for ORIG.
FLS Revision 0 Signatures and Dates IRE Other I and 2 The Revisions I and 2 calculation ORIG.
FLS title page, with Its original Initials and signatures, is retained as IRE Other sheet 2 3
REVISED. Revision 3 Is a General 216 ORIG.
FLS Revision which does not contain Mark Drucker NFM revision bars. Rev. 3 Incorporates CCN-2 and CCN-3 215 IRE Brigitte Gossett Other
-~~
_m 4
To account for up to 1000 CFM unfiltered Control Room Inleakage.
241 ORIG.
- 7 AbId Rustaey NFM Rev. 4 also Incorporates CCNs 4,5 241 IRE and 6. This Is a complete revision.
Tom I I
Space for RPE Stamp, identify use of an alternate calc., and notes as applicable.
I Computer Codes used in the Revision 4 analysis are:
NE-319 'LOCADOSE Code', Version 3.0 (executed on RISC 6000 with AIX Version 4.2.1 Operating System)
NE-602 SOURCE2 Code. Version D2-5 (executed on RISC 6000 with AIX Version 4.2.1 Operating System)
NE-319 "LOCADOSE Code-, Release 6.0 (executed on NT workstation D088771. with Operating System Version 4.00)
As discussed in Section 2.3, the results of this calculation support the following Technical Specification Umiting Conditions for Operations (LCOs) and Licensee Controlled Specifications (LCSs):
Tech Spec LCO 3.3.8:
Containment Purge Isolation Signal (CPIS)
Tech Spec LCO 3.3.9:
Control Room Isolation Signal (CRIS)
Tech Spec LCO 3.7.11:
Control Room Emergency Air Cleanup System (CREACUS)
RECEIED COM Tech Spec LCO 3.9.3:
Containment Penetrations Tech Spec LCO 3.9.6:
Refueling Water Level LCS 3.3.100:
RPSIESFAS Response Times APR 1 2 2002 LCS 3.9.101:
Decay Time LCS 3.9.102:
Communications LCS 3.9.106:
Containment Purge Isolation Signal SITE FILE COPY This calc. was prepared for the identified DCPIFCN/ISCO ECP. DCPIFCNJECP completion and turnover acceptance to be verified by receipt of a memorandum directing DCN/ECN Conversion. Upon receipt, this calc. represents the as-built condition. Memo date _
by
k fe-v.
L/-
R.5
,r;v C
Z4.1 CALCULATION TITLE PAGE ICCN NOJ I PRELIM. CCN NO.
IPAGE-OF_
II CCII CO Caic. No. N4072-003 DCPIFIDCNIFCN No. & Rev. NIA N
Subject Fuel Handling Accident FHAI Inside Containment - Control Room & OfMsite Doses System Number/Primary Sttion System Desgnator 1501 &1510 l XB&
KA SONGS Unit 2and 3 Tech. SpecJLCS Affecting? O NO YES. Section No. see below Equipment Tag No. Utf Ste ProgramProcoue Impact? E NO OYES, AR No.
NVERSION:
X CCN.
Sheet-J-ofj
.Ciass 11 CONTROLLED COMPUT£R PROGRAMI DATABASE 1 PROGRAM 0 DATABASE PROGRAMIDATABASE NAME(S)
VERSIONIRELEASE NO.(S)
E ALSO. USTEO BELOW St* below
_ee below
.1 RECORDS OF ISSUES
- REV,
~~~~~~~TOTALPRPRDAROE D.
DESCRIPTION ST PREPAREDAR DISC.
C~~~~~~AST SHT.
(Slax I
8W CDM lMcroache Records for ORIG.
FRS Other Rev_
lsn 0 Sgniatrs and Dae IRE Other Oter 1 and The Revisions and I calcuation ORIG Other t__
e page, with ls erigial Initials and signars, Is retained as RE Other Other sheet 3
REYISED. Revision 3 s a General 215 Other RevIsion which does not contain Mar_____
NFM mevslabar. FPt.3nopotsl 21
_ouW Other CC.CCNC N
8,/
4 ORIG FLS Othur IRE Ober
~~~~~~~~~Or
- ~
~
~__
Space for RPE Stamp, identify use of an alternate calc., and notes as applicable.
Computer Codes used In the Revision 3 analysis (all are executed on RiSC 6000 with AIX Version 4.2.1 Operating system) are NE-319 LOCADOSE Code, Version 3.0 NE-602 SOURCE2 Code. Version D2-6 As discussed in Section 23, the results of this calculation support the following Technical Specification Umiting Conditlons for Operations (LCOs) and Licensee Controlled Specifications (LCSsY Tech Spec LCO 3.3.8:
ContaInment Purge Isolation Signal (CPIS)
Tech Spec LCO 3.3.:
Control Room Isolation Signal (CRIS)
Tech Spec LCO 3.7. 11:
Control Room Emergency Air Cleanup System (CREACUS)
Tech Spec LCO 3.9.3:
Containment Penetrations REEIED COM Tech Spec LCO 3.9.6 Refueling Water Level LCS 3 3.10 RPSIESFAS Response Times LCS 3.9.101:
Decay Time FEB 03 1999 LCS 3.9.102:
Communications LCS 3.9.106:
Containment Purge Isolation Signal
- IT E
Fll Yv Will. I RAhVVI I This cab. was prepared inr the idertifled OCPIFCN. OCPIFCN comptellon end turnover acceptance in be verified by receipt of a memorandum directing Thils cakc. was ppared forlthe I d6 DCPIFCN. DCP/FCN emptu ad rx acceptance ID be weried by receipt f a memorandum direcft DCN Conmesion. Upon eceipt, this calc. epresents he s-buil condion. Memo date by smt.w1 v.Aw" uitEFaOOCS01w
.lt Site File Copy
&v L-,
Rap~z 3 Pf29-I CALCULATION TITLE PAGE
-IccN Noi IPRELIM CON NO.
1PAGE O IjCN CONVEisION:
Calc. No.
N4072-003 DCP/MMPIFIDCN/FCN No. & Rev.
N/A lNo.CCN Subject Fuel Handlin Accident (RA) bside Containment - Control Room and Offsite Doses System Number/Primary Station System Designator XBI & GKA / 1501 & 1510 SONGS Unit 2 Sheet I
Q-CIass Tech Spec Affecting?
MNO
[YES, Section No.
Equipment Tag No.
NIA I.
V 1
CONTROLLED COMPUTER PROGRAM
. DATABASE r-1 PROGRAM EJ DATABASE Ul ACCORDOANCE WnTH NESII 414 PROGRAMIDATABASE NAME(S)
O ALSO LISTED BELOW uLOCADOSE" NE-319 3SOURCE2 NE602 VERSIONIRELEASE NO(S).
AM qer'GOns 40.
I Relo 3
I.I RECORD OF ISSUES REV.
TOTAL DESCRIPTIlON SHTS PREPARED APPROVED DISC STO (Print nameAftial)
(Signatrel 0
SEE MICROFICHE RECORDS NA ORIG Sor FOR REVISIONO
. N-
- N SIGNATURES AND DA#ES NA RtEDMAT 1
SE NOTE BELOW 89 CR \\0r ISSUEND FP USE R_____A___6_____
BPC N
.8 E'D>
A T. Remeick:
all 3
/
2 WSElDEb (Ph'A C
OA=
.i "
WM
¢S#{
/laA~a
)-b V___
4_.D TE Space f*w RPE samp, Idemify use ef an alternate caic., and note ae applicable.
Note 1:
The purpose of Revision is as follows:
1 To esolve OIRs92-019, 92-070 and 92-071 which state that data given in Revision 0 of this calculation is inconsistent with design data.
Ws analysis will cpede Revision 0.
2 Pages added - currently numbered pages I through 89 Pages revised - none Pages deleted - Revision 0 in its entirety
$BC t.tci 2 PK4E COANUE5 Are':
V 4 5ct. AcAA 6A bA, LfAo, Z7A,230 A S A, r
- 4V.z b
42'b.
PaCL~s EesuA : 5, (O 8.0, 14, I5, Tb Z, Z-1 'Bo, 3 I 33v6 36, Ss 3'7, 369, 42, 50,..
Pay s't, tcL:
1cOA I p5f z,3 ro4 an 51 52.
This cae. was prepared for the identified DCPIMMP.
DCP completion and turnover acceptance to be verified by recept of a memorandum directing DCN conversion. Upon receipt, this coa., represents the as-built condition. Memo date by
CALCULATION CROSS-INDEX I ICCN NOJ IPRELIM. CCN NO.
I PAGE_OF_
CCN CONVERSION:
CCN NO. CCN-Calculation No.
N-4072-003.
Rev. 4 Sheet 4
of 241 These interfacing calculations and/or OUTPUTS Does the Idntfyouuint Calc. rev.
documents provide input to the subject Results and conclusions of the subject output interface dlc coueunnt number dcluainanifrvsdmyeqrecalculation are used in these interfacing calc/documentca/dum t responsible caluition, nd if revised ay uin calculations and/or documents require revision?
ECP, CCN, DCN Fi initia TCN/Rev., FIDCN, or Cac/Document No.
Rev, No.
Calc/Document No.
Rev. No.
tracking number.
Rev. 4
_1 A
Units 2&3 Calculation C-257-01.06.01 2
& CCN 3 UFSAR Sections 15.7 and 15.10.7.3.9 UFSAR Appendices ISB and 15.IOB 16 No Units 2&3 Calculation M-0073-041 8
Accident Analysis DBD -
4 No
& CCN 20 DBD-SO23-TR-AA, Section 4.3.21 Units 2&3 Calculation M-0073-095 3
Radiation Monitoring DBD-4 No
& CCNs I to 4 DBD-SO23-690 Units 2&3 Calculation N-4010-001 5
Plant Protection System DBD -
6 No
& CCNs 2 & 3 DBD-S023-710 Units 2&3 Calculation N-4010-002 I
Units 2&3 Calculation A-92-NF-003 0
No
& CCNs I to 3 Units 2&3 Calculation N-0450-002 I
Units 2&3 Calculation N-0720-014 0
No Units 2&3 Calculation N-4060-008 5
Units 2&3 Calculation J-SPA-139 0
No
& CCNs 3 & 4 CCNs I to 3 Units 2&3 Calculation N-4060-016 2
Units 2&3 Calculation -SPA-179 0
No CCNs 2 to 4 Units 2&3 Calculation M-0022-008 2
Units 2&3 Calculation J-SPA-219 0
No
& CCNs I to 4 CCNs 3 4 Units 2&3 Calculation -SPA-179 0
Units 2&3 Calculation J-SPA-289 0
No
& CCNs 2 to 4 CCNs I to 3 Unit 2 Calculation NFM-2-FP-0002 0
Units 2&3 Calculation J-SPA-329 0
No I
c4
.W7 RFV 4 71 IRFFFRFNCE: S0123-XXIV.7.tS
CALCULATION CROSS-INDEX ICCN NO.I IPRELIM. CCN NO.
I PAGE f_
F_
I CCN CONVERSION:
CCN NO. CCN-Calculation No.
N-4072-003.
Rev. 4 Sheet 5
of 241 I
INPUTS OTUSDe h
Calc. rev These interfacing calculations and/or Results and conclusions of the subject output interface Identify output interface Ca~~c. rev, documnents provide input to the subject calculatio ad cousion of the sutbjct c ou untc caic/docunment number and calculation, and if evised my require calculations and/or documents.
ca docurnentE responsible revision of the subject calculation.
cacltosado ouet.rqiervsoECP.
CCN. DCN FLS initials TCN/Rev., FIDCN. or and date Calc/Document No.
ocument No.
Rev No YES/NO tracking number.
_No-alc(Docut No.
Rv. o.
YS I
Rev. 4
.m Unit 2 Calculation NFM-2-PH-0003 0
Unit 2 Technical Specifications LCO 3.3.8 LCO 3.3.9 LCO 3.7.11 LCO 3.9.3 LCO 3.9.6 132 132 181 127 134 No No No No No I
Unit 3 Calculation NFM-3-PH-0003 0
Unit 3 Technical Specifications LCO 3.3.8 121 No LCO 3.3.9 121 No LCO 3.7.11 172 No LCO 3.9.3 116 No LCO 3.9.6 123 No Unit 2 Calculation NFM-2-PH-0004 0
Unit 2 Licensee Controlled Specifications LCS 3.3.100 6 (1/00)
No LCS3.9.101 1 (5/00)
No LCS 3.9.102 2 (5/00)
No LCS 3.9.106 1 (5/00)
No Unit 3 Calculation NFM-3-PH-0004 2
Unit 3 Licensee Controlled Specifications LCS 3.3.100 6 (1/00)
No LCS 3.9.101 1 (5/00)
No LCS 3.9.102 2 (5100)
No LCS 3.9.106 1 (5/00)
No Unit 2 Calculation NFM-2-PH-0005 0
Unit 2 Calculation NFM-2-PH-0016 I
Unit 3 Calculation NFM-3-PH-0016 0
Units 2&3 Docurnent 2
S023-990-C299 I
I I
I I
WCt 7.A PFV 4 71 IRnF!tENCE: SMA3.XxIWv7.qS
CALCULATION CROSS-INDEX ICCN NOJ PRELIM. CCN NO.
PAGE _Of_
N-4072-003.
Rev. 4 Sheet 6 of 241 I
CCN CONVERSION:
CCN NO. CCN-Calculation No.
INPUTS OUTPUTS Does the Cale. ret.
These interfacing calculations and/or Results and conclusions of the subject output interface Identify output interface nunnber and documents provide input to the subject calculation are used n ttn d/
interfacing calc/docunent calcdocument numbersiand calcultion. and if reised may require calculations and/or documents.
require revision?ECCNDN resoniile revision of the subject calculation.
EP.N C
FLSntials TCN/Rcv., FIDCN, or Cale/Document No.
Rev. No.
CaIc/Document No.
Rev. No.
YES/NO tracking number.
p Rev. 4 4'd4L7i Ks3<
Units 2 &3 Calculation NFM-23-PH-116 0
Units 2&3 Drawing 40173A 20 Unit 3 Drawing 40173AS03 5
Units 2&3 Drawing 40173C 21 Units 2&3 Drawing 23060 6
Units 2&3 Drawing 40096 17 Units 2&3 Drawing 40098 9
Units 2&3 Drawing 31394 19 Unit 3 Technical Specifications to Amnend. 13 Unit 2 Licensee Controlled to 8/13/2001 Specifications (LCS)
Unit 3 Licensee Controlled to 8113/2001 Specifications (LCS)
Units 2&3 UFSAR 16 Procedure S023-1-2.44 6
(& TCN 6-4) 1 RrF 7A474 REV. 4 10I
IREFERENCE:
S0123-XXIV7.15
CALCULATION CROSS-INDEX ICCN NOJ PRELIM. CCN NO.
I PAGE _OF I
CCN CONVERSION:
CCN NO. CCN-Calculation No.
N-4072-003.
Rev. 4 Sheet 7
of 241 INPUTS OUTPUTS Does the C21C. rev.
TIese interfacing calculations and/or Results and conclusions of the subject output interface Identify output interface number and documents provide input to the subject calculation arm used in these interfacing calc/document cac/document responsible calculation. and if revised may require calculations and/or documents.
require revision?
T CN DCN FLS initials revision of the subject calculation.
CCN C I
~~~~~~~~~TCNIRev.,
FIDCN, or and date Calc/Document No.
l Rev. No.
Calc/Document No.
Rev. No.
YES/NO tracking number.
Rev._4cocum3-N.
1 Rev. 4 A0l S023-1l3-206._
I 4 I
ac 1R-4 Rr 4 t01
IREFERENCE:
S0123-XXIV-? 151
E&TS DEPARTMENT ICCN NOJ CALCULATION SHEET PRELIM. CCN NO.
l PAGE oject or DCPIFCNIECP Caic. No. N4072-003 CCN CONVERSION:
CCN NO. CCN --
Subject Fuel Handling Accident (FHA) Inside Containment - Control Room & Offsite Doses Sheet OF_
R A nf A1
.2 -
REV ORIGINATOR DATE IRE DATE REV ORIGINATOR DATE IRE DATE R l 4 A. Rust1e T. Remick 1
i TABLE OF CONTENTS Section DescriDtion Page CALCULATION COVER SHEET 1
REVISIONS 1, 2 CALCULATION COVER SHEET
................. 2 CALCULATION CROSS INDEX.........................................
3 TABLE OF CONTENTS.........................................
8 1.0 PURPOSE......................................... 10 1.1 Task Description.
10 1.2 Criteria, Codes and Standards......................................
1 1
2.0 RESULTS/CONCLUSIONS AND RECOMMENDATIONS
.12 2.1 Results/Conclusions.12 2.2 Comparison with Previous Dose Results.14 2.3 Relationship between this Calc, the Tech Specs, and the LCS's..
I5 2.4 Recommendations.17 3.0 MODELING ASSUMPTIONS..
18 4.0 DESIGN INPUTS................
26 5.0 METHODOLOGY.
48 5.1 General Methodology.
48 5.2 Computer Code Descriptions..
49 5.3 EAB, LPZ and Inside Control Room Cloud Doses......
................ 52 5.4 Gamma Radiation Shine Contributions to the Whole Body Dose.....
...... 56
6.0 REFERENCES
61 7.0 NOMENCLATURE....................................................
69 8.0 COMPUTATIONS 72 8.1 Fuel Rod Gap Inventories.............
72 8.2 Refueling Water Iodine Removal.........
79 8.3 Initial Containment Airborne Activity Profile..........................
1 8.4 Modeling of Control Room Isolation.
88
E&TS DEPARTMENT CALCULATION SHEET ICCN NOJ PRELIM. CCN NO.
PAGE _
OF _
roject or DCP/FCN/ECP Caic. No. N-4072-003 CCN CONVERSION:
CCN NO. CCN -
Subject Fuel Handling Accident (FHA) Inside Containment - Control Room & Offsite Doses Sheet 9 of 241 REV ORIGINATOR L DATE IRE DATE l REV ORIGINATOR DATE l
IRE DATE R
~~~~~~~~~1
~~~~~~~~~~~~~~E 4 A.
Rustaey T.
Remick v
......mm...m...mI...mmI L{mmuj.mmm..L..mI ~ I 8.5 8.6 8.7 8.8 8.9 8.10 Modeling of Flow from Containment to CR HVAC Intake Filter.....
...... 91 LOCADOSE Code Time Steps.................
94 Containment Shine Dose..........................................
95 Environmental Cloud Shine Dose..............
..................... 98 Control Room HVAC Filter Shine Dose.........
.................... 102 Summation of Offsite and Control Room Doses.......
................ 107 9.0 COMPUTER PROGRAM INPUT AND OUTPUT FILES
............. 109 9.1 Files Created During This Analysis.................................
109 9.2 LOCADOSE Code Input and Output Files...........................
III 9.3 SOURCE2 Code Input and Output Files.............................
169 10.0 QUATTRO-PRO SPREADSHEET EVALUATIONS.........
............... 211 1 0. 1 QUATTRO-PRO Evaluation of Containment Shine......
.............. 211 10.2 QUATTRO-PRO Evaluation of Environmental Cloud Shine.....
........ 220 10.3 QUATTRO-PRO Evaluation of Control Room HVAC Filter Shine....
.... 229 11.0 COPIES OF MISCELLANEOUS REFERENCES.
238
E&TS DEPARTMENT CALCULATION SHEET IPRELIM.CCN N(
oject or DCPIFCNfECP Caic. No. N4072-003 cc
.RIh;ar4 roIPI H in4linn Aelrirlant MI4AN Ineirlm fnnhinmant
- r-nnfrni Dennm &t rlff~iatf Mlriz 0kheko 1
f lAi REV ORIGINATOR DATE IRE DATE REV ORIGINATOR DATE IRE DATE R
4 A. Rustaey T. Remick v
7
=
={~~~_
1.0 PURPOSE 1.1 Task Description This calculation determines the control room and offsite doses due to a design basis fuel handling accident (FHA) inside the containment building (FHA-IC).
The results of this calculation support the FHA-IC discussion provided in UFSAR Chapter 15, and in Design Basis Document Accident Analysis Topical Report Section 4.3.21. As discussed in Section 2.3, the results of this calculation also support the following Technical Specification Limiting Conditions for Operations (LCOs) and Licensee Controlled Specifications (LCSs):
Tech Spec LCO 3.3.8:
Tech Spec LCO 3.3.9:
Tech Spec LCO 3.7.1 1:
Tech Spec LCO 3.9.3:
Tech Spec LCO 3.9.6:
LCS 3.3.100:
LCS 3.9.101:
LCS 3.9.102:
LCS 3.9.106:
Containment Purge Isolation Signal (CPIS)
Control Room Isolation Signal (CRIS)
Control Room Emergency Air Cleanup System (CREACUS)
Containment Penetrations Refueling Water Level RPSIESFAS Response Times Decay Time Communications Containment Purge Isolation Signal I
I I
I I
I I
I 1.1.1 Revision 4 Changes I
A license amendment has been proposed that would allow for the containment equipment hatch to remain open during fuel handling operations. This condition can be adequately addressed in the current model, which assumes that all of the FHA activity is released from containment within two hours. As such, the EAB, LPZ and control room doses should not change with the approval of this license amendment. However, the NRC has questioned the unfiltered inleakage assumption made in submittals made by other licensees. To address this issue, in the absence of definitive inleakage test results, Revision 4 of this calculation updates the FHA-IC calculation model to assume the conservatively high control room unfiltered inleakage rate of 1000 cfn.
The EAB and LPZ doses are not impacted by this modeling change. Revision 4 of this calculation still presents the results for the 10 cfin inleakage rate which remains the current analysis of record.
In addition to the above changes, Revision 4 of this calculation incorporates CCNs 4, 5 and 6.
Revision 4 of this calculation also uses LOCADOSE Release 6.0.
E&TS DEPARTMENT ILUUN No./
CALCULATION SHEET IPRELIM. CCN N1 roject or DCP/FCN/ECP Calc. No. N4072-003 CC Subject Fuel Handling Accident (FHA) Inside Containment - Control Room & Offsite Doses Sheet 11 of 241 REV ORIGINATOR DATE IRE DATE IREV ORIGINATOR DATE IRE DATE R
_A
-teT.
Remick 1 1 L
i 1.2 Criteria, Codes and Standards 1.2.1 Control Room Operator Dose Criteria The Control Room Operator dose criteria are defined in the General Design Criteria of Appendix A to 10 CFR 50 (Reference 6.4a). General Design Criterion 19 states that the control room personnel must be able to occupy the control room during accident conditions without receiving radiation exposures in excess of 5 Rem whole body, or its equivalent to any part of the body, for the duration of the accident. Standard Review Plan 6.4 (Reference 6.41,Section II) clarifies the dose guidelines by indicating that the dose to the control room personnel during the entire period of the postulated accident should not exceed:
Control Room (accident duration dose): Whole body gamma dose of 5 Rem Thyroid dose of 30 Rem Beta skin dose of 30 Rem 1.2.2 Offsite Dose Criteria Offsite dose criteria are defined in 10 CFR 100 (Reference 6.4b). Per Standard Review Plan 15.7.4 Revision 0 (Reference 6.4m), the dose to an individual should be well within the 10 CFR 100 exposure guidelines. This criterion is clarified in SRP 15.7.4 Revision 1, which states that the plant site and the dose mitigating engineered safety features are acceptable with respect to the radiological consequences of a postulated FHA if the calculated whole-body and thyroid doses at the Exclusion Area Boundary (EAB) and at the Low Population Zone (LPZ) are well within the 10 CFR 100, Section 100.11 exposure guidelines. SRP 15.7.4 Revision I defines "well within" as meaning 25 percent or less of the 10 CFR 100 exposure guidelines.
Per the exposure guidelines of 10 CFR 100, Section 100.1 1(a)(1), the dose at the EAB is for the two hours immediately following onset of the postulated accident. Per SRP 15.7.4 Rev. I, the EAB dose for this initial 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> period should not exceed:
EAB (2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> dose):
Whole body gamma dose of 6 rem Thyroid dose of 75 rem Per the exposure guidelines of 10 CFR 100, Section 100.11 (a)(2), the dose at the outer boundary of the LPZ is for the entire period of the postulated accident. Per SRP 15.7.4 Rev. 1, the LPZ dose during the entire period of the postulated accident should not exceed:
LPZ (accident duration dose):
Whole body gamma dose of 6 rem Thyroid dose of 75 rem
E&TS DEPARTMENT CALCULATION SHEET ICCN NOJ PRELIM. CCN NO.
PAGE _
OF _
)ject or DCP(FCN/ECP Caic. No. N4072-03 l CCN CONVERSION:
I CCN NO. CCN -
Subject Fuel Handling Accident (FHA) Inside Containment - Control Room & Offsite Doses Sheet 12 of 241 REV ORIGINATOR DATE IRE DATE REV ORIGINATOR DATE IRE DATE R
4 A. Rustaey T. Remick v
2.0 RESULTS/CONCLUSIONS AND RECOMMENDATIONS The doses presented in this summary of results section are rounded off to the nearest one-tenth rem, and doses smaller than 0.1 rem are listed as "< 0.1 rem". This round-off presentation style is in recognition of the inherent uncertainty in this type of dose analysis, and the fact that the dose criteria is given in multiple rem.
Section 2.1 present the results/conclusions of this calculation. Section 2.2 presents a comparison of the current calculation results with previously calculated doses. Section 2.3 presents the relationships between this calculation, the Technical Specifications, and the Licensee Controlled Specifications. In addition, Section 2.4 presents recommendations for the review of various documents that may need to be revised to reflect the methodology and/or results of this calculation.
2.1 Results/Conclusions Based on the calculations presented in Section 8, Table 2.1-1 presents the Control Room, Exclusion Area Boundary (EAB), and Low Population Zone (LPZ) doses following a design basis fuel handling accident inside containment (FHA-IC). The design basis case is the failure of 226 fuel rods, representing 16 fuel rods in the dropped fuel bundle, and 210 fuel rods in one or two impacted fuel bundles. A review of Table 2.1-1 shows that the FHA-IC offsite and control room doses meet the dose criteria listed in Section 1.2.
Table 2. 1 -1 summarizes the dose consequences of a FHA-IC involving high burnup fuel.
Section 8.1 has determined that the FHA-IC involving high burnup fuel yields dose consequences that bound those of an EHA-IC involving once burned fuel. Per Design Input 4. 1, once burned fuel is fuel that has been irradiated for a single fuel cycle, and high burnup fuel is fuel that has been irradiated for more than one fuel cycle, with a burnup as high as 60 GWD/MTU.
The 16 high burnup fuel rods in the dropped fuel bundle are modeled with an iodine fuel rod gap fraction of 12 percent and a radial peaking factor (RPF) of 1.75. The 210 high bumup fuel rods in the impacted fuel bundles are modeled with an iodine fuel rod gap fraction of 12 percent and an average RPF of 1.40.
The design basis case failure of 226 fuel rods is sufficient to induce a high radiation Control Room Isolation Signal (CRIS) which within 3 minutes places the Control Room HVAC system into the high radiation isolation mode and starts both trains of the control room essential HVAC system. In the event of a CRIS failure, the control room doses reported in Table 2. 1-1 remain valid should manual Operator Action be taken to isolate the control room within this same 0
W
E&TS DEPARTMENT CALCULATION SHEET ICCN NOJ PRELIM. CCN NO.
PAGE _
OF _
ject or DCPIFCNIECP Caic. No. N4072-003 CCN CONVERSION:
CCN NO. CCN-Subiect FuI Handilinn ArccirdAnt (FHAI Insii Containment - Control Room. nffnitp nse Sheet 1 of 241 J
o
-la wlFXf Snfi o_
__j ww
~
__l
-i -x -rv I I -t Ax _@
a_
I REV ORIGINATOR DATE IRE DATE REV ORIGINATOR DATE IRE DATE R
4 A. Rustae T. Remick E
=
=
ffi 3 minute period. Operator Action within 3 minutes is reasonable due to the Units 2&3 Licensee Controlled Specification 3.9.102 (References 6.4e & 6.4f) requirement for direct communications between the Control Room and the Senior Reactor Operator supervising core alterations, and due to the Abnormal Operating Instruction S023-13-20 (Reference 6.5c) requirement for immediate manual CRIS initiation if an FHA with high radiation occurs. As such, the Control Room Operators can be made immediately aware of a dropped fuel assembly, and the potential need to isolate the control room.
This calculation assumes that the containment purge is not isolated by a containment purge isolation signal (CPIS) following a fuel handling accident inside containment. Because no credit is taken for CPIS actuation, this calculation does not model any CPIS related instrument delays or uncertainties, nor does this calculation model a containment purge isolation valve closure time.
I Table 2.1-1 Fuel Handling Accident Inside Containment Doses Location Criteria (Rem)
Dose (Rem)
Control Room (event duration dose) 1l CFM I000 CFM
[CR isolated at 3 minutes]
CR Unfiltered CR Unfiltered In-leakage In-leakage Thyroid Inhalation 30 14.8 25.4 Beta Skin Immersion 30 3.6 3.6 Whole Body Gamma Immersion 5
0.2 0.3 EAB (2-hour dose)
- Thyroid Inhalation 75 56.4 56.4 Beta Skin Immersion no dose criterion 0.3 0.3 Whole Body Gamma Immersion 6
0.3 0.3 LPZ (event duration dose)
- Thyroid Inhalation 75 1.6 1.6 Beta Skin Immersion no dose criterion
< 0.1
< 0.1 Whole Body Gamma Immersion 6
<0.1
< 0.1 I
I I
I I
I I
I I
I I
I
- EAB and LPZ were not re-evaluated for increased CR In-leakage
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2.2 Comparison with Previous Dose Results Table 2.2-1 presents a comparison of the doses calculated in this revision with the doses calculated in Revision 3. As shown in Table 2.1-2, Except for Control Room thyroid inhalation dose, the current doses are all either less severe than, or equivalent to, the previously calculated doses. The Controlroom thyroid doses are higher since the Rev. 4 analysis assumed 1 000 CFM of unfiltered control room inleakage which is significantly higher than O CFM assumed in Revision 3.
I I
I I
I Section 1. 1. Identifies changes made in the Revision 4 of this calculation.
I I
I The following represent the primary reasons for the change in dose exposure:
Previous revisions of this calculation modeled an unfiltered control room inleagkage flow of O CFM. Revision 4 models 1000 CFM of unfiltered inleakage flow to the control room.
I I
Table 2.2-1 Comparison of Current and Previously Calculated FHA Inside Containment Doses Location i
Revision 3 Revision 4 Calculated Doses (Rem),
CurrentDoses (Rem)
Control Room (event duration dose) 10 CFM 1000 CFM 10 CFM 1000 CFM
[CR isolated at 3 minutes]
Inleakage Inleakage Inleakage Inleakage Thyroid Inhalation 14.8 N/A 14.8 25.4 Beta Skin Immersion 3.6 N/A 3.6 3.6 Whole Body Gamma Immersion 0.2 N/A 0.2 0.3 EAB (2-hour dose)
(not re-analyzed)
Thyroid Inhalation 56.4 56.4 Beta Skin Immersion 0.3 0.3 Whole Body Gamma Immersion 0.3 0.3 LPZ (event duration dose)
(not re-analyzed)
Thyroid Inhalation 1.6 1.6 Beta Skin Immersion
< 0.1
< 0.1 Whole Body Gamma Immersion
< 0.1
< 0.1
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2.3 Relationship between this Calc, the Tech Specs, and the LCS's The results of this calculation support several Technical Specification Limiting Conditions for Operations (LCOs) and Licensee Controlled Specifications (LCSs). The following items list these LCOs and LCSs, and address the relationship between these licensing commitments and this calculation:
(1) Tech Spec LCO 3.3.8 addresses the need for at least one operable Containment Purge Isolation Signal (CPIS) channel during core alterations and during the movement of irradiated fuel assemblies within containment. The Tech Spec Bases for this LCO note that containment purge is not isolated following the fuel handling accident inside containment.
This calculation is consistent with the Tech Spec Bases statements.
(2) Tech Spec LCO 3.3.9 addresses the need for at least one operable Control Room Isolation Signal (CRIS) channel during Mode 6 and the movement of irradiated fuel assemblies.
Surveillance Requirement 3.3.9.6 requires verification that the CRIS response time is within limits. The Tech Spec Bases for this LCO note that CRIS, in conjunction with the Control Room Emergency Air Cleanup System (CREACUS) (see Tech Spec LCO 3.7.1 1), maintains the control room atmosphere within conditions suitable for prolonged occupancy throughout the duration of any one of the accidents (e.g., FHA-IC) discussed in UFSAR Chapter 15.
The Tech Spec Bases for Surveillance Requirement 3.3.9.6 note that the response time testing acceptance criteria are included in LCS 3.3.100. This calculation is consistent with the Tech Spec Bases statements.
(3) Tech Spec LCO 3.7.1 1 addresses the need for two operable CREACUS trains during Mode 6 and the movement of irradiated fuel assemblies. The Tech Spec Bases for this LCO note that in the CREACUS Emergency Mode the control room is isolated to protect operational personnel from radioactive exposure through the duration of any one of the postulated limiting faults (e.g., FHA-IC) discussed in UFSAR Chapter 15. This calculation is consistent with the Tech Spec Bases statements.
(4) Tech Spec LCO 3.9.3 addresses the containment penetration status during core alterations and during the movement of irradiated fuel assemblies inside containment. The Tech Spec Bases for this LCO note that 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. This LCO ensures that the containment personnel airlock can be closed after containment evacuation in the event of a fuel handling accident. This calculation is consistent with the Tech Spec Bases statements.
(5) Tech Spec LCO 3.9.6 addresses the refueling water level to be maintained above the reactor vessel flange during the movement of fuel assemblies or control element assemblies within
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the reactor pressure vessel when either the fuel assemblies being moved or the fuel assemblies seated within the reactor pressure vessel are irradiated, and during the movement of irradiated fuel assemblies within containment. The Tech Spec Bases for this LCO note that a minimum water level of 23 feet above the top of the reactor vessel flange is required to ensure that the iodine fission product activity is retained in the water in the event of a fuel handling accident. This calculation is consistent with the Tech Spec Bases statements.
(6) LCS 3.3.100 addresses the Reactor Protection System (RPS) and Engineered Safety Features Actuation System (ESFAS) response times. Table 3.3.100-2 gives a CRIS response time of 120 seconds for placing the control room ventilation in emergency mode due to high control room ventilation airborne radiation. This calculation is consistent with the LCS 3.3.100 Bases in that it explicitly models a 3 minute high radiation induced CRIS response time which is greater than the 2 minute response time required by the LCS.
(7) LCS 3.9.101 addresses the need for the reactor to 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 /> during the movement of irradiated fuel in the reactor pressure vessel. The Bases for this LCS state that this 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> decay time ensures that sufficient time has elapsed to allow the radioactive decay of the short lived fission products, and that this decay time is consistent with the assumptions used in the accident analyses. This calculation is consistent with the Tech Spec Bases statements.
(8) LCS 3.9.102 addresses the need to maintain direct communications between the Control Room and the Senior Reactor Operator supervising core alterations. The Bases for this LCS do not provide a relationship between this requirement and the fuel handling accident inside containment scenario. However, as noted in Section 2.1.1, in the event of a CRIS failure the control room doses reported in Table 2.1-1 will remain valid should manual Operator Action be taken to isolate the control room within a 3 minute period, and that Operator Action within 3 minutes is reasonable due to the LCS 3.9.102 communications requirement, and due to the Abnormal Operating Instruction S023-13-20 (Reference 6.5c) requirement for manual CRIS initiation (9) LCS 3.9.106 addresses the need for the Containment Purge Isolation System to remain operable during Core Alterations and during the movement of irradiated fuel assemblies within containment. The Bases for this LCS note that the operability of this system is required to restrict the release of radioactive material from the containment atmosphere to the environment. However, as noted in Section 2.1.1, this analysis assumes that the containment purge is not isolated following a fuel handling accident.
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Subject Fuel Handling Accident WFHA) Inside Containment - Control Room & Offsite Doses Sheet 17 of 241 a
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2.4 Recommendations 2.4.1 The following documents may need to be revised to reflect the methodology and/or results of this calculation. Action Requests (as listed in the Calculation Cross-Index) have been initiated to track these recommendations.
(1) None 2.4.2 It is also recommended that future calculations should double the CR gamma immersion dose to account for direct containment shine, environmental cloud shine, and CR intake/recirc filter shine.
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I
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= I t 3.0 MODELING ASSUMPTIONS 3.1a Configuration Symmetry Between Units 2 and 3 This analysis and its conclusions are applicable to both Units 2 and 3. References are provided to show that assumptions and design input data that have unit applicability are representative of both units.
3.1b Applicability of Cycle 10 Data to Future Cycles This calculation in parts uses Cycle 10 data which are mostly generic in nature. The applicability of this data to any future cycle shall be verified.
3.2 Reactor Shutdown (Decay) Time Regulatory Guide 1.25 Section C.la (Reference 6.4i) states that the fuel handling accident should be assumed to occur at a time after shutdown identified by the Technical Specifications as the earliest time fuel handling operations may begin, and that radioactive decay of the fission product inventory during the interval between shutdown and commencement of fuel handling operations may be taken into consideration. Units 2&3 Licensee Controlled Specification 3.9.101 (References 6.4e & 6.4f) requires that the reactor shall 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 /> prior to movement of irradiated fuel in the reactor vessel. Therefore, it is assumed that 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of reactor shutdown time will have elapsed prior to the occurrence of the fuel handling accident. This 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> delay, which is reflected in the Design Input 4.1 source term, is sufficient time to allow the radioactive decay of short lived fission products that would have otherwise contributed to the fuel handling accident dose consequences, and to allow a reduction in the fuel rod pressures thereby validating the modeled spent fuel pool iodine decontamination factors (refer to Section 8.2.2).
3.3 Fuel Rod Gap Iodine Inventory Species Composition Per Section C. If of Regulatory Guide 1.25, the iodine gap inventory is composed of 99.75 percent inorganic species (i.e., elemental and particulate iodine), and 0.25 percent organic species (i.e., organic iodide). It is assumed that this same relative proportion of inorganic and organic iodine species exists in both once burned and high burnup fuel.
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3.4 Release Duration In accordance with Regulatory Guide 1.25 Section C.li (Reference 6.4i), all of the radioactive material released into the Containment atmosphere is assumed to be released to the outside enviromnent over a two hour interval.
3.5 Containment Air Circulation Units 2&3 Technical Specification (References 6.4c & 6.4d) LCOs 3.6.6.1 and 3.6.6.2 require that the Containment Cooling Train Fans should be operable in MODES 1, 2, 3, and 4.
Units 2&3 Technical Specification LCO 3.6.8 requires that the Containment Dome Air Circulators should be operable in MODES I and 2. Since there is no Tech Spec requirement to have any air circulation units operable in MODE 6 (Refueling), it is assumed that the units will not be operating during the fuel handling accident.
Because of the assumed absence of air circulation in the containment, this analysis will not take credit for activity dilution within the containment dome air space. The modeling of a smaller containment air dilution volume will yield higher containment activity concentrations, which will yield higher and therefore conservative containment shine doses to offsite and control room locations.
Per Assumption 3.4, all of the radioactive material released into the Containment atmosphere is assumed to be released to the outside environment over a two hour interval. As such, the modeling of a smaller contaimnent air dilution volume will not change the activity release to the environment, and there will be no impact on the calculation of offsite and control room immersion doses.
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3.6 Containment Purge Isolation Units 2&3 Technical Specification (References 6.4c & 6.4d) LCO 3.3.8 requires that one Containment Purge Isolation Signal (CPIS) channel should be operable in MODES 1, 2, 3, and 4, during core alterations, and during the movement of irradiated fuel assemblies within containment. Nevertheless, this calculation assumes that the containment purge is not isolated by a CPIS following a fuel handling accident. This assumption is conservative in that it ensures a maximum release of containment airborne activity to the environment over the two hour accident duration (per Assumption 3.4).
Because no credit is taken for CPIS actuation, this calculation does not model any CPIS related instrument delays or uncertainties, nor does this calculation model a containment purge isolation valve closure time.
3.7 Activity Decay During Transport to Dose Receptors The activity released from the containment is assumed to be instantaneously transported by atmospheric dispersion to the control room HVAC intake and to the offsite EAB and LPZ dose receptors. No credit is taken for radioactive decay of the isotopes in transit to these locations.
3.8 Control Room Emergency Filtration Damper Stroke Times The stroke time of the HVAC isolation dampers in the CR HVAC system is assumed to be 6 seconds, based on an E-Mail from F. Santa Ana to T. Remick (Reference 6.3c).
3.9 Unfiltered Control Room Inleakage The previous revisions of FHA-IC analysis assumed 10 cfm of control room unfiltered inleakage. The NRC Staff has let it be known in documents such as RIS 2001-19 that the inleakage should be validated in all future dose-related licensing submittals. In the absence of a formal validation test, this new FHA-IC evaluation assumes a conservatively large value of 1000 CFM of control room unfiltered inleakage. Both scenarios (10 and 1000 CFM) will be presented.
3.10 Control Room and Offsite Breathing Rates Regulatory Guide 1.4 Section C.2c (Reference 6.4h) addresses at work and at rest breathing rates of individuals present at offsite Exclusion Area Boundary (EAB) and Low Population Zone (LPZ) locations during a loss of coolant accident. These breathing rates, which are presented in
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Table 3.10-1, will be used in the determination of the thyroid inhalation doses to individuals at the EAB and LPZ during the fuel handling accident. The at-work breathing rate will also be used in the determination of the thyroid inhalation doses to Control Room Operators.
Table 3.10 Control Room and Offsite Breathing Rates Time Interval Control Room EAB LPZ Breathing Rate Breathing Rate Breathing Rate (m 3/sec)
(m 3/sec)
(m 3/sec) 0 to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 3.47e-04 3.47e-04 3.47e-04 2 to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 3.47e-04 Not Applicable 3.47e-04 8 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 3.47e-04 Not Applicable 1.75e-04 I to 30 days 3.47e-04 I
Not applicable 2.32e-04 3.11 Daughter Product Isotopes The effects of daughter products are considered in this analysis. The daughter products generated by the radioactive decay of Iodine are noble gas (Krypton, Xenon) isotopes. The daughter products generated by the radioactive decay of the noble gas isotopes are Rubidium and Cesium.
The Bechtel LOCADOSE computer program (NE-3 19, Reference 6.6a) classifies each isotope as a member of one of eleven isotope groups. The isotopes in each group share common characteristics (e.g., applicable filter efficiencies), which are entered into the LOCADOSE input file. Rubidium and Cesium are in LOCADOSE isotope group 5. This grouping is consistent with NUREG-001 7 (Reference 6.4n, Table 2-6), which places Rb and Cs in the same element class. Per NUREG-00l 7 Table 2-17 the Cs isotopes are particulates; it is therefore assumed that the Rb isotopes are also particulates. Therefore, LOCADOSE Group 5 is considered to consist of particulates which are modeled as being subject to High Efficiency Particulate Air (HEPA) filtration.
Although not associated with the FHA source term, it is noted that the isotopes in LOCADOSE groups 6 (Te, Sb), 7 (Sr, Ba), 8 (noble metals), and 9 (rare earths) are also listed as particulates in NUREG-0017 Table 2-17. Therefore, LOCADOSE Groups 6 to 9 are considered to consist of particulates which are modeled as being subject to HEPA filtration. In addition, LOCADOSE Group 10 (other isotopes, including H-3 and N-16) and Group 11 (halogens other than iodine) which are not particulates are not modeled as being subject to HEPA or charcoal filtration.
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3.12 Control Room Filter Efficiency Model for Activity Loading Per Design Input 4.4.2, in the high radiation isolation mode the makeup air entering the control room passes through filters SA15IMA206 and SAI 5OMA207 (A206 and A207) of the Emergency Ventilation System train, and then through filters SAI51OME418 and SA 1510ME419 (E418 and E419) of the Emergency Air Conditioner filter train. For the purpose of determining the shine dose contribution from the control room HVAC filters, control room HVAC intake filters A206 and A207 are assumed to be 100% efficient at removing iodine and particulates from the incoming air. This maximizes the amount of iodine and particulates retained on A206 and A207, and thus maximizes the shine dose from these two filters. In reality, iodine and particulates that are not trapped on the intake filters will eventually be trapped on recirculation filters E418 and E419, which are located in the vicinity of the intake filters per Drawing 40002 Sheet I (Reference 6.2m). However, the results of the filter shine dose calculation would not be significantly different, since the geometry of the direct shine pathways from E418 and E419 is similar to the geometry of the direct shine pathways from A206 and A207. Modeling the filters with a retention efficiency of 100 percent is consistent with UFSAR Appendix 15B Table I SB-5 footnote b.
3.13 Control Room Filter Volume Based on Vendor Drawing S023-41O-1-1 (Reference 6.2n) and an E-mail from D. Higgins to F. Santa Ana (Reference 6.3b), a control room HVAC intake charcoal filter is approximated by a rectangular solid 27%" long x 24" wide x 19" tall. The volume of a filter is:
Filter volume = 27.75 in x 24 in x 19 in x int3 = 7.32 ft 3 1728 in3 The 19" height approximates three stacked carbon trays, each with a height of 6 /32 -
3.14 Control Room Post-Accident HVAC Operation Consistent with UFSAR Appendix 15B (Reference 6.4g), this analysis models two trains of emergency HVAC in operation during the first 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> of the accident. Operator action is assumed to be taken within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> to deactivate one train of emergency intake and recirculation units.
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el kiftu#
M I1ma 4onflinn Arpir-Ant /IWA% IncI.A t-nntfinmelnt - r~ntrtnr Df~frsn Rk rffeita riniec
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-V REV ORIGINATOR DATE IRE DATE REV ORIGINATOR DATE IRE DATE R
4 A. Rustaey T. Remick 3.15 Control Room Filter Accumulation Time Per Assumption 3.4 all of the radioactive material is assumed to be released to the outside environment over a two hour interval. Per Assumption 3.14, both control room filter trains are assumed to operate during the first 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> of an accident (after 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> one filter train is secured by Operator action). Therefore, the activity released from the containment and dispersed to the control room HVAC intake is assumed to evenly accumulate onto each of the charcoal filters during the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> release duration. After 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, the activity on each charcoal filter is allowed to decay for the duration of the accident.
3.16 Control Room Filter Model Flows As discussed in Section 5.4.3, a LOCADOSE node was used to model the CR HVAC intake filter. This was essentially done by assuming that there was only inflow into this node. The LOCADOSE node recirculation filters were modeled to eliminate the noble gases that are not retained on the charcoal filters, by assuming a high recirculation filter flowrate (1e6 cfn) with respect to the filter volume (7.32 cubic feet per Assumption 3.13). The LOCADOSE node recirculation filters were modeled with 0% iodine removal efficiencies, a 100% noble gas removal efficiency, and 0% all other material removal efficiencies. These assumptions result in the recirculation filters removing the noble gases, while leaving the iodines and particulates in the charcoal filter. Of note is that the LOCADOSE node "recirculation filters" are not recirculation filters SA151OME418 and SAI5IOME419 (E418 and E419) of the Control Room Emergency Air Conditioner filter train.
3.17 Control Room Dose Point Locations Per Section 5.4, Dose Multiplication Factors (DMFs) are used to evaluate Containment Shine, Environmental Cloud Shine and Control Room Filter Shine gamma doses. Per Design Inputs 4.10 through 4.12, Calculations N-0450-002 (Reference 6.1 f) and N-4060-016 (Reference 6. lh) determined DMFs for the 16 control room dose point locations shown in Figure 3-1. To simplify this analysis, doses will only be calculated at the four dose points representing the actual contr.ol board area (dose points 9, 10, 15, and 16). The maximum dose at these four dose points will be used to represent the Control Room Operator dose.
Per Calculations N-0450-002 and N-4060-016, Body IA/2A in Figure 3-1 represents control room HVAC intake charcoal filter SA 151 OMA206, Body 1 B/2B represents control room HVAC intake charcoal filter SA151OMA207, and Bodies Fl through F20 represent the control room fire walls that provide some shielding of the gamma radiation shine from the filters.
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I jE 1'~- _.111lIIIIl Figure 3 Control Room Dose Point Locations
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u l nAv 4 inn A ii-ot 4
n ICUA I neil P nftinm~ntr
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'ffeita Mlaa cQkpmt 'Dr rf 91A REV ORIGINATOR DATE IRE DATE REV ORIGINATOR DATE IRE DATE R
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3.18 Refueling Pool Water Temperature Per Calculation M-0022-008 (Reference 6.1i, CCN-I page A-5), the maximum spent fuel pool water temperature is 160 0F for the case of a full core offload with consolidated spent fuel storage. For the purposes of this calculation, the maximum refueling pool water temperature is assumed to be 160 OF.
3.19 Hot Fuel Rod Characteristics During Power Operations Design Analysis NFM-2-FP-0002 (Reference 6.1k) presents the Unit 2 Cycle 10 Fuel Performance Analysis. Per Design Analysis NFM-2-FP-0002 Section 6.1 Table 6-1, the Urnit 2 Cycle 10 Batch M erbia hot rod FATES standard output file is s2clOmeh.out. Per this output file, the temperature difference between the outer surface of the fuel cladding and the coolant is less than 50 F for a linear heat rate of 8.62 kw/ft (which exceeds the core average linear heat rate of 5.311 kw/ft [Reference 6.1k, Table 2-2]).
In addition, per this output file the temperature in the fuel rod gap space during power operations is approximately 300'F greater than the bulk temperature of the surrounding cooling water. For conservatism, it is assumed that the temperature at the inner surface of the fuel cladding during power operations is no more than 5007F greater than the bulk temperature of the surrounding cooling water.
3.20 Radioactive Decay Heat at 72 Hours after Shutdown Per Assumption 3.2 refueling operations begin 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after shutdown. Per Branch Technical Position ASB 9-2 (Reference 6.4p, Figures I through 3), the heat generation rate associated with residual radioactive decay 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> (2.59e5 seconds) after shutdown is less than 0.5 percent of the heat generation rate during power operations. To address additional decay heat that may be present with the SONGS Units 2&3 high bumup fuel management, it is assumed that the heat generation rate associated with residual radioactive decay at 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> is less than one percent of the heat generation rate during power operations.
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t.i~.
r,.C1 Wlllim-A,,rd4en* 1=JA Icrl, tf.n*finmrnt
- r.tontrnl Dt^rtrn f
R rffcifP noccbc
'h A f
3AI REV ORIGINATOR DATE IRE DATE REV ORIGINATOR DATE IRE DATE R
4 A. Rustae T. Remick E
l
=
=a~~~~~~~~~~~~
4.0 DESIGN INPUTS Note: This calculation in parts uses Cycle 10 data which are mostly generic in nature. The applicability of this data for use in any future cycle is verified as part of reload process.
4.1 Fuel Handling Accident Source Term Note: The FHA source term has increased by 2% to the values shown in Reference 6.1v (Section 4.3 and its reference table 3.3-1 at time 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />). However, this analysis continues to use those values stated in Table 4.1-1. To account for this increase radial peaking factors of Design Input 4.1.4 will be adjusted accordingly.
Design Input 4.1 presents the activity profile and isotopic characteristics of an average fuel rod.
These characteristics are dependent upon fuel bumup. This calculation evaluates fuel handling accidents involving once burned fuel (i.e., fuel that has been irradiated for a single fuel cycle) and high bumup fuel (i.e., fuel that has been irradiated for more than one fuel cycle).
Per Unit 2 Cycle 10 Physics Input Design Analysis NFM-2-PH-0005 (Reference 6.1 r, Section 4.3c and Table 3.3.3.3-5), once burned Batch M fuel assemblies have a maximum burnup of less than 33 GWDJMTU, and high bumup Batches K and L fuel assemblies (i.e.,
assemblies that have been irradiated for more than one fuel cycle) have a maximum burnup of 60 GWD/MTU.
4.1.1 Fuel Rod Inventory. Per Assumption 3.2, the fuel handling accident is assumed to occur after 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of reactor shutdown time. Per Regulatory Guide 1.25 Section C.ld (Reference 6.4i), iodine and noble gas gap activity is released from the damaged fuel rods.
Table 4. 1-1 presents the iodine and noble gas activity inventory in a single fuel rod after 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of decay, as documented in Unit 2 Cycle 10 Source Term Design Analysis NFM-2-PH-0016 (Reference 6.1s, Section 4.0). Per the design analysis, this source term is valid for enrichments from 4 to 5 w/o Uranium-235, for fuel managements where the maximum burnup of any rod within 10% of the limiting radial peaking factor is less than 60 GWD/MTU, for an average power per fuel rod of less than or equal to 0.0666 MWth, and for a maximum number of 300 non-fuel rods in the core. Per Unit 3 Cycle 10 Source Term Design Analysis NFM-3-PH-0016 (Reference 6.1t, Section 4.0), the U3C1O source term is bounded by the U2C1O source term.
Design Analysis NFM-2-PH-0016 documents the presence of tritium in an average fuel rod, but not the presence of other particulate fission products that are also present in spent fuel.
Table 4.1-1 does not address these isotopes since Regulatory Guide 1.25 Section C.Id does not
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Subject Fuel Handling Accident (FHA) Inside Containment - Control Room & Offsite Doses Sheet 27 of 241 REV ORIGINATOR [
DATE IRE DATE REV ORIGINATOR DATE IRE DATE R
4 A. Rustsey T. Remick v
~~ 7 7
=
=
{~~~~
require consideration of non-iodine or non-noble gas gap activity releases from the damaged fuel rods.
i Table 4.1-1 Average Fuel Rod Inventory (See note below)
Isotope Average Fuel Rod Inventory after 72 Hours Decay
[per Design Analysis NFM-2-PH-0016, Section 4.0]
(curies/rod)
Iodine-129 1.18e-04 Iodine-130 3.09e-02 Iodine-131 1.40e+03 Iodine-132 8.53e-07 Iodine-133 3.25+02 Iodine-134 0.00e+00 Iodine-I 35 1.79e+00 Krypton-83m 3.68e-10 Krypton-85 2.58e+01 Krypton-85m 8.I5e-03 Krypton-87 1.02e-14 Krypton-88 4.21 e-05 Xenon-131m 1.91e+0I Xenon-33m 6.42e+0 1 Xenon-133 2.81 e+03 Xenon-135m 2.74e-0 Xenon-135 3.69e+01 Xenon-1 38
_.00e+00 I
Note: The FHA source term has increased by 2% to the values shown in Reference 6. v (Section l
4.3 and its reference table 3.3-1 at time 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />).
I I
4.1.2 Fuel Rod Gap Inventory. Per Regulatory Guide 1.25 Section C. I d (Reference 6.4i), all of I
the gap activity in the damaged fuel rods is released from the damaged fuel rods. This gap activity consists of 10 percent of the total noble gases other than Krypton-85, 30 percent of the Krypton-85, and 10 percent of the radioactive iodine in the rods at the time of the accident.
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[jJA. Rustaey T. Remick I
T
=~~~~~~~~~~~~~~~~~~~~~
Per NUREG/CR-5009 (Reference 6.4k, Section 3.2.2 and Table 3.6), operation with fuel discharge burnup extended to 60 GWD/MTU increases the fuel rod gap inventory of several isotopes. At high fuel discharge bumup, all but the iodine isotopes are calculated to have fuel rod gap inventories less than that recommended by Section C.ld of Regulatory Guide 1.25 for use in fuel handling accident analyses. Per NUREG/CR-5009, at high bumups the fuel rod gap Iodine-131 inventory increases to 12 percent, which is greater than the 10 percent value that is assumed in Regulatory Guide 1.25. This analysis will assume that the fuel rod gap inventories of all iodine isotopes can potentially increase to 12 percent.
4.1.3 Number of Failed Fuel Rods. ABB Calculation A-SCE-FMDE-002 (Reference 6.1u) evaluated the extent of any damage produced by a fuel bundle dropping either horizontally or vertically onto one or more other fuel bundles in the reactor vessel during fuel handling operations. Per ABB Calculation A-SCE-FMDE-002 Sections 3.0 and 5.2, the maximum number of fuel rods predicted to fail will occur as a result of the vertical drop of the fuel assembly onto the fuel bundles in a partially loaded core. In this scenario, a total of 226 fuel rods will fail, representing 16 fuel rods in the dropped fuel bundle, and 210 fuel rods in one or two impacted fuel bundles.
4.1.4 Average Radial Peaking Factor.
Per Design Input 4.1.3, in the event of a fuel handling accident inside containment a total of 226 fuel rods will fail, representing 16 fuel rods in the dropped fuel bundle, and 210 fuel rods in one or two impacted fuel bundles. Regulatory Guide 1.25 Section C.le assumes a radial peaking factor (RPF) of 1.65. However, this RPF may be non-conservative when a small number of fuel rod failures are considered, and overly conservative when a large number of fuel rod failures are considered. To address this concern, Section B of Regulatory Guide 1.25 allows alteration of assumptions due to site specific characteristics, plant design features, and major changes in fuel composition or management.
I
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Subject Fuel Handling Accident (FHAI Inside Containment - Control Room & Offsite Doses Sheet 29 of 241 REVI ORIGINATOR DATE IRE DATE IREVl ORIGINATOR DATE l IRE DATE R
-I I
I t
1IE=I 4 A Rutay j
T. Remick j
_Iv I
mmmmmin~~~~mL 1 2.mm]
.W
.U~~~
4.1.4.1 Average RPF for Failure of 16 Fuel Rods Unit 2 Cycle 10 calculated a maximum full power tilted RPF of 1.6010 (Reference 6. p, Section 4.8 and Table 4.8-1). Unit 3 Cycle 10 calculated a maximum full power tilted RPF of 1.6290 (Reference 6.Iq, Section 4.8 and Table 4.8-1). This calculation models a maximum full power tilted RPF of 1.71 when 16 fuel rod failures are considered. This RPF of 1.71 will be modeled for both once burned fuel rods and for high bumup fuel rods.
Per Unit 2 Cycle 10 Physics Input Design Analysis NFM-2-PH-0005 (Reference 6.1r, Section 4.3c and Table 3.3.3.3-4), the relative power fall-off curves show a maximum peaking factor drop-off for burnups greater than 31 GWD/MTU. As such, modeling of an RPF of 1.71 for high burnup fuel rods is conservative.
4.1.4.2 Average RPF for Failure of 210 Fuel Rods Per Unit 2 Cycle 10 Fuel Performance Analysis NFM-2-FP-0002 (Reference 6. k, Section 3.3.6), the ratio of the hottest rod normalized power to the normalized average power of 60 rods is greater than the pin-to-box ratio of 1.03 used in the FATES3B computer code average rod cases. As such, the power characteristics of 60 rods are representative of the power characteristics of an entire fuel assembly. For this reason, when 60 or more fuel rods in an assembly are failed, the average radial peaking factor of the failed rods is equivalent to the failed assembly's Relative Power Density (RPD).
Section 8.1 determines if it is conservative to model an RPD of once burned fuel or an RPD of high bumup fuel.
Based on Unit 2 Cycle 10 Models and Depletion Analysis NFM-2-PH-0003 (Reference 6.11, Section 4.15 and Appendix A Figures 5-3, 5-4 and 5-5), and based on Unit 3 Cycle 10 Models and Depletion Analysis NFM-3-PH-0003 (Reference 6.1m, Section 4.14 and Appendix A Figures 5-3, 5-4 and 5-5), the RPDs of once burned fuel assemblies were found to be below 1.40, and at the end of any cycle the highest RPD was found to be below 1.30. This calculation models an RPD of 1.37 when 210 once burned fuel rod failures are considered.
Based on Unit 2 Cycle 10 Models and Depletion Analysis NFM-2-PH-0003 (Section 4.15 and Appendix A Figures 5-3, 5-4 and 5-5), and based on Unit 3 Cycle 10 Models and Depletion Analysis NFM-3-PH-0003 (Reference 6.1m, Section 4.14 and Appendix A Figures 5-3, 5-4 and 5-5), the RPDs of high bumup fuel assemblies (i.e., assemblies that have been irradiated for more than one fuel cycle) were found to be below 1.30, and
E&TS DEPARTMENT CALCULATION SHEET ICCN NOJ PRELIM. CCN NO.
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Subject Fuel Handling Accident FHA) Inside Containment - Control Room & Offsite Doses Sheet 30 of 241 REV ORIGINATOR DATE I
IRE DATE REV ORIGINATOR DATE IRE DATE R
4 A. Rustae T. Remick I
at the end of any cycle the highest RPD was found to be at or below 1. 15. As such, it is conservative for this calculation to model a RPD of 1.37 when 210 high bumup fuel rod failures are considered. Of note is that the NRC Staff employed an assembly averaged high bumup radial peaking factor of 1.20 in the Safety Evaluation Report enclosed with an October 8, 1996 NRC letter (Reference 6.4o) issuing Amendment No. 132 to SONGS Unit 2 License NPF-10, and Amendment No. 121 to SONGS Unit 3 License NPF-15.
4.1.5 Fuel Rod Gap Iodine Inventory Species Composition. Per Section C. f of Regulatory Guide 1.25, the iodine gap inventory is composed of 99.75 percent inorganic species (i.e.,
elemental and particulate iodine), and 0.25 percent organic species (i.e., organic iodide). Per Assumption 3.3, it is assumed that this same relative proportion of inorganic and organic iodine species exists in both once burned and high burnup fuel.
4.2 Refueling Water Clean-Up of Isotopes 4.2.1 Refueling Water Iodine Removal. Per Regulatory Guide 1.25 Section C.lg (Reference 6.4i), the iodine gap activity released to the water is subject to clean-up due to pool scrubbing. The pool decontamination factors for the inorganic iodine and organic iodide species are 133 and 1, respectively. Per Regulatory Guide 1.25 Sections C.lc and C.lb, these iodine decontamination factors are valid for a minimum 23 foot water depth between the top of the damaged fuel rods and the fuel pool surface, and for a maximum fuel rod pressure of less than 1200 psig. Section 8.2 assesses design compliance with these conditions.
4.2.2 Refueling Water Noble Gas Removal. Per Section C.Ih of Regulatory Guide 1.25, the noble gas activity released to the pool is not subject to clean-up due to pool scrubbing. The retention of noble gases in the pool is negligible (i.e., a decontamination factor of one).
E&TS DEPARTMENT CALCULATION SHEET ICCN NOJ PRELIM. CCN NO.
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I CCN NO. CCN --
OUJie#
um Lnr4luline r
en* (MHAI Ineirla _rUntca:inmont -
-nnftre Pnem R. rffet-foln nVe QkHgt '21 f ')A4 REV ORIGINATOR DATE IRE DATE REV ORIGINATOR DATEI IRE DATE R
4 A. Rustae T. Remick V
._Y I
_mmh~
I
=~~W 4 4.3 Containment Cbaracteristics 4.3.1 Containment Free Volume. The gaseous activity released from the damaged fuel rods is diluted by the containment air volume. Per Assumption 3.5, the containment cooling train fans and the containment dome air circulators are not operating during MODE 6 (Refueling).
Therefore, no credit will be taken for activity dilution within the air of the hemispherical containment dome space. The effective containment free volume that will dilute the airborne activity is:
Vex = V V1 V
V
=
(
n Veff= V ( 2 3 dome
- where, Veff effective containment free volume in which activity is mixed (ft3)
V
=
minimum containment free volume = 2.305e6 ft3 (per Calculation C-257-01.06.01 [Reference 6.1 a, Section 2.1.1])
Vdome
=
containment dome volume Rde
=
containment dome radius = 74' 1 1 34" (per Sections B & E of Drawing 23060, 75'0" (radius) minus 1/4" (steel liner) [Reference 6.2d])
Therefore, Vear = 2.305e6f 3
- (l)(4),r(74.98t) 3 Ve = 2.305e6ft3
- 0.883e6ft3 = 1.422e6f 3 4.3.2 Containment Release Flow Rate. Per Assumption 3.4, the containment airborne activity is released from the containment building over a two hour period. In this calculation this requirement will be met by modeling a containment release flow rate large enough to ensure that at least 99.9 percent of the containment airborne activity will have been released from the containment within the first two hours of the fuel handling accident. The containment release flow rate necessary to achieve this removal effectiveness can be determined through use of the time dependent activity removal equations defined in Introduction to Health Physics (Reference 6.6k, page 73, Equations 4.18 and 4.19):
Al A ei' whiere X
Iiii A
N At-0
~At 0
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Subject Fuel Handling Accident (FHA) Inside Containment - Control Room & Offsite Doses Sheet 32 of 241 REV ORIGINATOR DATE IRE DATE REV ORIGINATOR DATE IRE DATE R
Ei mm_____
4.Rustae T. Remick j
Vn Disregarding radioactive decay, the change in the number of radioactive atoms (AN) is equivalent to the product of the initial atom concentration (atoms per unit volume), the containment release rate (volume per unit time), and the given time interval (At):
AN = - (
°
) x Flowrate x At Volume Nv x Flowrate x At)
N
-(N0 Volume x tI
.. ) =
limit At-0
= Flowrate Volume At Flowrate xme A, = A e
= A e Volume For the Design Input 4.3.1 effective containment free volume of 1.422e6 ft3, the containment release flow rate necessary to achieve a 99.9 percent removal effectiveness (i.e., at the end of two hours, the activity in the containment [A,] will be 0.1% of the initial activity [A.] in the containment) is:
Flowrte. Tme 0.001 = e Volume
_ F_______
Flow rate x 1 0 m ln(O.001)
- Flowrate x Time = _x 120 min Volume I.422e6 ft 3 Flowrate = - ln(0.00) x x1.422e6 ft 3 120 min Flowrate = 81857cfm = 82000cfm For conservatism, the release flow rate from the containment to the outside environment will be rounded up and modeled as 82,000 cfm (equivalent to 38.7 m3/sec).
As confirmed by the Section 9 LOCADOSE code output, using this release flow rate and considering radioactive decay yields a containment airborne activity at 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> that is less than 0.1% of the initial containment airborne activity. This validates the use of the modeled containment release flow rate.
E&TS DEPARTMENT CALCULATION SHEET Project or DCPIFCNIECP Calc. No. N4072-003 0-1--f C-a Unftl1n AMir-#
ICLIAI neiArlv f~ntevM~
0ontro loo t
McF nnc-Theis 2
f 15A
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ll%
vr w
I lrPr l loluxw, ~Wl ILOSI IIII-l IL -
.*JI MlVI II o
IJI IvVL
,JJ WeI 414W I REV ORIGINATOR DATE IRE DATE REV ORIGINATOR DATE IRE DATE R
4 A. Rusta T. Remick I
I v
=
I
{~~~~~~
4.4 Control Room Characteristics 4.4.1 Control Room Net Free Volume. The control room net free volume is 266,920 ft3 per Calculation M-0073-095 (Reference 6.1c, CCN-3 page A6-3).
4.4.2 Control Room Charcoal and HEPA Filter Efficiencies. Per Control Building Air Flow Diagram Drawings 40096 (Reference 6.2i) and 40098 (Reference 6.2j), in the high radiation isolation mode the makeup air entering the control room passes through filters SAI 51 OMA206 and SA1510MA207 of the Emergency Ventilation System (EVS) train, and then through filters SAI5IOME418 and SAI5 10ME419 of the Emergency Air Conditioner (EAC) filter train.
However, per UFSAR Appendix 15B (Reference 6.4g), credit is only taken for the EAC filters.
The recirculation air passes only through the EAC filter train.
Table 4.4-1 presents the elemental, organic, and particulate iodine removal efficiencies modeled for the LOCADOSE Code intake and recirculation filters. The EAC iodine filter efficiency is based on a charcoal filter depth of 2 inches (per an E-mail from D. Higgins to F. Santa Ana
[Reference 6.3b]) and the corresponding charcoal filter efficiency values given in Regulatory Guide 1.52 Table 2 (Reference 6.4j). The EAC particulate efficiency is based on the HEPA efficiencies given in Regulatory Guide 1.52 Section C.5c. Regulatory Guide 1.52 efficiencies are applicable since Technical Specification 5.5.2.12 (References 6.4c and 6.4d) tests the EAC filters to this Regulatory Guide.
Table 4.4 Control Room Charcoal and HEPA Filter Efficiencies Contaminant EVS Filter EAC Filter Intake Recirculation Beaing n
Removal Removal Removal Efficiency RmvlEfficiency RB ved Efficiency Efficiency modeled in the modeled in the Removed (A206/A207)
(E418/E419)
LOCADOSE Code LOCADOSE Code Elemental Iodine 0%
95%
95%
[
95%
Organic Iodide 0%
95%
95%
95 Particulates 9
/0 990/0 99%_99
E&TS DEPARTMENT CALCULATION SHEET Project or DCP/FCNIECP C
Subject Fuel Handling Accident FHA) Inside Containment - Control Room & Offsite Doses Sheet 34 of 241 REV ORIGINATOR DATE IRE I
DATE REV ORIGINATOR DATE IRE DATE R
1+/- A.Rustae T. Remick v
'---By
=
JJ~~~~~~~~~~~~
4.4.3 Control Room HVAC Operation Characteristics. The control room HVAC parameters during normal and high radiation isolation operation are presented in Tables 4.4-2 & 4.4-2a; the HVAC configuration is shown on Control Room Complex HVAC Drawings 40173A (Reference 6.2a), 401 73AS03 (Reference 6.2b), and 401 73C (Reference 6.2c).
To conservatively maximize the introduction of airborne radionuclides into and retention within the control room, the high radiation isolation mode outside air intake flow rate will be increased to its upper tolerance of 2050 +/-150 cfm as given in Calculation M-0073-041 (Reference 6.lb, CCN N-7, page 15), and the recirculation flowrate will be decreased to its lower tolerance of 35,705 cfm +10% as given in Calculation M-0073-041 (CCN N-7, page 15). These flowrates and their tolerances are consistent with those given in Maintenance Procedure S023-1-2.44 (Reference 6.5a).
Per Assumption 3.14, consistent with UFSAR Appendices 1 5B and 5.1 OB, this analysis models two trains of emergency HVAC in operation during the first 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> of the accident. Operator action is assumed to be taken within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> to deactivate one train of emergency intake and recirculation units.
I I
I I
Table 4.4 Control Room HVAC Operation Flowrates and Inleakage Rates Parameter CR Normal HVAC Operation CR High Radiation Isolation Operation (single train operation flowrates)
(single train operation flowrates)
Value
[
Reference Value l
Reference Filtered 0 cfm dwg 40096 (Ref. 6.2i) 2050 + 150 Calc M-0073-041 (Ref 6. lb)
Intake 2200 cfm dwg 40096 (Ref 6.2i) dwg 40098 (Ref 6.2j)
S023-1-2.44 (Ref 6.5a)
Unfiltered 5820 cfm dwg 40096 (Ref. 6.2i) 0 (inflow) + 10 (inleakage) dwg 40096 (Ref 6.2i)
Intake
= 10 cfm dwg 40098 (Ref 6.2j)
Assumption 3.9 Filtered Ocfm dwg 40096 (Ref. 6.2i)
(35,705 - 10%) - 2200 Calc M-0073-041 (Ref 6.1b)
Recirculation
= 29,934.5 cfm '1) dwg 40096 (Ref 6.2i) dwg 40098 (Ref 6.2j)
S023-1-2.44 (Ref 6.5a)
Unfiltered 29,885 cfm dwg 40096 (Ref. 6.2i) 0 cfm dwg 40096 (Ref. 6.2i)
Recirculation dwg 40098 (Ref. 6.2j)
(a)
The filtered recirculation flowrate is rounded down to 29,934 cfm for single train operation, and modeled as 59,870 cfm for two trains operation.
I
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Q.H.k;,., Csil 14nl Ar.,44e-rl fFCIA% Ineil r.nntniVnm n -
frnItl Deunm k f cit, Flrneme Qhkeft
'AK nf 15A4 ouLDJ~u t
I Eu fllu 1UU-:
t II U
- lj a
E5uI.J.
Eta U
.E Ev W 'r..r EE.aEt v
%.F:;Gt
%I I I REV ORIGINATOR DATE IRE DATE REV ORIGINATOR DATE IRE DATE Ri A. Rustae T. Remick I
E I
Table 4.4-2a Control Room HVAC Operation Flowrates and Increased Inleakage Rates I
I Parameter CR Normal HVAC Operation CR High Radiation Isolation Operation (single train operation flowrates)
(single train operation flowrates)
Value I
Reference Value I
Reference Filtered 0 cfm dwg 40096 (Ref. 6.2i) 2050 + IS0 = 2200 cfm Calc M-0073-041 (Ref 6.1b)
Intake dwg 40096 (Ref 6.2i) dwg 40098 (Ref 6.2j)
S023-1-2.44 (Ref 6.5a)
Unfiltered 5820 (inflow) + 1000 dwg 40096 (Ref. 6.2i) 0 (inflow) + 1000 (inleakage) dwg 40096 (Ref 6.2i)
Intake (inleakage) = 6820 cfm Assumption 3.9 1000 cfm dwg 40098 (Ref 6.2j)
Assumption 3.9 Filtered 0 cfm dwg 40096 (Ref. 6.2i)
(35,705 - 10%) - 2200 Calc M-0073-041 (Ref 6.1b)
Recirculation
- 29,934.5 cfm "
dwg 40096 (Ref 6.2i) dwg 40098 (Ref 6.2j)
S023-1-2.44 (Ref 6.5a)
Unfiltered 29,885 cfrn dwg 40096 (Ref. 6.2i) 0 cfrn dwg 40096 (Ref. 6.2i)
Recirculation dwg 40098 (Ref. 6.2j)
(a)
The filtered recirculation flowrate is rounded down to 29,934 cfrn for single train operation, and modeled as 59,870 cfm for two trains operation.
I I
I I
I I
I I
I I
I I
I I
I I
I I
I I
I I
I
E&TS DEPARTMENT CALCULATION SHEET CCN NO./
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Cl.k;ars FI.DI Uftnnin^ Ar-A~n IUAI ncirm
-nirbm
-A*t
_oS..1 DEUrfl'e gi Anna
- km*
U1tQ ff OiA4 REV ORIGINATOR DATE IRE DATE REV ORIGINATOR DATE IRE DATE R
4 A. Rustae T. Remick v
E
_ I 4.5 Control Room Radiation Monitor Characteristics 4.5.1 Transfer to High Radiation Isolation Mode. Units 2&3 Technical Specification LCO 3.3.9 (References 6.4c & 6.4d) requires that one Control Room Isolation Signal (CRIS) channel shall be operable in MODES 1 through 6, and during the movement of irradiated fuel assemblies. Per Elementary Drawings 31394 (Reference 6.2k) and 31395 (Reference 6.21), the control room will automatically transfer to the high radiation isolation mode if a Safety Injection Actuation Signal (SIAS) is generated, or if radiation monitor 2/3RE782G1 or 2/3RE7825G2 senses high radiation.
4.5.2 Radiation Monitor Higb Radiation Trip Initiation Time. Calculation J-SPA-179 (Reference 6. 1j) shows that the control room radiation monitor trip initiation time is based on the plenum fill time and the radiation monitor algorithm response time.
Plenum Fill Time. The plenum fill time is the time required for the activity concentration inside the control room HVAC intake plenum to build up to the outside activity concentration present at the control room HVAC intake louvers. Per Calculation J-SPA-179 Section 8.16.1, the intake plenum activity reaches 90 percent of the incoming outside air activity concentration within 8.5 seconds.
Radiation Monitor Algorithm Response Time. The radiation monitor algorithm response time is the time required for the count rate averaging algorithm to produce a representative concentration value that is greater than the monitor trip set point. Per Calculation J-SPA-179 Section 8.16.2, the monitor response time is a function of the plenum activity concentration as presented in Figure 4.5-1. For any plenum activity concentration greater than 8.5e-7 pCi/cc the monitor response time will be less than 60 seconds. Note that increasing activity causes the response time to decrease to a minimum detector response time of 3 seconds. Consistent with Calculation J-SPA-1 79 Section 8.16.2, an additional allowance, to a minimum time of 60 seconds is used for the algorithm response time.
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R A. Rustae IT. Remick l
l l
v Figure 4.5-1: Control Room Monitor Algorithm Response Time
[per Calculation J-SPA-179 Figure 8.16.2]
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Project or DCPIFCNJECP Calc. No. N-4072-003 CC4 CONVERSION:
I CCN NO. CCN-Auiha P
a4at Ia nAlin Anrldent fT-ILS In side Contatnment _ Control onm IL Affsite ne Shaat QSR f 9A V
REV ORIGINATOR DATE I
IRE DATE REV ORIGINATOR DATE IRE DATE 4 A. Rustae
- r. Remick I
_u I
4.6 Occupancy Factors The control room occupancy factors (CROFs) are given in Standard Review Plan 6.4 Figure/Table 6.4-1 (Reference 6.40). Per 10 CFR 100.11 (Reference 6.4b), the offsite post-accident doses are to be calculated for an individual who is exposed for the entire interval of interest (i.e., an occupancy factor of 1,0). Based on the time intervals of interest specified in the Section 1.2 dose criteria, the occupancy factors are:
Exclusion Area Boundary Low Population Zone Boundary 1.0 for 0 hour0 days <br />0 hours <br />0 weeks <br />0 months <br /> to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 1.0 for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> 1.0 for 0 hour0 days <br />0 hours <br />0 weeks <br />0 months <br /> to 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> Control Room 1.0 0.6 0.4 for 0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> for 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> to 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> N-4072-003
E&TS DEPARTMENT CALCULATION SHEET ICCN NOJ PRELIM. CCN NO.
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Subject Fuel Handling Accident (FHA) Inside Containment - Control Room & Offsite Doses Sheet 39 of 241 REV ORIGINATOR DATE IRE DATE REV ORIGINATOR DATE IRE DATE R
1 A. Rusta T. Remick I
4.7 Atmospheric Dispersion Factors Table 4.7-1 presents the San Onofre site-specific five percentile meteorology atmospheric dispersion factors (X/Q) for design basis releases from the Containment to the control room and offsite EAB and LPZ dose receptors. The Control Room x/Qs are per Calculation N-4010-001 (Reference 6. d, CCN-2 Section 2.0), and are presented with and without the control room occupancy factors (CROFs) defined in Design Input 4.6. The offsite x/Qs are per Calculation N-4010-002 (Reference 6.le, Revision 1 page 3).
The offsite x/Qs presented in Table 4.7-1 differ from those used by the NRC Staff in their independent Safety Evaluation Report analysis of this fuel handling accident inside containment event (Reference 6.4o). However, the San Onofre site-specific X/Qs are consistent with UFSAR Appendix 1 SB (Reference 6.4g, Table I SB-4), which presents the models used to calculate offsite radiological doses that would result from the release of radioactivity due to various postulated accidents, including a FHA.
Table 4.7 Atmospheric Dispersion Factors Receptor Design Basis Case Five Percentile X/Q (sec/m3) 0-2 Hours l 2-8 Hours 1 8-24 Hours l 24-96 Hours 196-720 Hours Control Room 3.1 e-03 3.1 e-03 I.Se-03 5.9e-04 9.6e-OS (with CROF)
Control Room 3.1e-03 3.le-03 I.8e-03 9.8e-04 2.4e-04 (without CROF)
EAB 2.72e-04 Not Applicable Not Applicable Not Applicable Not Applicable LPZ 7.72e-06 7.72e-06 4.74e-06 3.67e-06 2.67e-06
E&TS DEPARTMENT ICCN NOJ1 CALCULATION SHEET IPRELIM. CCN N1 rroject or DCP/FCN(ECP_
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4.8 Dose Conversion Factors Table 4.8-1 lists the dose conversion factors (DCFs) modeled in this calculation as input into the LOCADOSE computer program (Reference 6.6a) library file listed in Section 9.1. Table 4.8-1 lists only those iodine and noble gas isotopes which are present in the damaged fuel rod source tenn after 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of decay (as identified in Design Input 4.1). DCFs are not presented for the particulate daughter products, since particulate isotopes are not released to the environment.
The thyroid inhalation DCFs are consistent with the requirements of the Technical Specification Section 1.1 definition for "DOSE EQUIVALENT I-131" (References 6.4c and 6.4d). Per this definition, the thyroid DCFs shall be those listed in ICRP Publication 30, Supplement to Part 1, pages 192-212, in tables titled "Committed Dose Equivalent in Target Organs or Tissues per Intake of Unit Activity" (Reference 6.60. The ICRP-30 iodine inhalation DCFs are presented in units of Sieverts per Becquerel (SvIBq). Table 4.8-1 converts these DCFs to units of rem per curie as required for input into the LOCADOSE computer program by using conversion factors of 3.7elO Bq/ci, and 100 rem/Sv.
The default whole body gamma immersion and beta skin immersion DCFs in the Bechtel LOCADOSE computer program will be used in this calculation. These DCFs are consistent with UFSAR Appendix I 5B Table 1 5B-6 (Reference 6.4g).
E&TS DEPARTMENT CALCULATION SHEET j ICCN NOJ l
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project or DCP/FCN/ECP Calc. No. N-4072-003 CCN CONVERSION:
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Subiect Fuel Handlina Accident (FHA) Inside Containment - Control Room & Offsite Doses Sheet 41 of 241 REV ORIGINATOR DATE IRE DATE REV ORIGINATOR DATE IRE DATE R
4 A. Rustaey T. Remick E
r X~~~~~~~~~~~~~~~~~~~
TABLE 4.8 Dose Conversion Factors Isotope Thyroid Beta Skin Whole Body Gamma nhalation DCF (
Immersion DCF Immersion DCF (Sv/Bq).(rem/curie)
(rem-m3 /curie-sec)
(rem-r 3 /curie-sec)
Iodine-129 1.6e-06 5.92e+06 3.710e-04 3.024e-03 lodine-I 30 2.0e-08 7 AOe+04 4.990e-02 4.980e-01 Iodine-131 2.9e-07 1.07e+06 3.170e-02 8.720e-02 Iodine-I 32 I.7e-09 6.29e+03 1.320e-01 5.130e-01 lodine-133 4.9e-08 1.81 e+05 7.350e-02 I.550e-01 Iodine-135 8.5e-09 3.15e+04 1.290e-O 4.210e-O Kiypton-83m 0
0 0
2.396e-06 Krypton-85 0
0 4.246e-02 5.102e-04 Krypton-85m 0
0 4.626e-02 3.708e-02 Krypton-87 0
0 3.083e-01 1.876e-01 Krypton-88 0
0 7.510e-02 4.658e-01 Xenon-13 1m 0
0 1.508e-02 2.899e-03 Xenon-133m 0
0 3.150e-02 7.954e-03 Xenon-133 0
0 9.697e-03 9.316e-03 Xenon-135m 0
0 2.253e-02 9.887e-02 Xenon-1 35 0
0 5.894e-02 5.736e-02 (a) Sample Calculation for Iodine-1 31 Thyroid Inhalation Dose Conversion Factor
= (2.9e-7 Sv/Bc) x (3.7elOBalci) x (10o rem/Sv)= 1.07e6rem/ci
E&TS DEPARTMENT CALCULATION SHEET ICCN NO./
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a oject or DCP/FCNIECP Calc. No. N-4072-003 CCN CONVERSION:
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Si ehirt Fi eII Hnneilinn Accident IFHA1 Insiria Cnntainment - Cnntrnl Pnnm. Offsite Dlnesp Sheet A9 nf A1 J--,
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4.9 SOURCE2 Code Source Strength Spectrum Energy Structure The gamma energy source strengths from the SOURCE2 output will be grouped into a "BASE1O" gamma energy structure. The "BASE10" gamma energy structure is presented in Table 4.9-1. This energy structure is part of the SOURCE2 Code library (Reference 6.6c, Table 2-5).
TABLE 4.9 "BASE10" Gamma Energy Structure Energy Group Gamma Energy Ranges Number (MeVy-disintegration)
I 0.0 sE sO.1 2
0.1 <EsO.4 3
0.4 < Es 0.9 4
0.9<Es 1.35 5
1.35 < E s 1.80 6
1.80 < E s 2.20 7
2.20 < E s 2.60 8
2.60 < E s3.00 9
3.00 < E s5.00 10 5.00<Es 15.00
E&TS DEPARTMENT CALCULATION SHEET ICCN NOJ PRELIM. CCN NO.
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_ CCN CONVERSION:
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4.10 Containment Shine Dose Multiplication Factors - Control Room Per Calculation N-0450-002 (Reference 6.1 f), the containment shine dose multiplication factors (DMFs) as functions of BASE1O gamma energies are given in Table 4.10-1 for the 16 Control Room dose locations identified in Assumption 3.17. These DMFs are in units of Rem/hr per MeV/cc-sec of containment cloud volumetric source strength. The dose point 9, 10, 15 and 16 locations represent the control board area that will be used to represent the Control Room Operator doses (per Assumption 3.17). It is noted that these DMFs were calculated with a containment free volume of 2.284e6 ft' rather than the 2.305e6 ft3 given in Design Input 4.3.1; this modeling is acceptable as discussed in Calculation N-0450-002 Design Input 4.14.
Table 4.10 Containment Shine Dose Multiplication Factors BASEIO Effective Containment Airborne Cloud Shine Dose Multiplication Factors Gamma Energy Gamma
[Remn/hr per MeV/ec-sec I Range Energy____
IMeVl I MeVI at DP I
&tDP 2 at DP 3 atDP4 atDP5 tDP6 atDP7
&tDP8 0.0s E s 0.1 0.1 l.OOE-38 I.OOE-38 L.OOE-38 1.00E-38 l.OOE-38 L.OE-38 l.OOE-38 l.OOE-38 0.1 <E sO.4 0.4 7.91E-32 2.43E-23 S.i3E-23 5.17E-23 4.73E-23 9.09E-23 8.68E-23 8.00E-23 0.4 < E s 09 0.8 2.08E-25 8.66E-49' 1.61E-18 1.70E-18 1.72E-18 2.89E-18 2.54E-18 2.24E-18 0.9 < E s 1.35 1.3 2.64E-21 6.18E-16 1.07E-15 1.16E-15 1.24E-15 1.93E-lS 1.60E-15 1.37E.15 1.35 < Es 1.8 1.7 1.98E-19 I.ISE-14 192E-14 2.13E-14 2.35E-14 3.52E-14 2.82E-14 2.38E-14 1.8< E 2.2 2.18 8.75E-18 1.47E-13 2.37E-13 2.66E-13 3.02E-13 4.38E-13 3.42E-13 2.85E-13 2.2 < E s 2.6 2.S 5.79E-17 5.12E-13 8.19E-13 9.25E-13 1.06E-12 I.52E-12 1.17E-12 9.66E-13 2.6 < E s 3.0 2.8 2.22E-16 1.23E-12 l.95E-12 2.21E-12 2.55E-12 3.62E-12 2.76E-12 2.27E-12 3.0 < E s 5.0 4.0 9.44E-15 1.3SE-11 2.08E-11 2.40E-11 2.84E-11 3.93E-11 2.82E-1 2.29E-ll 5.0 < E s 10.0 6.2 2.76E-13 l.09E-10 1.63E-10 1.91E-10 2.32E-10 3.12E-10 2.21E-10 1.77E-10 BASE10 Effective Containment Airborne Cloud Shine Dose Multiplication Factors Gamma Energy Gamma
,Rem/hr per MeV/cc-seI Range Energy IMCVI I MVl I atDP9 atDPiO atDPIl 1
atDP12 atDP13 atDP14 atDPI5 atDP16 0.0 s Es 0.1 0.1 1.OOE-38 ll.00E-38 1L.OOE-38 I.00E-38 1.00E-38 1.00E-38 I.00E-38 1.0E-38 0.1 < E s 0.4 0.4 9.96E-23 1.21E-22 1.43E-22 1.49E-22 2.32E-22 1.91E-22 1.28E-22 9.15E-23 0.4 < Es 0.9 0.8 2.55E-1 B 3.20E-8 4.05E-18 4.58E-18 6A471-I8 4.90E-1 8 3.16E-I8 2.26E-1IS 0.9< Es 1.35 1.3 1.46E-15 1.88E-15 2.50E-15 3.01E-15 3.98E-I15 2.82E-15 1.79E-15 1.27E-15 1.35 <E s 1.8 1.7 2.50E-14 3.20E-14 4.38E-14 5.44E-14 6.96E-14 4.77E-14 3.01E-14 2.11E-14 1.8< E s 2.2 2.18 2.92E-13 3.79E-13 5.27E-13 6.76E-13 8.35E-13 5.57E-13 3.48E-13 2.44E-13 2.2< E s 2.6 2.5 9.79E 13 1.27E-12 1.80E-12 2.34E-12 2.84E-12 1.88E-12 1.14E-12 8.17E-13 2.6 < E s3.0 2.8 2.28E-12 2.98E-12 4.23E-12 5.56E-12 6.70E-12 4.39E-12 2.66E-12 1.90E-12 3.0< E s5.0 4.0 2.30E-II T3.03E-1I 4.42E-I 6.02E-1 17.01E-11 4.44E-1I 2.68E-1 1.91E-1I S.0< E s 10.0 l
6.2 1.69E-10 2.24E-I0 3.32E-l0 4.79E-10 5.40E-l0 3.32E-10 1.98E-I I.43E-l0J
E&TS DEPARTMENT CALCULATION SHEET ICCN NOJ PRELIM. CCN NO.
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Subject Fuel Handling Accident (FHA) Inside Containment - Control Room & Offsite Doses Sheet 44 of 241 REV OIIAO DATE IRE DATE R
EV ORIGINATOR__
DAE IE DATE R
1 r
~~~~~~~~~-E l 4 A. utey I
[T. Remick H
zVi+/-
m
-~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~I 4.11 Environmental Cloud Shine Dose Multiplication Factors - Control Room Per Calculation N-0450-002 (Reference 6.1f), the environmental cloud shine dose multiplication factors (DMFs) as functions of BASEI0 gamma energies are given in Table 4.11-1 for the 16 Control Room dose locations identified in Assumption 3.17. These DMFs are in units of Rem/hr per MeVtcc-sec of containment cloud volumetric source strength. The dose point 9, 10, 15 and 16 locations represent the control board area that will be used to represent the Control Room Operator doses (per Assumption 3.17). Per Calculation N-0450-002, these DMFs consider the presence of contaminated air in the cable riser galleries to the north and south of the control room.
Table 4.11 Environmental Cloud Shine to Control Room Dose Multiplication Factors BASE10 Effective Environmental Cloud Shine Dose Multiplication Factors Gamma Energy Gamma jRenx/r per MeV/cc-sec I Range Energy IMeV I McVi at DP 1 atDP2 itDP3 at DP 4 atDPS at DP 6 t DP 7 at DP 8 0.0 E s 0.1 0.1 2.74E-04 2.36E404 1.13E&04 9.33E.05 6.76E-OS 5.00E-05 6.64E-05 6.26E-05 0.1 < E s 0.4 0.4 454E-04 3.61E4D4 1.83E-04 1.69E-04 1.55E.04 1.03E-04 I.20E-04
.I1OE-04 0.4 < Es 0.9 0.8 4.36E-04 3.44E-04 1.78E-04 1.71E-04 1.65E-04 1.12E-04 1.23E-04 1. IE-04 0.9< Es 1.35 1.3 4.34E.04 3A1E-04 1.81E04 1.79E.04 l.80E-04 1.27E-04 1.32E-04 l.19E-04 1.35 < Es i.8 1.7 431E-64 339E-04 I.83E-04 1.84E.04 1.88E-04 136E-04 138E-04 1.23E-04 1.8 < E s 2.2 2.18 439E-04 346E-04 U91E-04 1.95E-04 2.03E-04 J.49E-04 1.48E-04 132E-04 2.2 < E s 2.6 2.5 4.44E-04 3.50E-04 1.97E-04 2.02E-04 2.1 E-04 I S8E-04 1.55E-04 1.38E-04 2.6 < E s 3.0 2.8 4A5E-04 3.52E-04 2.OOE-04 2.05E-04 2.16E-04 1.63E-04 1.59E-04.142E-04 3.0 < E s 5.0 4.0 4.49E-04 3.59E-04 2.12E-04 2.20E-04 2.35E-04 1.81E-04 1.73E-04 I.55E-04 5.0 < E s 10.0 6.2 4.66E-04 3.81E-04 2.35E-04 2.46E-04 2.67E-04 2.1 IE-04 1.9E-04 1.77E-04 BASE10 Effective Environmental Cloud Shine Dose Multiplication Factors Gamma Energy Gamma lRem/hr per MeV/cc-sec Range Energy IMeVi I MeVl at DP 9 at DP 10 at DP 11 at DP 12 at DP 13 at DP 14 at DP 15 at DP 16 0.0 s E s 0.1 0.1 5.22E-05 6.42E-05 5.95E-05 1.22E-04 1.22E-04 6.05E-05 5.78E-05 5.OOE-05 0.1 <E s 0.4 0.4 8.90E-05
- 1. 1 IE-04 1.01 E-04
.68E-04 I.59E-04 9.35E-05 9.66E-05 8.44E-05 0.4 < E s 0.9 0.8 9.26E-05 1.1 3E-04
.06E-04 1.57E-04 I.46E-04 9.59E-05 9.89E-05 8.82E-05 0.9 < E s 1.35 1.3 1.03E-04
.22E-04 1i18E-04 I.60E.04 E48E-04 1.07E-04
.09E-04 9.92E-05
.35 < Es 1.8 1.7 1.IOE-04
.29E-04 1.26E-04 1.63E-04 1.51E-04 1.14E-04
.72E-04 1.59E-04 1.26E-04 1.27E-04 1.1I8E-04 2.2 < E s 2.6 2.5 1.27E-04
.47E-04
.46E-04 1.77E-04 1.64E-04 1.33E-04 1 34E-04 11.25E-04 2.6 < Es 3.0 2.8 1.32E-04
.51E-04
.51E-04
.81E-04 1.68E-04 1.38E-04 138E-04 1.30E-04 3.0 < E s 50 4.0 1.48E-04 1.67E-04
.69E-04 1.94E-04 1.81E-04 1.55E-04 1.55E-04 1.47E-04 5.0 < E s 10.0 6.2 1.75E-04 1.93E-04 1.98E-04 2.17E-04 2.04E-04 1.82E-04 1.82E.04 1.74E-04
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Subject Fuel Handling Accident (FHA) Inside Containment - Control Room & Offsite Doses Sheet 45 of 241 REV ORIGINATOR DATE IRE DATE REV ORIGINATOR DATE IRE DATE R
4 A.
Rustae T.
Remick IZ ZIJI I
I IE 4
A. Rustaey T. Remick l
V I
$~~~~
l l
_~~~~~~~~~~~~~~~~~~~~~~~~~
4.12 Control Room Filter Shine Dose Multiplication Factors Per Calculation N-4060-016 (Reference 6. lh), the charcoal filters A206 and A207 shine dose multiplication factors (DMFs) as functions of BASEIO gamma energies are given in Tables 4.12-1 and 4.12-2, respectively, for the 16 Control Room dose locations identified in Assumption 3.17. These DMFs are in units of Rem/hr per MeV/cc-sec of charcoal filter volumetric source strength. The dose point 9, 10, 15 and 16 locations represent the control board area that will be used to represent the Control Room Operator doses (per Assumption 3.17).
Table 4.12-1 -- Charcoal Filter A206 Shine Dose Multiplication Factors BASE10 Effective Filter A206 Shine Dose Multiplication Factors Gamma Energy Gamma lRem/hr per MeV/cc-sec 1 Range Energy IMeV]
I MeVI atDPI atDP2 satDP3 atDP4 atDP5l atDP6 atDP7 at DP8 0.0 s E s 0.1 0.1 7.93E-07 3.31E-09 4.19E-10 1.02E-10 6.42E-09 1.38E-09 4.45E-10 1.13E-10 0.1 <E s 0.4 0.4 1.52E-06 7.54E-09 2.54E-09 1.95E-09 2.36E-08 7.67E-09 3.33E-09 1.41E-09 0.4 < E s 0.9 0.8 1.31E-06 6.77E-09 2.92E-09 2.85E-09 2.27E-08 8.1SE-09 3.79E-09 1.B4E-09 0.9 < E 1.35 1.3 l.15E-06 6.06E-09 3.04E-09 3.39E-09 2.11E-08 8.09E-09 3.89E-09 2.05E09 1.35 <E s 1.8 1.7 1.06E-06 5.60E-09 2.99E-09 3.52E-09 1.98E-08 7.79E-09 3.80E-09 2.07E-09 1.8 < E s 2.2 2.18 9.83E-07 5.25E-09 2.96E-09 3.64E-09 I.88E-08 7.55E-09 3.74E-09 2.09E-09 2.2 < E s 2.6 2.5 9.43E-07 5.06E-09 2.93E-09 3.68E-09 1.82E-08 7.39E-09 3.68E-09 2.09E-09 2.6 < E s 3.0 2.8 9.05E-07 4.86E-09 2.86E-09 3.64E-09 1.76E-08 7.17E-09 3.58E-09 2.05E-09 3.0 < E s 5.0 4.0 7.97E-07 4.29E-09 2.63E-09 3A6E-09 1.57E-08 6.49E-09 3.27E-09 1.91E-09 5.0 < E s 10.0 6.2 6.92E-07 3.74E-09 2.36E-09 3.20E-O9 1.38E-08 5.77E-09 2.93E-09 1.74E-091 BASE10 Effective Filter A206 Shine Dose Multiplication Factors Gamma Energy Gamma lRemlhr per McV/cc-sec I Range Energy 1
IMeVI I MeVJ atDP9 atDP 10 atDPll atDP12 atDP13 atDP14 atDPI5 atDP 16 0.0 s E s 0.1 0.1 2.19E-10 6.15E-10 6.96E-10 2.79E-09 1.75E-09 3.72E-10 4.41E-10 3.39E-10 0.1 <E S 0.4 0.4 1.74E-09 3.28E-09 4.19E-09 7.97E-09 5.12E-09 2.54E-09 2.43E-09 I.90E-09 0.4 < E s 0.9 0.8 2.02E-09 3.45E-09 4.58E-09 7.57E-09 4.91E-09 2.89E-09 2.58E-09 2.02E-09 0.9 <E s 1.35 1.3 2.09E-09 3.38E-09 4.60E-09 7.03E-09 4.57E-09 2.97E-09 2.55E-09 2.OOE-09 1.35 < E s 1.8 1.7 2.05E-09 3.24E-09 4.45E-09 6.60E-09 4.30E-09 2.90E-09 2.45E-09 1.92E-09 1.8 < E s 2.2 2.18 2.02E-09 3.1 3E-09 4.34E-09 6.27E-09 4.09E-09 2.85E-09 2.37E-09 I.86E-09 22< E s 2.6 2.5 l.99E-09 3.06E-09 4.26E-09 6.08E-09 3.97E-09 2.81E-09 2.32E-09
.82E-09 2.6 < E s 3.0 2.8 1.95E-09 2.97E-09 4.14E-09 5.86E-09 3.83E-09 2.73E-09 2.25E-09 1.77E-091 3.0 < E s 5.0 4.0 1.78E-09 12.68E-09 3.76E-09 5.23E-09 l3.42E-09 2.50E-09 2.04E-09 I.60E-09 5.0 < E 10.0 6.2 1.60E-09 2.38E-09 3.35E-09 4.61E-09 3.02E-09 2.24E-09 1.81 E-09
.42E-09
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Project or DCP/FCN/ECP Calc. No. N4072-003 CCN CONVERSION:
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Cieft LJnn,.r-
^' A tn ICU A% triciaI
- r.
- n.
t-rAJ 0~r k frVeifem frmem A9~ctR f 14
%JUuJ Lt SUel *tusMlt SI titL I
I IWI It..fl.JI
%as I
WIt u
l sU 96rvu.,
e u
tt::
s,aCn Ul I
REV ORIGINATOR DATE IRE DATE REV ORIGINATOR DATE IRE DATE R
A. Rustae T. Remick E
Table 4.12 Charcoal Filter A207 Shine Dose Multiplication Factors BASEIO Effective Filter A207 Shine Dose Multiplication Factors Gamma Energy Gamma _Remlbr per MeV/cc-sec Range Energy_________
IMeVI I MeV1 at DP I at DP 2 at DP 3 at DP 4 fatDPS atDP6 at DP 7 atDP8 0.0! E s 0.1 0.1 1.05E-14 4.29E09 7.29E-10 4AIE-10 5.04E-10 3.62E-11 1.44E-10 2.76E-10 0.1 < E s 0.4 0.4 3.88E-12 9.62E-09 3.84E-09 2.36E-09 1.74E-09 4.96E-10 1.36E-09 2.35E-09 OA E s 0.9 0.8 1.67E-11 8.59E-09 4.19E-09 2.61E-09 1.75E-09 7.DlE-10 1.69E-09 2.78E-09 0.9 < Es 1.35 1.3 4.12E-11 7.68E-09 4.22E-09 2.64E-09 I.69E-09 823E-10 1.82E-09 2.92E-09 1.35 < E s 1.8 1.7 6.00E-11 7.09E-09 4.09E-09 2.56E-09 1.61E-09 8.52E-10 1.82E-09 2.88E-09 1.8 < E s 2.2 2.18 836E-11 6.64E-09 4.00E-09 2.51E-09 1.55E-09 8.80E-10 1.82E-09 2.85E-09 2.2 < E s 2.6 2.5 9.77E-11 6.39E-09 3.93E-09 2A7E-09 1.5E-09 8.87E-10 1.81E-09 2.82E-09 2.6 < E s 3.0 2.8 1.07E-10 6.14E-09 3.82E-09 2.40E-09 1.46E-09 8.78E-10 1.78E-09 2.75E-09 3.0 < Es 5.0 4.0 1.35E-10 SA2E-09 3A8E-09 2.19E-09 1.32E-09 8.36E-10 1.65E-09 2.53E-09 5.0< Es 10.0 6.2 1.58E-10 4.73E-09 3.10E-09 1.96E-09 1.17E-09 7.73E-10 1.49E-09 2.28E-09 BASEIO Effective Filter A207 Shine Dose Multiplication Factors Gamma Energy Gamma
[RemAir per MeVIcc-sec I Range Energy IMeVJ I MeV at DP9 atDPlO atDPl1 atDP12 atDP13 atDP14 atDP15s atDP 16 0.0!sEs 0.1 0.1 1.1 IE-10 1.26E-10 730E-11 239E-11 4.2E-11 7.24E-1 2.96E-11.91E-tO 0.1 <Es 0.4 0.4 1.34E-09 1.21E-09 7.27E-10 3.37E-10 4.46E-10 7.08E-10 5.52E-10 1.55E-D9 0.4 <Es 0.9 0.8 1.74E-09 1.49E-09 9.18E-10 4.78E-10 5.73E-10 8.74E-10 8.04E-10 1.8E-09 0.9 E s 1.35 1.3 1.93E-09 1.60E-09
.OOE-09 5.61E-10 634E-10 9.42E-10 9.57E-10 1.89E-09 1.35 < E s 1.8 1.7 1.95E-09 1.59E-09 1.01E-09 5.80E-10 6.39E-10 9.392-10 9.94E-10 1.86E-09 1.8 < E s 2.2 2.18 1.97E-09 1.59E-09 1.OE-09 5.99E-10 6A5E-10 9.39E-O 1.03E-09 1.83E-09 2.2 <E s 2.6 2.5 1.96E-09 1.58E-09 1.01E-09 6.03E-10 6A3E-10 9.31E-10 1.04E-09 1.81E-09 2.6 < E s 3.0 2.8 1.93E-09 1.54E-09 9.88E-10 5.97E-10 6.31E-10 9.12E-10 1.03E-09 1.77E-09 3.0 < E s 5.0 4.0 1.80E-09 1.43E-09 9.20E-10 5.68E-10 5.89E-10 8A5-10 9.83E-10 1.62E-09 5.0<Es 10.0 6.2 1.64E-09 1.29E-09 8.35E-10 5.252-10 5.37E-10 7.64E-10 9.10E-10 1.46e-09
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Subject Fuel Handling Accident (FHAI Inside Containment - Control Room & Offsite Doses Sheet 47 of 241 REV ORIGINATOR DATE IRE DATE REVI ORIGINATOR DATE I IRE DATE l R 4A.Rustae T. Remick I
I 1
~~~~~~~
4.13 Average Fuel Rod Maximum Internal Pressures Design Analysis NFM-2-FP-0002 (Reference 6.1k) presents the Unit 2 Cycle 10 Fuel Performance Analysis. Per Design Analysis NFM-2-FP-0002 Section 4.12 and Table 4-15, the average rod maximum internal pressures as a function of void (rod gap) volume and plenum temperature occur for a Batch L/M Erbia rod axial burnup of 50.5 GWD/MTU:
Plenum Temperature Void Volume Rod Internal Pressure (with temporary gas release) 2 170.4 psia 613OF 70°F 0.8388 in3 1.0843 in3 829.4 psia
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Subject Fuel Handlina Accident FHA) Inside Containment - Control Room & OMsite Doses Sheet 48 of 241 REV ORIGINATOR DATE I
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L ATE[~
4 A. Rustaey lT.
Remick l
l v
l I
=
=
$~~~~~
5.0 METHODOLOGY 5.1 General Methodology Following a fuel handling accident inside containment (FHA-IC), activity from the damaged fuel rods is dispersed into the containment atmosphere, and from there, to the Control Room and the offsite Exclusion Area Boundary (EAB) and Low Population Zone (LPZ) locations.
The control room dose evaluation considers the following radiation sources:
Immersion and inhalation within the cloud inside the control room (Section 5.3)
Gamma radiation shine from the contaminated Containment Building air (Section 5.4.1)
Gamma radiation shine from the environmental cloud outside the Control Room Envelope (Section 5.4.2)
Gamma radiation shine from the emergency control room charcoal filter (Section 5.4.3)
The EAB and LPZ dose evaluations consider the following radiation sources:
nImmersion and inhalation within the environmental cloud at the Offsite locations (Section 5.3)
Gamma radiation shine from the contaminated Containment Building air (Section 5.4.1)
This calculation evaluates the more severe failure of either once burned fuel or high bumup fuel.
As discussed in Design Input 4.1, once burned fuel is modeled with an iodine fuel rod gap fraction of 10 percent, a single fuel pin radial peaking factor (RPF) of 1.71, and an assembly averaged RPF of 1.40. High burnup fuel is modeled with an iodine fuel rod gap fraction of 12 percent, a single fuel pin RPF of 1.71, and an assembly averaged RPF of 1.37. Section 8.1 determines which of these scenarios will yield the most severe dose consequences, and it is that case which is evaluated in the remainder of this calculation.
The FHA is evaluated with credit taken for a high radiation induced Control Room Isolation Signal (CRIS) which places the Control Room HVAC system into the high radiation isolation mode, and starts both trains of the control room essential HVAC system.
This calculation uses the LOCADOSE (Reference 6.6a) and SOURCE2 (Reference 6.6c) computer codes to evaluate the doses of interest. These computer codes are described in Section 5.2.
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Subject Fuel Handling Accident (FHAI Inside Containment - Control Room & Offsite Doses Sheet 49 of 241 ORIGINATOR DATE IRE DATE _ ORIGINATOR DATE DATE R
ElA Rtaey_
IT.
Remick V
I[I z
5.2 Computer Code Descriptions 5.2.1 LOCADOSE Computer Code Description.
The inhalation and immersion doses due to the fuel handling accident activity release mechanism will be evaluated with Bechtel Computer Program NE-3 19, "LOCADOSE". The LOCADOSE Code consists of three modules: an activity transport program, a dose calculation program, and a filter loading program. The first two modules are used in this calculation.
The activity transport module calculates activities, integrated activities, and releases over a time period using a multi-region model that can accommodate up to nine regions and fifty time steps.
Daughter isotope production can be performed by this program. Activities, time integrated activities and releases are saved on a file for use by the other modules.
The dose calculation module uses the file generated by the activity transport program, the isotope library file and a user-generated data file to calculate doses and dose rates. Doses and dose rates can be obtained for all the regions used by the activity transport program and for up to twenty offsite locations.
The LOCADOSE Code used in Revision 3 was executed on the Nuclear Fuel Management RISC 6000 workstation. Use of the LOCADOSE Code on the NFM RISC 6000 workstation has been verified and validated as detailed in a Software Installation Report (Reference 6.6b).
Four error notices have been issued against the LOCADOSE Code revision 3:
Bechtel Error Notice 93-1 (issued on July 30, 1993) relates to the modeling of containment spray operation. Since containment sprays are not modeled in this calculation, Bechtel Error Notice 93-1 does not affect this calculation.
SCE Error Number I (issued on August 22, 1997) relates to the formatting of the LOCADOSE Dose input file when requesting dose rate output. The concerns addressed in SCE Error Number I have been addressed by the proper formatting of the LOCADOSE dose input file.
SCE Error Number 2 (issued on November 9, 1998) addresses the same concern documented in SCE Error Number I (the error was addressed for a second time to provide closure to Bechtel LOCADOSE Code Error Notice 98-1). The concerns addressed in SCE Error Number 2 have been addressed by the proper formatting of the LOCADOSE dose input file.
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Iuicat Feil I-Innrilinn Arcident IFWA% Insid (nntainment - rnntrI Ptanm & CffcitP rln:ac
- hoct rn nf A1
-"J9 c
CNl~WW 9-WXs
'..~....l
- IVtFll;lWVxi iV REV ORIGINATOR DATE IRE DATE REV ORIGINATOR DATE IRE DATE R
4 A. Rustae T. Remick I
SCE Error Number 3 (issued on November 9, 1998) relates to an error present only in LOCADOSE Version 4.2. Since this calculation employs LOCADOSE Version 3.0 (see note below), no action is required.
l The LOCADOSE Code Release 6.0 was used in Revision 4 of this calculation and was executed I
on the Nuclear Fuel Management NT station (Device ID D088771). Use of the LOCADOSE l
Code on the NFM NT station has been verified and validated as detailed in a Software i
Installation Report (Reference 6.6g).
One error notice have been issued against the LOCADOSE Code Release 6.0:
I Error notice 20014 (issued on March 29, 2001) dealt with input file translation program l
TransLat.exe which converted releases 4 and 5 files to release 6. Since LOCADOSE
{
Release 6 was used in Rev. 4 of this calculation no conversion was necessary.
I Therefore, this error notice does not impact this calculation.
l Section 9.2 of this calculation presents the library, input and output files associated with the l
LOCADOSE Code analysis.
5.2.2 SOURCE2 Computer Code Description Calculation of Gamma radiation shine doses requires determination of gamma source strength spectra. Each gamma source strength spectrum will be evaluated with Bechtel Computer Program NE-602, "SOURCE2". The SOURCE2 Code determines the gamma source strength spectrum from a User specified energy grouping arrangement and a User specified activity profile. The activity profile represents the total number of curies of each isotope present in the radiation source.
To calculate control room gamma radiation shine dose rates in units of Rem/hr using the Dose Multiplication Factors presented in Design Inputs 4.10 through 4.12 requires input in the form of a gamma source strength spectrum in units of MeV/cc-sec. The SOURCE2 Code will calculate gamma source strcngth spectrum in units of MeV/cc-sec with input from LOCADOSE in the form of total Curies per isotope for the node of interest, and a Code Multiplier equivalent to one over the nodal volume (I/cc).
The SOURCE2 Code is executed on the Nuclear Fuel Management RISC 6000 workstation.
Use of the SOURCE2 Code on the NFM RISC 6000 workstation has been verified and validated as detailed in a Software Installation Report (Reference 6.6d).
E&TS DEPARTMENT CALCULATION SHEET PRELIM. CCN NO.
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No error notices have been issued against the SOURCE2 Code.
Section 9.3 of this calculation presents the input and output files associated with the SOURCE2 Code analysis.
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Siuhit-t Ft W Hnnrffinn Accideint (FWAI Insie Cnntainment - Cnntrnl Rnonm K tlffnitp Doses Sheet n f A1 as__~ ~
w g l lw M w
- Et*l-alB
__l1
-m wW
- . al By
--a A
s ax -,
vow =
REV ORIGINATOR DATE IRE DATE REV ORIGINATOR DATE IRE DATE R
4 A. Rusta T. Remick I
II V
5.3 EAB, LPZ and Inside Control Room Cloud Doses The immersion and inhalation doses within the environmental cloud at the Exclusion Area Boundary and at the outer boundary of the Low Population Zone, and the immersion and inhalation doses within the cloud inside the control room are evaluated with the LOCADOSE Code using the appropriate assumptions and design inputs from Sections 3 and 4.
The LOCADOSE Code model for the fuel handling accident inside containment (FHA-IC) sequence will be set up with four regions and two offsite dose receptors. As required by the LOCADOSE Code region numbering convention, Region 1 represents the outside environment, and the last region (Region 4) represents the control room. Offsite dose receptor I represents an individual at the EAB, and offsite dose receptor 2 represents an individual at the outer boundary of the LPZ. Additionally, a third offsite location is modeled to simulate control room HVAC intake.
Figures 5.3-1 through 5.3-3 show the LOCADOSE Code models used, including information on region volumes, flowrates, and filter removal of iodines (1), noble gases (NG), and particulates (P). The primary difference between the figures is the depiction of control room (CR) HVAC operation prior to and after control room isolation (as detailed in Design Input 4.4):
Figure 5.3-1 depicts CR HVAC operation prior to control room isolation, with the CR HVAC operating in normal mode.
Figure 5.3-2 depicts CR HVAC operation after control room isolation, with both trains of the CR HVAC operating in emergency mode during the first 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> of the FHA-IC event.
Figure 5.3-3 depicts CR HVAC operation after control room isolation, with one train of the CR HVAC operating in emergency mode after the first 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> of the FHA-IC event.
As will be discussed in Section 5.4.3, the control room charcoal filter shine dose is dependent on the CR HVAC intake filters' instantaneous activity loading at various times during the FHA event. Due to atmospheric dispersion, only a portion of the Containment activity release will become entrained in the Control Room HVAC intake flow. To analyze this dispersion and intake into the control room, Figures 5.3-1 through 5.3-3 show the LOCADOSE Code modeling a direct flowpath between the Containment Building (Region 2) and a Control Room HVAC Intake Filter (Region 3). Section 8.5 calculates an unfiltered flow rate of 264 cfm as equivalent to the portion of the Containment activity release which becomes entrained in the Control Room HVAC intake flow.
E&TS DEPARTMENT CALCULATION SHEET ICCN NO.I PRELIM. CCN NO.
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Siihiret Fl Hanilinn Acciient FHA) IniHe Cnntainment - Cnntrnl Rnnm A Offsite Onoe Sheept 53 f
OA1
_MV-w
- i sI- -I Iw - -
s~
.- *. *1
- -W I-----
-- 1.
ltl A ---
ii
~ V_
Vl REV ORIGINATOR DATE IRE DATE REV ORIGINATOR DATE IRE DATE R
4 A. Rustae T. Remick l
l E
I I 7
1 1
l ffi~~~~
Figure 5.3-1 LOCADOSE Model prior to CR Isolation CR HVAC In Normal Mode (Neither CR Isolation nor CREACUS Operation)
Region 2-Region 3 -
CONTAINMENT BLDG Q2.o3= A206 or A207 CR INTAKE FILT]
Vrplo7= 1.422E+6 ft3
= 0 cfm Vri.o 3 = 7.32 ft3 1
I t
UNFILTERED OUTFLOW Q2to 1= 82000 cfm PSEUDO RECIRCULATION FLOW (REMOVES GASES FROM FILTER)
Q3. 3 = e6 cfmn F: 0% 1, 100% NG, 0% Particulates I
I p-Region
- ENVIRONMENT V
=not modeled]
Atmospheric Dispersion (X/Q) to Control Room X/Q to I
EAB, LPZ UNFILTERED INFLOW FILTERED INFLOW Qunfii.nd flo = 5820 cfm &
Qfllerd inflow 0 cfm 6820 cfm (for increased CR inleakage)
F: 0% 1, 0% NG, 0% Particulates Region 4 - CONTROL ROOM v1 ezi0n4 = 2665920 ft3 I
I 11 Ii I
OUTFLOW Q2utflo = 5820 cfm &
6820 cfm (for increased CR inleakage)
FILTERED RECIRC. FLOW Qrccirc no%% - 29885 Cffil F: 0% I, 0% NG, 0% Particulates I
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Subject Fuel Handling Accident (FHAI Inside Containment - Control Room & Offsite Doses Sheet 54 of 241 A. RustT.RemkI I
= _ _ _L i i i
Figure 5.3-2 LOCADOSE Model after CR Isolation and prior to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 2 Trains of CR HVAC in Emergency Mode (by a High Radiation Induced CRIS)
Region 2-CONTAINMENT BLDG Qt3=
V~mi) ?= 1.422E+6 ft3 264 cfin I
UNFILTERED OUTFLOW Q2 to = 81736 cfm PSEUDO RECIRCULATION FLOW (REMOVES GASES FROM FILTER)
Q31. = e6 cfn F: 0% I, 100% NG, 0% Particulates t
I I
S.,
Region 1 - ENVIRONMENT Vm., = not modeled]
II Atmospheric Dispersion (X/Q) to Control Room X/Q to I
EAB, LPZ UNFILTERED INFLOW FILTERED INFLOW Qufilered innow = 1 cfm &
Qfilterw= 4400 cfin 2 trains]
1000 cfrn (for increased CR inleakage)
F: 95% le 95% I.g, 99% ;
0% NG, 99% Particulates Region 4 - CONTROL ROOM IVadmA=266 I
I ii 11 I
OUTFLOW QoutOlos = 4410 cfm &
5400 cfm (for increased CR inleakage)
FILTERED RECIRC. FLOW Qrecirc now = 59870 cm [2 trains]
F: 95% Ieem 95% 1oru. 99% paric.
0% NG, 99% Particulates I
I I
-p
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Subject Fuel Handling Accident (FHA) Inside Containment - Control Room & Offsite Doses Sheet 55 of 241 R
OG TDATE IRE DATE REV ORIGINATOR DATE IRE DATE R
I I
I I
I I~~~~~~~
Figure 5.3-3 LOCADOSE Model after 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 1 Train of CR HVAC in Emergency Mode (by a High Radiation Induced CRIS)
UNFILTERE Q2.0l =
I I
'I
- D OUTFLOW PSEUDO RECIRCULATION FLOW R1736 cfin (REMOVES GASES FROM F1LTER)
I I- _
Q3 to3 e6 efin F: 0% I, 100% NG, 0% Particulates Region 1 - ENVIRONMENT Vm,, = not modeled]
Atmospheric Dispersion (X/Q) to Control Room X/Q to I
EAB, LPZ UNFILTERED INFLOW FILTERED INFLOW Qunfiltemd inflow = 1 0 cfm &
Qfiftemdinflow = 2200 cfn [1 train]
1000 cfin (for increased CR inleakage)
F: 95% Ie 95% Io 99% Lc 0% NG, 99% Particulates Region 4 - CONTROL ROOM I -
4 = 266,920 ft3 I
I ii I
OUTFLOW Q~tflo
= 2210 cfm &
3200 cfm (for increased CR inleakage)
FILTERED RECIRC. FLOW Qrcirc flow = 29934 cfim [I train]
F: 95% lelem, 95% lorg. 99% lpanic.
0% NG, 99% Particulates t
I I
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Rus ae T. Remick v
_ev I.m...m..um... I 1
5.4 Gamma Radiation Shine Contributions to the Whole Body Dose In addition to thyroid inhalation, whole body gamma cloud immersion, and beta skin cloud immersion doses determined with the LOCADOSE Code, this calculation considers contributions to the whole body gamma dose from containment shine, environmental cloud shine, and control room charcoal filter shine. The general method used to calculate these shine doses is:
Determine shine instantaneous gamma source strengths as a function of time with the SOURCE2 Code using shine source activity loading and applicable dilution volume as code input parameters.
Determine shine dose rate as a function of time by scaling the shine instantaneous gamma source strength with the Dose Multiplication Factors (DMFs) presented in Design Inputs 4.10 through 4.12.
Determine the accident duration whole body gamma shine dose by integrating the shine dose rate profile and considering applicable occupancy factors.
5.4.1 Containment Shine Dose Rates The contaminated air inside the Containment Building is a gamma radiation shine source for the Control Room and offsite EAB and LPZ dose receptors.
The containment shine dose at distant locations such as the EAB and LPZ are typically much less than the offsite immersion doses due to the fuel handling accident inside containment (FHA-IC) activity release. This is confirmed by a comparison of the post-LOCA containment shine and offsite immersion doses documented in Calculation N-4060-008 (Reference 6. Ig, Sections 8.7 and 8.9). These post-LOCA doses show that the containment shine dose is at least two orders of magnitude (a factor of 100) smaller than the offsite immersion doses. If the FHA-IC immersion doses calculated at the EAB and LPZ dose receptors are small (e.g., less than I rem), then the direct gamma radiation shine dose from the containment to the EAB and LPZ will be negligible (i.e., < 0.01 rem), and it will not be necessary to calculate the containment shine dose at the offsite locations.
The SOURCE2 Code will be used to determine the containment cloud instantaneous gamma source strengths (MeV/cc-sec), which will be multiplied by the containment shine DMFs from Design Input 4.10 to calculate the containment shine contribution to the Control Room Operator whole body gamma dose rates. The dose rates will be time integrated using the Section 5.4.4
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T. Remick I
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l~~~~~~~~~~~~~~~
methodology to determine the accident duration containment shine dose. Input to the SOURCE2 Code consists of the BASE10 gamma energy structure (per Design Input 4.9), the containment cloud's instantaneous activity loading (ci) at various times during the FHA-IC event (based on the LOCADOSE transport activity program), and a SOURCE2 Code Multiplier representing one over the containment air dilution volume (cc-'). Per Design Input 4.3.1, the containment air dilution volume is 1.422e6 ft3 with a corresponding inverse volume of 2.483e-1I cm 3.
The containment air dilution volume of 1.422e6 ft3 does not credit dilution within the dome space. However, the containment shine DMFs were calculated using a model in which the contaminated containment air source fills the containment, including the dome space. As such, use of the containment shine DMFs with this calculation's FHA-IC source model conservatively overpredicts the Control Room Operator whole body gamma dose rates (and dose) due to containment shine.
5.4.2 Environmental Cloud Shine Dose Rates The contaminated air in the environmental cloud outside of the control room envelope is a gamma radiation shine source for the Control Room dose receptor (the environmental cloud dose contributions to the offsite EAB and LPZ dose receptors are considered in the offsite immersion dose as evaluated with the LOCADOSE Code).
To calculate the environmental shine contributions to the Control Room Operator whole body dose, the source term in the environmental cloud must be calculated. This is performed using the following equation which scales the containment cloud source term by the containment release flow rate (thereby determining the source term release rate to the environment), and then scaling by the control room atmospheric dispersion factor (thereby determining the environmental cloud source term at the control room HVAC intake):
Se(t) = St) x F x X ()
Q whcre:
S,(t) environmental cloud source term in MeV/cc-sec S,(t) containment cloud source term in MeV/cc-sec [calculated per Section 5.4.1]
F containment release flow rate in m3/sec [per Design Input 4.3.2]
x/Q(t) control room atmospheric dispersion factor in sec/m3 [per Design Input 4.7]
The environmental cloud source term is then multiplied by the environmental cloud shine DMFs from Design Input 4.11 to generate Control Room Operator dose rates. The dose rates will be
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_ T
_I_
-E 4
A. Rustaev IT. Remick i
v XinmtW ii time integrated using the Section 5.4.4 methodology to determine the accident duration environmental cloud shine dose.
5.4.3 Control Room Charcoal Filter Shine Dose Rates Per Design Input 4.4.2, in the high radiation isolation mode the makeup air entering the control room passes through filters SAI 510MA206 and SA15 10MA207 (A206 and A207) of the Emergency Ventilation System train, and then through filters SA1510ME418 and SA1510ME419 (E418 and E419) of the Emergency Air Conditioner filter train. Per Assumption 3.12, for the purpose of determining the shine dose contribution from the control room HVAC filters, CR HVAC intake filters A206 and A207 are assumed to be 100% efficient at removing iodine and particulates from the incoming air. This maximizes the amount of iodine and particulates retained on A206 and A207, and thus maximizes the shine dose from these two filters. In reality, iodine and particulates that are not trapped on the intake filters will eventually be trapped on recirculation filters E418 and E419, which are located in the vicinity of the intake filters per Drawing 40002 Sheet 1 (Reference 6.2m). However, the results of the filter shine dose calculation would not be significantly different, since the geometry of the direct shine pathways from E418 and E419 is similar to the geometry of the direct shine pathways from A206 and A207.
The SOURCE2 Code will be used to determine the control room HVAC intake filters instantaneous gamma source strengths (MeV/cc-sec), which will be multiplied by the control room filter shine DMFs from Design Input 4.12 to calculate the filter shine contribution to the Control Room Operator whole body gamma dose rates. The dose rates will be time integrated using the Section 5.4.4 methodology to determine the accident duration filter shine dose. Input to the SOURCE2 Code consists of the BASE 10 gamma energy structure (per Design Input 4.9),
the CR HVAC intake filters' instantaneous activity loading (ci) at various times during the FHA-IC event (based on the LOCADOSE transport activity program), and a SOURCE2 Code Multiplier representing one over the CR HVAC intake filter volume (cc-'). Per Design Input 4.4.6, the volume of each CR HVAC intake filter is 7.32 3 with a corresponding inverse volume of 4.824e-6 cm 3.
The LOCADOSE Code computer run will be used to determine the time dependent filter activity loading (ci) entered into the SOURCE2 Code. Due to atmospheric dispersion, only a portion of the Containment activity release will become entrained in the Control Room HVAC intake flow.
To analyze this dispersion and intake into the control room, the LOCADOSE Code will model a direct flowpath between the Containment Building (Region 2) and the Control Room HVAC intake filter (Region 3). The Region 2 to 3 flowrate will account for the atmospheric dispersion of the Containment activity release, and the actual intake flow rate into the CR HVAC intake
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I filter. Although atmospheric dispersion between the Containment and control room varies with time, per Design Input 4.3.2 virtually all of the activity is released from the containment during the first 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> of the event. Therefore, a single Region 2 to 3 flowrate determined by using the 0-8 hour Control Room atmospheric dispersion factor from Design Input 4.7 will be applied for the 30 day analysis duration. Section 8.5 calculates an unfiltered flow rate equivalent to the portion of the Containment activity release which becomes entrained in the Control Room HVAC intake flow.
Since a Control Room HVAC intake filter (Region 3) does not retain noble gases, the LOCADOSE Code will model a "Region 3 recirculation filter" (with an arbitrary flowrate of 1,000,000 cfm and a noble gas filter efficiency of 100 percent) to remove the noble gases that enter Region 3. To maximize the filter shine dose, the LOCADOSE Code will not model a "Region 3 exhaust flow"; this effectively retains 00 percent (less radioactive decay) of the iodine and particulates entering Region 3. Although a large filtered flowrate has been modeled in an effort to prevent the LOCADOSE Code from showing an accumulation of noble gases on the Control Room HVAC intake filter, a review of the LOCADOSE Code output may show the presence of some trace noble gas activity in Region 3 at the end of each time step. Since noble gas activity will not be collected on the CR HVAC intake filter, the SOURCE2 Code input file used to determine instantaneous gamma source strength spectra for the filter shine analysis will zero out the CR HVAC intake filter noble gas activity values.
5.4.4 Determination of 30-Day Time Integrated Doses The time integrated doses from the shine gamma sources is determined by integrating the instantaneous dose rates determined using the methodology of Sections 5.4.1 through 5.4.3 over the 30 day event duration. The integration method is based on the premise that the activity (and hence the dose rate) in any region between any two times (t, and t2) will behave as an exponential of the form:
DR(t) = DRe 4'-")
(1)
Therefore, at time t the dose rate is given by:
DR2
= DR,e -Mi 2 -'.)
(2)
The effective decay constant X for the t to t time interval can be calculated from the above equation to be:
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= - n(DR,/DR,)
(2 -ti)
(3)
Integrating the instantaneous dose rate DR(t) between times t and t, and applying the Control Room Occupancy Factor (CROF) from Design Input 4.6 yields:
D = fDR(t)d = CROF x fDRe
'")dt it
= - CROF DR, le '#'l
= - CROF x DR, e
= - CROF x 1 DRe
- DR, I 1.
(4)
(5)
(6)
(7)
Inserting equation 2 and then equation 3 into equation 7 yields the following time integrated dose for the t, to t2 time interval:
D = DR(t)dt = CROF x (DR, -DR,) (12 -1 ln(DR2/DR,)
(8)
NOTE 1:
The natural logarithm function in the equation 8 denominator will yield inaccurate integrated dose results under certain conditions. To eliminate this concern, the following limitations are placed on the use of equation 8:
l)if DR, =DR2 then D = (DR, + DR2)(t2 - t) / 2 I
- 2) if DR = 0, and/or DR2 0 O then D = (DR + DR2)(t 2 -t) / 2 (i.e., a trapezoidal approximation is employed to estimate the integrated dose)
I NOTE 2:
Due to inclusion of the Control Room Occupancy Factor in this derivation, the actual 30 day shine dose to equipment in the Control Room is higher than that calculated with Equation 8. The Section 10 Quattro-Pro spreadsheets determine 30 day shine doses with and without consideration of the CROF.
E&TS DEPARTMENT PI.CCN NC CALCULATION SHEET RELIMCCNN(
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On1.
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- 1 REV ORIGINATOR DATE IRE DATE REV ORIGINATOR DATE IRE DATE R
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6.0 REFERENCES
6.1 Calculations 6.1a Units 2&3 Calculation C-257-01.06.01, Revision 2 (including CCN 3), "Containment Shell Analysis - Containment Passive Heat Sink" 6.1b Units 2&3 Calculation M-0073-041, Revision 8 (CCN 20), "Auxiliary Building -
Control Area El 30' 0", Heat Load and Equipment Sizing Normal and Emergency" 6.1c Units 2&3 Calculation M-0073-095, Revision 3 (including CCNs 1 through 4),
"Infiltration into the Control Room Envelope from Surrounding Areas" 6.1d Units 2&3 Calculation N-4010-001, Revision 5 (including CCNs 2 and 3), "Control Room x/Qs" 6.le Units 2&3 Calculation N-4010-002, Revision 1, "EAB and LPZ X/Q values" 6.1f Units 2&3 Calculation N-0450-002, Revision 1, "Control Room Doses to Personnel" 6.Ig Units 2&3 Calculation N-4060-008, Revision 5 (including CCNs 3 and 4), "Post-LOCA Containment Leakage Doses to EAB, LPZ and Control Room" 6.1h Units 2&3 Calculation N-4060-016, Revision 2, "Post-LOCA Charcoal Filter Doses -
Control Building" 6.1i Units 2&3 Calculation M-0022-008, Revision 2 (including CCNs 1 through 4), "Spent Fuel Pool Heat Exchanger Performance" 6.Ij Units 2&3 Calculation J-SPA-179, Revision 0 (including CCNs 2 through 4), "Control Room Fuel Handling Building Monitor Set Points" 6.1k Design Analysis NFM-2-FP-0002, Revision 00, "SONGS 2 Cycle 10 Fuel Performance Analysis", dated October 8, 1998
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~
6.1!
Design Analysis NFM-2-PH-0003, Revision 00, "SONGS-2 Cycle 10 CORD/ROCS/MC Design Models, Depletions, Rodded Cases, and Integrated Files", dated March 27, 1998 6.1m Design Analysis NFM-3-PH-0003, Revision 00, "SONGS-3 Cycle 10 CORD/ROCS Design Models, Depletions, Rodded Cases, and Integrated Files", dated November 16, 1998 6.1n DELETED 6.1o DELETED 6.1p Design Analysis NFM-2-PH-0004, Revision 00, "SONGS-2 Cycle 10 Design Parameters and F. vs Power", dated April 23, 1998 6.lq Design Analysis NFM-3-PH-0004, Revision 02, "SONGS-3 Cycle 10 Design Parameters and F. vs Power", dated February 11, 1999 6.Ir Design Analysis NFM-2-PH-0005, Revision 00, "SONGS-2 Cycle 10 Physics Input to LOCA, TORC, and FATES Analysis and Pin Census for Pre-Trip Steam Line Break",
dated June 22, 1998 6.1s Design Analysis NFM-2-PH-0016, Revision 01, "SONGS 2 Cycle 10 LOCA and Non-LOCA Fission Product Source Term', dated October 7, 1998 6.1t Design Analysis NFM-3-PH-0016, Revision 00, "SONGS 3 Cycle 10 LOCA and Non-LOCA Fission Product Source Term", dated December 17, 1998 6.1u ABB Calculation A-SCE-FMDE-002, Revision 02, "Fuel Bundle Drop Evaluation for SCE Units 2 and 3", dated December 3, 1998 Note :This Calculation is in CDM as Document S023-990-C299 Revision 2 Notc 2:Calculation A-SCE-FMDE-002 Revision 02 is Enclosure 2 in ABB letter S-98-209, "Transmittal of Fuel Bundle Drop Evaluation for SONGS Units 2 and 3 (Calculation A-SCE-FMDE-002, Rev. 02)", dated December 7, 1998.
6.1 v Design Analysis NFM-2/3-PH-1116, Revision 00, "SONGS 2&3 Cycle 1 ILOCA and Non-LOCA Fission Product Source Term", dated 3/9/2000.
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=4 6.2 Drawings 6.2a Units 2&3 Drawing 40173A, Revision 20, P&I Diagram, Control Room Complex HVAC (Normal A.C.) - System No. 1510 6.2b Unit 3 Drawing 40173AS03, Revision 5, P&I Diagram, Control Room Complex HVAC (Computer Room) - System No. 1510 6.2c Units 2&3 Drawing 40173C, Revision 21, P&I Diagram, Control Room Complex HVAC (Emergency V.S. and A.C. Units) - System No. 15 10 6.2d Units 2&3 Drawing 23060, Revision 6, Containment Structure - Dome Liner & Inserts -
Sections & Details Sheet I 6.2e DELETED 6.2f DELETED 6.2g DELETED 6.2h DELETED 6.2i Units 2&3 Air Flow Diagram 40096, Revision 17, Train B - Control Building - El 30'-0" 6.2j Units 2&3 Air Flow Diagram 40098, Revision 9, Train A - Control Building - El 30'-O" 6.2k Units 2&3 Drawing 31394, Revision 19, Elementary Diagram - HVAC Plant Control Room Isolation System Train A 6.21 Unit 2 (and 3, although not specified) Drawing 31395, Revision 19, Elementary Diagram
- HVAC Plant Control Room Isolation System Channel B 6.2m Units 2&3 Drawing 40002 Sheet 1, Revision 33, General Arrangement - Unit 2 - Plan at El. 45'-0" to 30'-0" 6.2n Units 2&3 Drawing S023-410-1-1, Revision 13 (including DCN 2), Filter House Auxiliary Building Control Room Emergency HV Unit
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A. Rute
.Rmc 6.3 Correspondence 6.3a ABB Memorandum ST-98-427, "Maximum Internal Gas Pressure for Dose Calculations from Postulated Fuel Handling Accidents", dated August 10, 1998.
Note: Memorandum ST-98-427 is Enclosure 3 in ABB letter S-98-139, "Open Porosity Model, Axial Densification Factor, and Maximum Internal Gas Pressure -
SONGS Unit 2, Cycle 10 Support", dated September 3, 1998 6.3b E-Mail from D. Higgins (SCE) to F. Santa Ana (SCE) et. al., dated 5/13/93, "CR HVAC Intake Filter Housing" (a copy is included in Section 1 1. 1).
6.3c E-Mail from F. Santa Ana (SCE) to T. Remick (Bechtel), dated 9/16/93, "Control Room HVAC Infonnation" (a copy is included in Section 11.2).
6.3d DELETED
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.1 1-1 1
1
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=
4 6.4 Regulatory Documents 6.4a 10 CFR Part 50, "Domestic Licensing of Production & Utilization Facilities" 6.4b 10 CFR Part 100, "Reactor Site Criteria" 6.4c San Onofre Unit 2 Technical Specifications, i)
Section 1.1 page 1.1-3 [Amendment 145]
ii)
LCO 3.3.8 page 3.3-35 [Amendment 132]
iii)
LCO 3.6.6.1 page 3.6-18 [Amendment 127]
iv)
LCO 3.6.6.2 page 3.6-21 [Amendment 127]
v)
LCO 3.6.8 page 3.6-25 [Amendment 127]
vi)
LCO 3.9.6 page 3.9-10 [Amendment 134]
vii)
Section 5.5.2.12 page 5.0-19c [Amendment 140]
viii)
LCO 3.3.9 page 3.3-9 [Amendment 132]
6.4d San Onofre Unit 3 Technical Specifications, i)
Section 1.1 page 1.1-3 [Amendment 137]
ii)
LCO 3.3.8 page 3.3-35 [Amendment 121]
iii)
LCO 3.6.6.1 page 3.6-18 [Amendment 116]
iv)
LCO 3.6.6.2 page 3.6-21 (Amendment 116]
v)
LCO 3.6.8 page 3.6-25 [Amendment 116]
vi)
LCO 3.9.6 page 3.9-10 [Amendment 123]
vii)
Section 5.5.2.12 page 5.0-19c [Amendment 132]
viii)
LCO 3.3.9 page 3.3-9 [Amendment 121]
6.4e San Onofre Unit 2 Licensee Controlled Specifications, i)
LCS 3.9.101 pages 3.9-101-1 and 3.9-101-2 [Rev. 1, May25, 20001 ii)
LCS 3.9.102 pages 3.9-102-1 and 3.9-102-2 [Rev. 2, May 2000]
6.4f San Onofre Unit 3 Licensee Controlled Specifications, i)
LCS 3.9.101 pages 3.9-101-1 and 3.9-101-2 [Rev. 1, May 25, 2000]
ii)
LCS 3.9.102 pages 3.9-102-1 and 3.9-102-2 [Rev. 2, May 2000) 6.4g SONGS 2&3 Updated Final Safety Analysis Report, up to and including Amendment 1 5A 6.4h Regulatory Guide 1.4, Revision 2, "Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss of Coolant Accident for Pressurized Water Reactors", dated June 1974 I
I
E&TS DEPARTMENT CALCULATION SHEET ICCN NO.J l
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I E
6.4i Regulatory Guide 1.25 (Safety Guide 25), "Assumptions Used for Evaluating the Potential Radiological Consequences of a Fuel Handling Accident in the Fuel Handling and Storage Facility for Boiling and Pressurized Water Reactors", dated March 23, 1972 6.4j Regulatory Guide 1.52, Revision 2, "Design, Testing and Maintenance Criteria for Post Accident Engineered-Safety-Feature Atmosphere Cleanup System Air Filtration and Adsorption Units of Light-Water-Cooled Nuclear Power Plants", dated March 1978 6.4k NUREG/CR-5009, "Assessment of the Use of Extended Burnup Fuel in Light Water Power Reactors", published February 1988 6.41 Standard Review Plan Section 6.4, "Control Room Habitability System" i)
NUREG 75/087, Revision 1, dated December 1978 ii)
NUREG-0800, Revision 2, dated July 1981 6.4m Standard Review Plan Section 15.7.4, "Radiological Consequences of Fuel Handling Accidents" i)
NUREG 75/087, Revision 0, dated November 24, 1975 ii)
NUREG-0800, Revision 1, dated July 1981 6.4n NUREG-0017, Revision I, "Calculation of Releases of Radioactive Materials in Gaseous and Liquid Effluents from Pressurized Water Reactors (PWR-GALE Code)",
published April 1985 6.4o Safety Evaluation Report Related to Unit 2 License NPF-10 Amendment 132 and Unit 3 License NPF-15 Amendment 121; attached to October 8, 1996 Letter from Mel B.
Fields of the NRC to Harold B. Ray of SCE 6.4p Branch Technical Position ASB 9-2, Revision 2,"Residual Decay Energy for Light-Water Reactors for Long-Term Cooling", dated July 1981 Note: This Branch Technical Position is attached to NUREG-0800 Standard Review Plan Section 9.2.5, "Ultimate Heat Sink".
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6.5 Procedures 6.5a Maintenance Procedure SO23-I-2.44, Revision 6 (TCN 6-3), CREACUS - Control Room Emergency Air Clean Up System Operation and Operability Test Surveillance.
6.5b DELETED 6.5c Abnormal Operating Instruction S023-13-20, Revision 6, "Refueling Accidents" 6.6 Other Documents 6.6a LOCADOSE Code, Bechtel Standard Computer Program NE-3 19, Release 3, dated October 19, 1990. Computer Program Error Notice Numbers 1, 2 and 3 have been issued against this code.??
6.6b Software Installation Report, LOCADOSE (NE-319) Version 3.0, RISC 6000 Computer System - Device ID D037571, Operating System IBM AIX Version 4.2.1, Revision I Approved August 31, 1998. ??
6.6c SOURCE2 Code, Bechtel Standard Computer Program NE-602, Release D2-5, dated December 11, 1991. No Computer Program Error Notices have been issued against this code.
6.6d Software Installation Report, SOURCE2 (NE-602) Version D2-5, RISC 6000 Computer System - Device ID D037571, Operating System IBM AIX Version 4.2.1, Revision 2 Approved August 31, 1998.
6.6e DELETED 6.6f DELETED 6.6g Software Installation Report, LOCADOSE (NE-319) Release 6.0, Pentium I, 800 MHz Computer System - Device ID D088771, Operating System NT Version 4.00, Revision 2 Approved May 15, 2001.
6.6h Samuel Glasstone, Principles of Nuclear Reactor En neenng. published by D. Van Nostrand Company, Inc. of Princeton, New Jersey I
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E&TS DEPARTMENT tCCN NOJ CALCULATION SHEET PRELIM. CCN N(
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6.6j DE1ETED 6.6k Herman Cember, Introduction to Health Physics, published by Pergamon Press of New York, dated 1989.
6.61 ICRP Publication 30, "Limits for Intakes of Radionuclides by Workers", Supplement to Part 1, International Commission on Radiological Protection, adopted July 1978.
6.6m Frank Kreith, Principles of Heat Transfer. 2nd Edition, published by International Textbook of Scranton, Pennsylvania I
E&TS DEPARTMENT KX;N NUJ CALCULATION SHEET IPRELIM. CCN N1 vroject or DCP/FCN/ECP Caic. No. N-4072-003 cc Sc Subject Fuel Handling Accident (FHA) Inside Containment - Control Room & Offsite Doses Sheet 69 of 241 REV ORIGINATOR DATE IRE DATE REV ORIGINATOR DATE IRE DATE R
' U _ _ _ _ _ _ _ _ _ _ __
E 4 A. Rusae T. Remick v
S L.in m m i r n i n e j 7.0 NOMENCLATURE 7.1 Acronyms ABB-CE AR CCN CE CFR CPIS CR CREACUS CRG CRIS CROF DCF DCN DCP DMF EAB EAC ESFAS EVS FHA FHA-IC HEPA HVAC ICCN ICRP LCO LCS LOCA LPZ NRC RPD RPF RPS SIAS SRP UFSAR Asea Brown Boveri - Combustion Engineering Action Request Calculation Change Notice Combustion Engineering Code of Federal Regulations Containment Purge Isolation Signal Control Room Control Room Emergency Air Cleanup System Cable Riser Gallery (adjacent to the control room)
Control Room Isolation Signal Control Room Occupancy Factor Dose Conversion Factor Document Change Notice Design Change Package Dose Multiplication Factor Exclusion Area Boundary Emergency Air Conditioning Engineered Safety Feature Actuation System Emergency Ventilation Supply Fuel Handling Accident Fuel Handling Accident Inside Containment High Efficiency Particulate Air (filter)
Heating, Ventilating and Air Conditioning Interim Calculation Change Notice International Commission on Radiological Protection Technical Specification Limiting Condition for Operation Licensee Controlled Specification Loss of Coolant Accident Low Population Zone Nuclear Regulatory Commission Relative Power Density Radial Peaking Factor Reactor Protection System Safety Injection Actuation Signal Standard Review Plan (NUREG-75/087 or NUREG-0800)
Updated Final Safety Analysis Report I
I
E&TS DEPARTMENT CALCULATION SHEET ICCN NO.J PRELIM. CCN NO.
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}
lv 7.2 Symbols cP Gr h
k Nu Pr Q
q' T
V px/Q specific heat Grashof number heat transfer coefficient thermal conductivity Nusselt number Prandtl number volumetric source strength linear heat rate temperature thermal coefficient of volume expansion beta radiation gamma radiation viscosity density chi over Q, Atmospheric Dispersion Factor
E&TS DEPARTMENT CALCULATION SHEET Project or DCP/FCN/ECP Caic. No. N-4072-003 Sihipt-t Fel Hnndlino Arident FHA Inside Containment - Control Room & OffqitA Doses Sheet 71 of 241 REV ORIGINATOR DATE IRE DATE REV ORIGINATOR DATE IRE DATE R
4 A. Rustae T. Remick 7.3 Units Bq Btu cc cfin ci OR Ci cpm ft GWD/MTU hr in Ibm m
MeV min MWth psia psig sec Sv OF OR 1ci Becquerel British thermal unit cubic centimeters cubic feet per minute curies counts per minute (by a radiation monitor) feet Gigawatt-days per metric tonne of uranium hours inches pound-mass meters Million electron volts (mega-electron volts) minutes Megawatt-thermal pounds of force per square inch, absolute pounds of force per square inch, gauge seconds Sievert degrees Fahrenheit degrees Rankine microcuries (le-6 curies)
I
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-1 U--.4Z.
A.n ICU A l..,A,.-ci P
ntin.vrl
_ Ct^
trnel DoII C kffcit n
c Q
t 7
-f 9)A4
%JUUJtz1t UgI 101 IUIII 1 F-I.AAUCllLI C
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1 REV ORIGINATOR DATE IRE DATE REV ORIGINATOR DATE IRE DATE R
4 A. Rustaey T. Remick l
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'U.
8.0 COMPUTATIONS 8.1 Fuel Rod Gap Inventories Per Design Input 4.1.3, in the event of a fuel handling accident inside containment (FHA-IC) a total of 226 fuel rods will fail, representing 16 fuel rods in the dropped fuel bundle, and 210 fuel rods in one or two impacted fuel bundles. The dropped and impacted fuel assemblies may contain either once burned fuel or high bumup fuel. Per Design Input 4. 1, once burned fuel has been irradiated for a single fuel cycle, and high burnup fuel has been irradiated for more than one fuel cycle. This section calculates the gap inventory for both a once burned and a high bumup peak fuel rod when either 16 or 210 fuel rod failures are considered. These inventories are compared to determine which inventory set will yield the most severe dose consequences, and as such, which inventory set will be evaluated in the remainder of this calculation. The gap inventory of a representative failed fuel rod is then calculated by summing the products of the number of fuel rod failures and the peak fuel rod gap inventories for both the 16 and 210 fuel rod failures, and dividing by 226, the total number of failed fuel rods.
Design Input 4.1.1, Table 4.1 -1, and the power uprate correction factor of 1.02 present the fission product isotope inventory for a fuel rod. In Tables 8.1-1 through 8.1-4 this inventory is scaled by the various combinations of iodine and noble gas gap release fractions and radial peaking factors to determine the once burned and high burnup peak fuel rod gap inventories when either 16 or 210 fuel rod failures are considered.
Tables 8.1-1 through 8.1-4 do not address Iodine-134 or Xenon-138, in that Design Input 4.1 documents that these isotopes are not present when the fuel handling accident inside containment (FHA-IC) occurs 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> following reactor shutdown.
Per Design Input 4.1.2, consistent with Regulatory Guide 1.25, the gap activity in the damaged once-burned fuel rods consists of 10 percent of the total noble gases other than Krypton-85, 30 percent of the Krypton-85, and 10 percent of the radioactive iodine in the rods at the time of the accident. In addition, Design Input 4.1.2, notes that at high fuel discharge burnup, all but the iodine isotopes are calculated to have fuel rod gap inventories less than that recommended by Regulatory Guide 1.25 for use in fuel handling accident analyses; at high burnups the fuel rod gap iodine inventory increases to 12 percent, which is greater than the 10 percent value assumed in Regulatory Guide 1.25.
Per Design Input 4.1.4.1 it is conservative for this calculation to model an RPF of 1.71 when 16 fuel rod failures are considered. This RPF will be modeled for both once burned fuel rods and for high bumup fuel rods.
E&TS DEPARTMENT CALCULATION SHEET ICCN NOI PRELIM. CCN NO.
PAGE _
OF _
hu ject or DCP/FCN/ECP Calc. No. N4072-003 l CCN CONVERSION:
I CCN NO. CCN -
Subject Fuel Handling Accident (FHA) Inside Containment - Control Room & Offsite Doses Sheet 73 of 241 REV ORIGINATOR DATE IRE DATE REV ORIGINATOR I I I
A. Ruste T. Remick iv
{~~~~~
Per Design Input 4.1.4.2 when 60 or more fuel rods in an assembly are failed, the average radial peaking factor of the failed rods is equivalent to the failed assembly's Relative Power Density (RPD). Per Design Input 4.1.4.2, it is conservative for this calculation to model an RPD of 1.37 when 210 once burned fuel rod failures are considered, and to model an RPD of 1.37 when 210 high burnup fuel rod failures are considered.
A comparison shows that the Table 8.1-2 high burnup peak fuel rod gap inventory bounds the Table 8. 1-1 once burned peak fuel rod gap inventory when 16 fuel rod failures are considered.
This occurs because the iodine gap release fraction is greater in the high bumup case than in the once burned case. For this reason, this calculation will only evaluate the case of damage to high burnup fuel when 16 fuel rod failures are considered.
A comparison shows that the Table 8.1-4 high burnup peak fuel rod gap inventory bounds the Table 8.1-3 once burned peak fuel rod gap inventory when 210 fuel rod failures are considered.
This occurs because the iodine gap release fraction is greater in the high burnup case than in the once burned case. For this reason, this calculation will only evaluate the case of damage to high burnup fuel when 210 fuel rod failures are considered.
Table 8.1-5 determines the gap inventory of a representative failed fuel rod in the event of an FHA-IC. This inventory is calculated by summing the products of the number of fuel rod failures and the peak fuel rod gap inventories for both the 16 and 210 fuel rod failures, and dividing by 226, the total number of failed fuel rods.
E&TS DEPARTMENT CALCULATION SHEET ICCN NO.J PRELIM. CCN NO.
PAGE _
OF _
Project or DCP/FCN/ECP CaIc. No. N-4072-003 CCN CONVERSION:
CCN NO. CCN --
Subject Fuel Handling Accident (FHA) Inside Containment - Control Room & Offsite Doses Sheet 74 of 241 Table 8.1-1 Once Burned Peak Fuel Rod Gap Inventory Considering 16 Fuel Rod Failures Isotope la' Average Fuel Rod Power Uprate Once Burned Once Burned Once Burned Inventory Correction Gap Release Radial Peaking Peak Fuel Rod 72 hrs decay Factor Fraction Factor Gap Inventory (b) 16 Fuel Rod Failures 16 Fuel Rod
[per Table 4.1 -
(per Dl 4.1 I]
[per D 4.1.2]
[per Dl 4.1.4.1]
Failures (curies/rod)
(unitless)
(unitless)
(unitless)
(curies/rod)
Iodine-129 I.18e-04 1.02 0.10 1.71 2.07e-05 lodine-130 3.09e-02 1.02 0.10 1.71 5.41e-03 lodine-131 1.40e+03 1.02 0.10 1.71 2.45e+02 lodine-132 8.53e-07 1.02 0.10 1.71 1.49e-07 lodine-133 3.25e+02 1.02 0.10 1.71 5.69e+01 lodine-135 1.79e+00 1.02 0.10 1.71 3.13e-01 Krypton-83m 3.68e-10 1.02 0.10 1.71 6.44e-1 I Krypton-85 2.58e+01 1.02 0.30 1.71 1.35e+01 Krypton-85m 8.15e-03 1.02 0.10 1.71 1.43e-03 Krypton-87 1.02e-14 1.02 0.10 1.71 1.79e-15 Kryplon-88 4.21 e-05 1.02 0.10 1.71 7.37e-06 Xenon-131m 1.91 e+01 1.02 0.10 1.71 3.34e+00 Xenon-133m 6.42e+01 1.02 0.10 1.71 1.12e+01 Xenon-133 2.81 e+03 1.02 0.10 1.71 4.92e+02 Xenon-135m 2.74e-01 1.02 0.10 1.71 4.80e-02 Xenon-135 3.69e+01 1.02 0.10 1.71 6.46e+00 Rubidium-88 0.00e+00 1.02 0.00 1.71 0.00e+00 Cesium-135 0.00e+00 1.02 0.00 1.71 0.00e+00 (a)
The Rb-88 and Cs-135 daughter isotopes that the LOCADOSE Code library file created are each listed with zero initial (72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />) and adjusted gap activities.
(b)
Sample Calculation for 1 133 Gap Inventory, (325 ci/rod)
(1.02) x ( I n gap fraction) x(1.71
)F) = 56.9 ci/rod "Io$
I I
I I
I I
I I
I I
I I
I I
I I
I I
I I
I I
I I
I I
I I
I I
E&TS DEPARTMENT CALCULATION SHEET ICCN NOJ I
PRELIM. CCN NO.
PAGE OF _
i-roject or DCPIFCN/ECP Calc. No. N4072-003 CCN CONVERSION:
I CCN NO. CCN -
Subject Fuel Handling Accident (FHA) Inside Containment - Control Room & Offsite Doses Sheet 75 of 241 REV ORIGINATOR DATE IRE DATE REV ORIGINATOR DATE IRE DATE l R ZL-I I
4 A. R="tey T. Remick v
~~I_
. =
=
{~~~~~~~~~~~~
Table 8.1-2 High Burnup Peak Fuel Rod Gap Inventory Considering 16 Fuel Rod Failures Isotope Average Fuel Rod Power Uprate High Burnup High Burnup High Burnup nventory Correction Gap Release Radial Peaking Peak Fuel Rod 72 hrs decay Factor Fraction Factor Gap inventory "b 16 Fuel Rod Failures 16 Fuel Rod
[perTable4.1-1]
[perDl4.1.1]
[per DI 4.1.21
[per D1 4.1.4.1J Failures (curiestrod)
(unidess)
(unitless)
(unitless)
(curies/rod)
Iodine-129 l.l1
_e-04 1.02 0.12 1.71 2.48e-05 lodine-130 3.09e-02 1.02 0.12 1.71 6.49e-03 Iodine-131 l
.40e+03 1.02 0.12 1.71 2.94e+02 Iodine-132 8.53e-07 1.02 0.12 1.71 1.79e-07 lodine-133 3.25e+02 1.02 0.12 1.71 6.83e+01 Iodine-135 1.79e+00 1.02 0.12 1.71 3.76e-01 Krypton-83m 3.68e-10 1.02 0.10 1.71 6.44e-1 I Krypton-85 2.S8e+01 1.02 0.30 1.71 1.35e+01 Krypton-85m S. Sc-03 1.02 0.10 1.71 l.43e-03 Krypton-87 1.02e-14 1.02 0.10 1.71 1.79e-15 Krypton-88 4.21 e-05 1.02 0.10 1.71 7.37e-06 Xenon-131m 1.91 e+01 1.02 0.10 1.71 3.34e+00 Xenon-133m 6.42e+01 1.02 0.10 1.71 1.12e+01 Xenon-133 2.8e+03 1.02 0.10 1.71 4.92e+02 Xenon-135m 2.74e-01 1.02 0.10 1.71 4.80e-02 Xenon-135 3.69e+01 1.02 0.10 1.71 6.46e+00 Rubidium-88 0.00e+00 1.02 0.00 1.71 0.00e+00 Cesium-135 0.00e+00 1.02 0.00 1.71 0.00e+00 (a)
The Rb-88 and Cs-I 35 daughter isotopes that the LOCADOSE Code library file created are each listed with zero initial (72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />) and adjusted gap activities.
(b)
Sample Calculation for 1-133 Gap Inventory, (325 cilrod) x (1.02)
(0.12 gap fraction) x (1.71 RPF) 68.3 ci/rod
E&TS DEPARTMENT CALCULATION SHEET r-roject or DCPIFCNIECP C
ICCN NOJ PRELIM. CCN NO.
PAGE _ OF _
alc. No. N4072-003 CCN CONVERSION:
I CCN NO. CCN -
Subject Fuel Handling Accident (FHA) Inside Containment - Control Room 8 Offsite Doses Sheet 76 of 241 REV ORIGINATOR DATE IRE DATE REVS ORIGINATOR DATE I IRE DATE R
IA. Rustae T. Remick l
4 E
I-
=
=
Table 8.1-3 Once Burned Peak Fuel Rod Gap Inventory Considering 210 Fuel Rod Failures Isotopc &'
Average Fuel Power Uprate Once Burned Once Burned Once Burned Rod Correction Gap Release Radial Peaking Peak Fuel Rod Inventory Factor Fraction Factor Gap Inventory (b) 72 hrs decay 210 Fuel Rod 210 Fuel Rod Failures Failures
[per Table 4.1 -1]
[per DI 4.1.1]
[per DI 4.1.2]
[per DI 4.1 A.2]
(curies/rod)
(unitless)
(unitless)
(unitless)
(curies/rod)
Iodine-129 1.18e-04 1.02 0.10 1.37 1.65e-05 Iodine-130 3.09e-02 1.02 0.10 1.37 4.33c-03 Iodine-131 I.40e+03 1.02 0.10 1.37 1.96e+02 lodine-132 8.53e-07 1.02 0.10 1.37 1.19e-07 lodine-133 3.25e+02 1.02 0.10 1.37 4.55e+01 Iodine-135 1.79e+00 1.02 0.10 1.37 2.51e-01 Krypton-83m 3.68e-10 1.02 0.10 1.37 5.15e-11 Krypton-85 2.58e+0l 1.02 0.30 1.37 1.08e401 Krypton-85m 8.15e-03 1.02 0.10 1.37 1.14e-03 Krypton-87 1.02e-14 1.02 0.10 1.37 1.43e-15 Krypton-88 4.21 e-05 1.02 0.10 1.37 5.89e-06 Xenon-131 m 1.91e+01 1.02 0.10 1.37 2.67e+00 Xenon-133m 6.42e+01 1.02 0.10 1.37 8.99c+00 Xenon-133 2.81 e+03 1.02 0.10 1.37 3.93e+02 Xenon-135m 2.74e-01 1.02 0.10 1.37 3.84e-02 Xenon-135 3.69e+01 1.02 0.10 1.37 5.17e+00 Rubidiun-88 0.00e+00 1.02 0.00 1.37 0.00e+00 Cesiumn-135 0.00e+00 1.02 0.00 1.37 0.OOe+00 (a)
The Rb-88 and Cs-135 daughter isotopes that the LOCADOSE Code library file created are each listed with zero initial (72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />) and adjusted gap activities.
(b)
Samplc Calculation ror 1-133 Gap Inventory, (325 ci/rod) x (1.02) x (0.10 gap fraclion) x (1.37 RPF) = 45.5 ci/rod
E&TS DEPARTMENT CALCULATION SHEET ICCN NOJ PRELIM. CCN NO.
PAGE _
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rroject or DCP/FCNIECP Calc. No. N4072-003
_ CCN CONVERSION:
CCN NO. CCN -
4'tii~cv-
~Muil Wone4ffinn Ar-iji4nnt IMI-AI neiAn Cr~ninman -
-e~nfrnIl De~em R. (wf'mt rnc Qh~at 77 r^f 91 REV ORIGINATOR DATE IRE I
DATE REV ORIGINATOR DATE IRE DATE R
4 IA. Rustaey T. Remick l
l E
_ - -l _ _ _ _MA~
Table 8.1-4 High Burnup Peak Fuel Rod Gap Inventory Considering 210 Fuel Rod Failures Isotope ¢"
Average Fuel Rod Power Uprate High Burnup High Burnup High Burnup Inventory Correction Gap Release Radial Peaking Peak Fuel Rod 72 hrs decay Factor Fraction Factor Gap Inventory 01 210 Fuel Rod 210 Fuel Rod Failures Failures
[per Table 4. 1 -1]
[per DI 4.1.1]
[per DI 4.1.2]
[per DI 4.1 A.21 (curies/rod)
(unitless)
(unitless)
(unitless)
(curies/rod)
Iodine-i 29 1.18e-04 1.02 0.12 1.37 1.98e-45 Iodine-130 3.09e-02 1.02 0.12 1.37 5.19e-03 lodine-131 1.40e+03 1.02 0.12 1.37 2.35e+02 lodine-132 8.53e-07 1.02 0.12 1.37 1.43e-07 Iodine-133 3.25e+02 1.02 0.12 1.37 5.46e+01 Iodine-135 1.79e+00 1.02 0.12 137 3.01e-0I Krypton-83m 3.68e-10 1.02 0.10 1.37 5.15e-11 Krypton-85 2.58e+01 1.02 0.30 1.37 1.08e+01 Krypton-85m 8.15e-03 1.02 0.10 1.37 1.14e-03 Krypton-87 1.02e-14 1.02 0.10 1.37 1.43e-15 Krypton-88 4.21e-05 1.02 0.10 1.37 5.89e-06 Xenon-131m 1.91 e+01 1.02 0.10 1.37 2.67e+00 Xenon-133m 6.42e+01 1.02 0.10 1.37 8.99e+00 Xenon-133 2.81e+03 1.02 0.10 1.37 3.93e+02 Xenon-135m 2.74e-0l 1.02 0.10 1.37 3.84e-02 Xenon-135 3.69e+01 1.02 0.10 1.37 5.17e+00 Rubidium-88 0.00e+00 1.02 0.00 1.37 0.00e+00 Cesium-135 0.00e+00 1.02 0.00 1.37 0.00e+00 (a)
The Rb-88 and Cs-1 35 daughter isotopes that the LOCADOSE Code library file created are each listed with zero initial (72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />) and adjusted gap activities.
(b)
Sample Calculation for 1-133 Gap Inventory, (325 cilrod) x (1.02) x (0.12 gap fraction) x (1.37 RPF) = 54.6 ci/rod
E&TS DEPARTMENT CALCULATION SHEET 1CN O./~
fPRELIM. CCN N 1
PAGE _OF _
r-roject or DCPIFCNIECP Calc. No. N407Z-uua CCN NO. CCN-Subject Fuel Handling Accident (FHA) Inside Containment - Control Room & Offsite Doses Sheet 78 of 241 REV ORIGINATOR DATE IRE DATE REV ORIGINATOR DATE IRE DATE IR 4 A. Rustae T. Remick I
Table 8.1-5 Representative Failed Fuel Rod Gap Inventory to be Modeled Isotope ° High Bumup Peak Fuel Rod High Burnup Peak Fuel Rod Representative Fuel Rod Gap Inventory Gap Inventory Gap Inventory (1) 16 Fuel Rod Failures 210 Fuel Rod Failures 226 Fuel Rod Failures
[per Table 8.1-2]
[per Table 8.1-4]
(curiestrod)
(curies/rod)
(curies/rod)
Iodine-129 2.48e-05 1.98e-05 2.02c-05 Iodine-130 6A9e-03 5.19e-03 5.28e-03 Iodine-131 2.94e+02 2.35e+02 2.39e+02 Iodine-132 1.79e-07 1.43e-07 1.46e-07 Iodine-133 6.83e-01 5.46e+01 5.56e+01 lodine-135 3.76e-01 3.01 e-Ol 3.06e-01 Krypton-83m 6.44e-I 5.15e-I I 5.24e-1 1 Krypton-85 1.35e+01 1.08e+01 I.1Oe+0 Krypton-85m 1.43e-03 1.14e-03 1.16e-03 Krypton-87 1.79e-1 5 I A3e-15 1.46e-1 5 Krypton-88 7.37e-06 5.89e-06 5.99e-06 Xenon-1 31 rn 3.34e+00 2.67e+00 2.72e+00 Xenon-133m 1.12e+Ol 8.99e+00 9.15e+00 Xenon-133 4.92e+02 3.93e+02 4.OOe+02 Xenon-135m 4.80e-02 3.84e-02 3.91 e-02 Xenon-135 6.46e+00 5.1 7e+00 5.26e+00 Rubidium-88 0.O0e+00 O.OOe+00 O.OOe+00 Cesium-135 0.OOe+00 0.OOe+O O.OOe+00 (a)
The Rb-88 and Cs-135 daughter isotopes that the LOCADOSE Code library file created are each listed with zero gap activities.
(b)
Sample Calculation for 1-133 Gap Inventory
[(68.3 ci/rod) x (16 failed fuel rods)] + 1(54.6 ci/rod) x (210 failed fuel rods)]) + (226 failed fuel rods) 55.6 ci/rod I
I I
I I
I I
E&TS DEPARTMENT CALCULATION SHEET ICCN NOJ PRELIM. CCN NO.
PAGE _ OF _
,. ject or DCP/FCN/ECP Caic. No. N4072-003 CCN CONVERSION:
I CCN NO. CCN -
Subject Fuel Handling Accident (FHA) Inside Containment - Control Room & Offsite Doses Sheet 79 of 241 REV ORIGINATOR DATE IRE DATE REV ORIGINATOR DATE IRE DATE R
4 IA Rustaey T. Remick l
l E
l I I
I I_
=
{~~~~~
8.2 Refueling Water Iodine Removal The iodine gap activity released to the water is subject to clean-up due to pool scrubbing. Per Design Input 4.2.1, the pool iodine decontamination factors are valid for a minimum 23 foot water depth between the top of the damaged fuel rods and the fuel pool surface, and for a maximum fuel rod pressure of less than 1200 psig.
8.2.1 Compliance with 23 foot Water Depth Requirement. Per Technical Specification LCO 3.9.6 (References 6.4c and 6.4d), during movement of irradiated fuel assemblies within containment, the refueling water level above the top of the reactor vessel flange shall be greater than or equal to 23 feet. Except for the case where the damaged bundle is lying on the reactor vessel flange, there will be greater than 23 feet of water above the damaged bundle. Per Design Input 4.1.3, the maximum number of fuel rods predicted to fail will occur as a result of the vertical drop of the fuel assembly onto the fuel bundles in a partially loaded core. Since the damage occurs inside the reactor vessel, all of the damaged fuel rods are situated below the reactor vessel flange. As such, a minimum 23 foot water depth exists between the top of the damaged fuel rods and the fuel pool surface.
8.2.2 Compliance with 1200 psig Fuel Rod Pressure Requirement. Per Design Input 4.13, the average rod maximum internal pressures as a function of void (rod gap) volume and plenum temperature are:
Plenum Temperature Void Volume Rod Internal Pressure (with temporary gas release) 613OF 0.8388 in' 2170.4 psia 700F 1.0843 in' 829.4 psia Applying the Ideal Gas Law (PV = nRT) to each set of these conditions yields the same nR value:
nR = PV/T = (2170.4 psia)(0.8388 in3)/(613OF + 460'R) = 1.697 psia-in3/OR nR = PV/T = (829.4 psia)(l.0843 in3)/(70'F + 460'R) = 1.697 psia-in3/OR The fact that the same nR value is calculated for each set of conditions indicates that the fuel rod gap gases behave consistent with the ideal gas law over this temperature range. This conclusion is confirmed by a discussion in ABB Memorandum ST-98-427 (Reference 6.3a), which notes that internal pressure at desired conditions may be determined by application of the perfect gas law based on constant moles of gas (n).
E&TS DEPARTMENT CALCULATION SHEET ICCN NOJ PRELIM. CCN NO.
PAGE _
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rroject or DCP/FCN/ECP Calc. No. N-4072-003 CCN CONVERSION:
l CCN NO. CCN-icrt Fel Hndlinn Arieint (FHAI Insirde Cnntainment - Cnntrol Rnnm r. OfftitA Dosn Sheet 80 f 241 REV ORIGINATOR DATE IRE DATE REV ORIGINATOR DATE IRE DATE R
4 A. Rustae T. Remick I
I I
I Since the Ideal Gas Law is applicable, the product of the rod internal pressure and void volume (i.e., PV) is directly proportional to temperature (T). A discussion in ABB Memorandum ST-98427 notes that linear interpolation between the hot void volume at hot conditions (based on hot plenum temperature [613'F]) and cold void volume at 70'F will conservatively determine the void volume at spent fuel pool conditions. Linear interpolation over this temperature range determines that a temperature of 256.00F and a void volume of 1.0002 in3 will yield a rod internal pressure of 1200 psig.
V(at256.00F) -
V(at7O0 F) 256.0F-700F V(at6130F) - V(at70°F) 6130F-700F V(at256.0F) - 1.0843in3 256.0F - 700F 0.8388in3
- 1.0843in3 6130F - 700F V(at256.0F) = 1.0002in 3 P = nRT / V = (1.697 psia-in3IOR)(256.0cF + 460'R) / (1.0002 in3)
P = 1214.6 psia = 1200 psig Based on the preceding, if the plenum temperature at the time of the fuel handling accident is less than 256°F, then the rod internal pressure will be less than 1200 psig and the Regulatory Guide 1.25 pool iodine decontamination factors will be valid.
Per Assumption 3.18, the maximum refueling pool cooling water bulk temperature is 160'F.
Therefore, if it can be documented that the fuel rod gap space will be no more than 960F hotter than the bulk coolant temperature at shutdown conditions, then the fuel rod gap temperature will be less than 256°F, the rod internal pressure will be less than 1200 psig and the Regulatory Guide 1.25 pool iodine decontamination factors will be validated.
The temperature gradient between the fuel rod gap space and the bulk coolant can be determined using cylindrical fuel element heat transfer relationships documented in Principles of Nuclear Reactor Engineering (Reference 6.6h, Sections 11.86 through 11.91). The lemlperature differences between the inner and outer surfaces of the fuel cladding (positions I and 2, respectively) and the coolant (position are:
E&TS DEPARTMENT CALCULATION SHEET ICCN NOJ PRELIM. CCN NO.
PAGE _ OF_
. oject or DCP/FCN/ECP Caic. No. N-4072-003 -
CCN CONVERSION:
I CCN NO. CCN -
Subject Fuel Handling Accident WFHAI Inside Containment - Control Room & Offsite Doses Sheet 81 of 241 REV ORIGINATOR DATE IRE I
DATE REV ORIGINATOR DATE IRE DATE R
4 A. Rustaey
. Remick
__l___l_v l
I I
I I
I lt-Q a x ( 1 b +
I 2
kc a
hb
=
Qa 2 2 h b T2 -
Tf where:
T.
=
Temperature at inner surface of the fuel cladding (F)
T2
=
Temperature at outer surface of the fuel cladding (F)
Tr
=
Temperature of the bulk coolant (OF)
Q
=
Volumetric source strength (Btu/hr-ft')
a
=
inner radius of the fuel cladding (fit) b
=
outer radius of the fuel cladding (ft) kc
=
thermal conductivity of the fuel cladding (Btulhr-ft- 0F) h
=
heat transfer coefficient between the fuel cladding and coolant (Btu/hr-fit2-F)
The volumetric source strength (Q), which is a measure of the rate of heat release per unit volume of fuel, is equal to the quotient of the linear heat rate (q') and the cross-sectional area of the fuel:
Q
=
q
/
a2 A review of the preceding equations and parameters shows that the temperature gradient between the inner surface of the fuel cladding and the bulk coolant is directly proportional to the linear heat rate, and inversely proportional to the heat transfer coefficient between the fuel cladding and coolant:
T - Tf Qlh (T 1 -
Tf)t.shutdown (T 1 -
Tf)at shutdown
=
(T at power
=
(T f)a, power x (Q / h)., 5 hw wn (Q / h)otP0,,,r x
Qatshurdo'n x power Qtpower hot shutdoun This final relationship is solved to estimate the temperature gradient between the fuel rod gap space and the bulk coolant at shutdown conditions.
E&TS DEPARTMENT CALCULATION SHEET ICCN NOJ PRELIM. CCN NO.
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OF _
Project or DCPIFCN/ECP Caic. No. N4072-003 CCN CONVERSION:
I CCN NO. CCN -
qqiict-ial I-Innlinn Arcritiont lFHMAI Insidep Containment. - Cnntrnl Ronm lX Offsita flncac Sheht a9 f 9A1 REV ORIGINATOR DATE IRE DATE REV ORIGINATOR DATE IRE DATE R
4 A. Rustae" T. Remick I
v (T - Tf)tpower Per Assumption 3.19, the temperature at the inner surface of the fuel cladding during power operations is assumed to be no more than 500 F greater than the bulk temperature of the surrounding cooling water.
Qa shutdown / Qat power Per Assumption 3.20, when refueling operations begin at 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after shutdown, the heat generation rate associated with residual radioactive decay is assumed to be less than one percent of the heat generation rate during power operations.
ba power Manipulations of previously defined equations allow for the heat transfer coefficient between the fuel cladding and the coolant to be calculated with:
Q a 2 q
h
=
_a q
2 b (T2 - T1 )
2 b (T 2 - T)
Per Assumption 3.19, for a linear heat rate of 8.62 kw/ft and a fuel rod outer diameter of 0.382 inches, the temperature difference between the outer surface of the fuel cladding and the coolant is less than 50'F. These values indicate that the heat transfer coefficient between the fuel cladding and the coolant during power operations is approximately:
h
= (8.62kw/ft) x (3413Btulkw-hr) x (12in/ft) 2950Btu/hr ft 2 -F 2
x (0.382in) x (50OF) bat shutdown The heat transfer coefficient between the fuel cladding and the coolant during shutdown conditions can be estimated using natural convection heat transfer from vertical cylinders relationships documented in Principles of Heat Transfer. 2"" Edition (Reference 6.6m, Equation 6-4 on page 268, and Figure 7-4 on page 335):
c= Nu f L
when Gr x Pr - le (laminar region),
Nu
= 0.555 (GrPr)J4 when GrxPr > e9 (turbulent region),
Nu
= 0.0210 (GrPr)2"5 where:
hc
=
Convective heat transfer coefficient (Btu/hr-ftk-0F)
Nu
=
Average Nusselt number based on length of the fuel rod (unitless) kf
=
Thernal conductivity of the coolant (Btu/hr-ft-°F)
L
=
Characteristic Length [fuel rod length] (ft)
Pr Prandtl number = cp p / kr (unitless) [Reference 6.6m, page 3331
E&TS DEPARTMENT CALCULATION SHEET ICCN NOJ PRELIM. CCN NO.
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Project or DCPIFCN/ECP Caic. No. N4072-003 1
CCN CONVERSION:
I CCN NO. CCN -
Subject Fuel HandlincI Accident FHA) Inside Containment - Control Room & Offsite Doses Sheet 83 of 241 REV ORIGINATOR DATE IRE I
DATE REV ORIGINATOR DATE IRE DATE R
4 A. Rustae T. Remick i
v I
$~~~~~~~
Gr
=
Grashof number = L3 p2 g 3 AT / li2 (unitless) [Reference 6.6m, page 333]
CP Specific heat of the coolant (BTU/lbm-0F)
Ii
=
Viscosity of the coolant (Ibmn/ft-sec) p
=
Density of the coolant (Ibm/ft3) g
=
Gravity of 32.2 ft/sec2
(
=
Thermal coefficient of volume expansion of the fluid (1/0 F)
AT Temperature difference between the outer surface of the fuel cladding and the coolant (F)
Per Assumption 3.18, the maximum refueling pool cooling water bulk temperature is 160'F. Assuming a 500F temperature difference between the outer surface of the 12.5 foot long fuel cladding and the coolant, the above coolant thermal properties will be evaluated at an average coolant temperature adjacent to the fuel rod of 185 0F. Per Principles of Heat Transfer. 2nd Edition (Table A-3 for water):
k=
0.391 Btu/hr-ft-0 F Pr 2.14 p g /u
=
9.09e8 OF-ft-3 For these conditions, the Grashof number, and the product of the Grashof and Prandtl numbers are:
Gr = L 3ATx(p 2 g(P/p 2 )
Gr
= (12.5ft)3 x (5 0 °F) x (9.09e8F-'-f1 3)
= 8.9e3 GrPr = (8.9e)3) x (2.14)
= 1.9eJ4 Since the product of the Grashof and Prandtl numbers is greater than I e9, flow is turbulent and the heat transfer coefficient between the fuel cladding and the coolant during shutdown conditions is:
h
= [0.0210 (GrPr)2 5] x if c
[ 1 1.L h
[0.0210 x (1.9e14)2 5
.9Buh-l 0
338Btu/hr ft2
'F C
i ~~~~~~~~~2.5fl
E&TS DEPARTMENT CALCULATION SHEET ICCN NOJ PRELIM. CCN NO.
PAGE _
OF _
Project or DCP/FCNIECP Caic. No. N 4072-003
_ CCN CONVERSION:
CCN NO. CCN -
Subject Fuel Handling Accident (FHA) Inside Containment - Control Room & Offsite Doses Sheet 84 of 241 REV ORIGINATOR DATE IRE DATE REV ORIGINATOR DATE IRE DATE 4
A. Rustaey T. Remick E
L__
=
=
I Using the preceding parameters, the temperature gradient between the inner surface of the fuel cladding (position 1) and the bulk coolant (position f) at shutdown conditions is approximately:
(T Tf)at shutdown
= (T, -
Tf)apowver XQo shutdown xhatpower Qrnpower hatshutdown (T
Tf),tshuldown
= (500 °F) x (0.01) x (2950Btufhr-ft2 - 0F) 338Btulhr -.f2-°F (T
T)at shutdown
= 440F This temperature gradient between the inner surface of the fuel cladding and the bulk coolant is consistent with the'assumed 50 0F temperature difference between the outer surface of the fuel cladding and the coolant, thereby validating the use of an average coolant temperature adjacent to the fuel rod of 185 F in determining the coolant thermal properties.
Under the premise that the temperature gradient across the fuel rod gap space is minimal at equilibrium conditions (such as at shutdown), the temperature in the fuel rod gap space can be approximated by the temperature at the inner surface of the fuel cladding. Therefore, the temperature gradient between the fuel rod gap space and the bulk coolant at shutdown conditions is less than 50F.
Based on the preceding, since the fuel rod gap space will be less than 96 0F hotter than the bulk coolant temperature at shutdown conditions, the rod internal pressure will be less than 1200 psig and use of the Regulatory Guide 1.25 pool iodine decontamination factors is validated.
E&TS DEPARTMENT CALCULATION SHEET ICCN NOJ PRELIM. CCN NO.
PAGE _ OF _
Project or DCP/FCN/ECP Caic. No. N-4072-003 CCN CONVERSION:
I CCN NO. CCN -
Subject Fuel Handling Accident (FHA) Inside Containment - Control Room & Offsite Doses Sheet 85 of 241 REV ORIGINATOR DATE IRE DATE REV ORIGINATOR DATE IRE DATE R
=stae~
T. Remick IIE I
I 8.3 Initial Containment Airborne Activity Profile Table 8.3-1 presents the initial containment airborne activity profile to be input into the LOCADOSE Code.
The quantity of iodine evolving into the containment air space is dependent on the iodine species makeup of the fuel rod gap iodine inventory, the ability of the water surrounding the damaged fuel to retain the released iodine, and the total number of failed fuel rods. Per Design Input 4.1.5, the fuel rod iodine gap inventory is composed of 99.75 percent inorganic species (i.e., elemental and particulate iodine), and 0.25 percent organic species (i.e., organic iodide).
Per Design Input 4.2.1 (as justified in Section 8.2), the pool decontamination factors for the inorganic iodine and organic iodide species are 133 and 1, respectively. Per Design Input 4.1.3, in the event of a fuel handling accident inside containment a total of 226 fuel rods will fail.
Applying these fractions and factors and the total number of failed fuel rods to the representative failed fuel rod iodine gap inventory identified in Table 8.1-5 results in the initial containment airborne iodine activity profile presented in Table 8.3-1.
The quantity of noble gases evolving into the containment air space is dependent on the fuel rod gap noble gas inventory, the ability of the water surrounding the damaged fuel to retain the released noble gas, and the total number of failed fuel rods. Per Design Input 4.2.2, the retention of noble gases in the pool is negligible (i.e., a decontamination factor of one). Per Design Input 4.1.3, in the event of a fuel handling accident inside containment a total of 226 fuel rods will fail. Applying the decontamination factor and total number of failed fuel rods to the representative failed fuel rod noble gas gap inventory identified in Table 8. 1-5 results in the initial containment airborne noble gas activity profile presented in Table 8.3-1.
The initial refueling pool activity for isotopes of interest are shown in Table 8.3-la. Per Design Input 4.1.5, the fuel rod iodine gap inventory is composed of 99.75 percent inorganic species (i.e., elemental and particulate iodine), and 0.25 percent organic species (i.e., organic iodide).
Per Design Input 4.2.1 (as justified in Section 8.2), the pool decontamination factors for the inorganic iodine and organic iodide species are 133 and 1, respectively. Therefore, the Iodine fractions used in LOCA DOSE are as follows, Organic:
0.0025 I = 0.0025 Inorganic:
0.9975 + 133 = 0.0075
E&TS DEPARTMENT CALCULATION SHEET ICCN NO./
PRELIM. CCN NO.
PAGE OF _
Project or DCP/FCN/ECP Calc. No. N4072-003 CCN CONVERSION:
I CCN NO. CCN-Subject Fuel Handling Accident (FHA) Inside Containment - Control Room & Offsite Doses Sheet 86 of 241 REV ORIGINATOR DATE IRE DATE REV ORIGINATOR DATE IRE DATE I R 41 A ta T. Remick
-u=-
Table 8.3-1 Initial Containment Airborne Activity Profile isotope ')
Representative Fuel Rod Iodine Species Iodine Species Number of Initial Gap Inventory Fractions Decontamination Failed Fuel Containment 226 Fuel Rod Failures Factor Rods Airborne
[per Table 8.1 -5]
[per Dl 4.1.5]
[per DI 4.2.11
[per DI 4.1.3]
Activity Mb)
(curies/rod)
(unitless)
(unitless)
(rods)
(curies) 1-129 elemental 2.02e-05 0.9975 133 226 3.42e-05 1-129 organic 2.02e-05 0.0025 I
226 1.14e-05 1-130 elemental 5.28e-03 0.9975 133 226 8.95e-03 1-130 organic 5.28e-03 0.0025 I
226 2.98e-03 1-131 elemental 2.39e+02 0.9975 133 226 4.05e+02 1-131 organic 2.39e+02 0.0025 1
226 1.35e+02 1-132 elemental 1.46e-07 0.9975 133 226 2.47e-07 1-132 organic 1.46e-07 0.0025 1
226 f.25e-08 1-133 elemental 5.56e+01 0.9975 133 226 9.42e+01 1-133 organic 5.56e+01 0.0025 1
226 3.14e+01 1-135 elemental 3.06e-01 0.9975 133 226 S.19e-01 1-135 organic 3.06e-01 0.0025 1
226 I.73e-O1 Krypton-83m 5.24e-I 1 226 1.18e-08 Krypton-85 I.IOe+01 226 2.49e+03 Krypton-85m 1.16e-03 226 2.62e-01 Krypton-87 IlA6e-15 226 3.30e-13 Krypton-88 5.99e-06 226 1.35e-03 Xenon-131 m 2.72e+00 226 6.1Se+02 Xenon-133m 9.15e+00 226 2.07e+03 Xenon-133 4.00e+02 226 9.04e+04 Xenon-135m 3.91e-02 226 8.84e+00 Xenon-135 5.26e+00 226
- 1. I 9e+03 Rubidium-9R 0.OOe+0O 226 O.Oe+0 Cesium-135 0.00e+00 226 0.OOeO0 (a)
The Rb-88 and Cs-35 daughter isotopes that the LOCADOSE Code library file created are each listed with zero initial containment airborne activities.
(b)
Sample Calculations for Initial Containment Airbome Activities:
Iodine-I 31 Elemental = 1(239 ci/rod) x (0.9975) + (133)] x (226 failed fuel rods) = 4.05e2 ci Krypton-85 = (11.0 ci/rod) x (226 failed fuel rods) = 2.49e3 ci
E&TS DEPARTMENT CALCULATION SHEET ICCN NOJ PRELIM. CCN NO.
PAGE _
OF _
Project or DCP/FCN/ECP Calc. No. N4072-003 CCN CONVERSION:
I CCN NO. CCN -
.¢gmhji.
Mmi I-Ian, H
nr A Irri nt IC1I-IA)
I,.eneoA m
-ninrn,
- frntrnfDl 0 m k li elffgin rf
-ea Skhmgt R7 f '3A 1 tJJoU INU
- JN NNNI' WNNI'dULN ru IrNIowl
- N*UtJ~~JIVCi*
Irnn Ill1u kuaIlllS VII.
19 UIU I NIN II 0 lNJlZ V
V:C 5.I I%:I Ul SI
& 1 REV ORIGINATOR DATE IRE DATE REV ORIGINATOR DATE IRE DATE R
4 A. Rustae T. Remick I
v ffi~~~~~~
I Table 8.3-la Initial Refueling Pool Activity Profile Isotope b Representative Fuel Rod Number of Initial Gap Inventory Failed Fuel Rods Refueling Pool 226 Fuel Rod Failures 1per Dl 4.1.3]
Activity
[per Table 8.1-5]
(rods)
(curies)
(curies/rod)
I 1-129 2.02e-05 226 4.57e-03 1-130 5.28e-03 226
).19e+00 1-131 2.39e+02 226 5.40e+04 1-132 1.46e-07 226 3.30e-05 1-133 5.56e+01 226 1.26e+04 1-135 3.06e-0l 226 6.92e+01 Krypton-83m 5.24e-1 1 226 1.1 8e-08 Krypton-85 1.10e+01 226 2.49e+03 Krypton-85m 1.16e-03 226 2.62e-01 Krypton-87 1.46e-15 226 3'.30e-13 Krypton-88 5.99e-06 226 1.35e-03 Xenon-131m 2.72e+00 226 6.15e+02 Xenon-I 33m 9.15e+00 226 2.07e+03 Xenon-133 4.00c+02 226 9.04e+04 Xenon-I 35m 3.9 1e-02 226 8.84e+00 Xenon-135 5.26e+00 226 1.19e+03 Rubidiun-88 0.00e+00 226 0.00e+00 Cesium-135 0.00e+00 226 0.00c+O0 (a)
The Rb-88 and Cs-1 35 daughter isotopes that the LOCADOSE Code library file created are each listed with zero initial containment airborne activities.
(b)
Sample Calculations for Initial Airborne Activities:
Iodine-I 31 - (239 ci/rod) x (226 failed fuel rods) = 5.40e+04 ci
E&TS DEPARTMENT CALCULATION SHEET ICCN NO.1 I
PRELIM. CCN NO.
PAGE _ OF -
Project or DCP/FCNIECP Calc. No. N4072-003 CCN CONVERSION:
CCN NO. CCN -
Subject Fuel Handling Accident (FHA) Inside Containment - Control Room & Offsite Doses Sheet 88 of 241 REV ORIGINATOR DATE IRE I
DATE REVI ORIGINATOR DATE IRE DATE R
i aeIT.
Remick i
I 8.4 Modeling of Control Room Isolation Per Design Input 4.5.1 the control room HVAC system will automatically transfer to the high radiation isolation mode if high radiation is sensed by either 2/3RE7824G1 or 2/3RE7825G2.
The time period to isolate the control room is based on the following:
time needed for the contaminated air to travel from the Containment release point to the control room normal HVAC intake time needed for the intake plenum to fill time needed for the radiation monitor to initiate the isolation signal time needed for the isolation dampers to close 8.4.1 Activity Concentration at Control Room Air Intake. The activity release rate to the outside environment may be converted to an activity concentration at the control room intake as follows:
Activity Release (C) x x( sec m 3 o
10 ACi = Concentration(
QC time interval (sec)
Q m 3 106 CC Ci CC The LOCADOSE activity transport output file included in Section 9.2 contains a time step at 10 seconds (0.00278 hours). This interval, and the design basis 0 to 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Control Room atmospheric dispersion factor (X/Q per Design Input 4.7) yields the following equation for converting the activity release during the first LOCADOSE time step to an activity concentration at the control room intake:
I Concentration CR Intake
) =
CC Activity Release )-0.00278hr (ci) 3.lxO0 -3 sec x (m 3 -1s) 10sec m 3 (CC-Ci)
Concentration CR lntae (pa*)
= Activity Release 0. 002 78h, (Ci) x 3.lxlO-4(
)Ci)
CC cc-C, 8.4.2 Intake Plenum Fill Time. Per Design Input 4.5.2, the time necessary for the plenum activity concentration (at the detector location) to reach 90% of the outside air activity concentration is 8.5 seconds.
E&TS DEPARTMENT CALCULATION SHEET ICCN NOJ PRELIM. CCN NO.
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eroject or DCPIFCN/ECP Caic. No. N-4072-003 CCN CONVERSION:
I CCN NO. CCN -
Subject Fuel Handling Accident (FHAI Inside Containment - Control Room & Offsite Doses Sheet 89 of 241 REV ORIGINATOR DATE IRE DATE REV ORIGINATOR DATE IRE DATE R
I I
TI.~~~~~
1~ E 4 A.RusneyT.Remick
-i II V
8.4.3 Radiation Monitor Trip Initiation Time. The radiation monitor response time and efficiency values are based on exposure to noble gas isotopes. Per Calculation J-SPA-179 (Reference 6. 1j, Section 2.2), control room radiation monitors 2/3RE7824GI and 2/3RE7825G2 measure the gross activity concentration in the outside air entering the control room area. A CRIS is produced if the activity concentration exceeds a trip setpoint. Per Design Input 4.5.2, the control room radiation monitor response time is slowest when the activity concentration in the outside air flow entering the control room is smallest. Since Krypton-85 is one of the many isotopes present in the "gross activity" accident release, the Krypton-85 concentration in the outside air flow entering the control room area will always be less than the gross activity concentration in the outside air flow entering the control room area. Therefore, this section calculates Krypton-85 concentration in the control room HVAC intake plenum to determine the maximum time required for the averaging algorithm to produce a representative concentration value that is greater than the alarm trip setpoint. This maximum time equates to the longest delay prior to the generation of a CRIS. Due to its long half-life of approximately 10.73 years, the activity profile of Krypton-85 in stored spent fuel will not be appreciably different from that of freshly irradiated fuel. Note: for completeness, Xenon-133 concentration data is also calculated.
Per the LOCADOSE activity transport output file (Section 9.2.3), the activity releases in the first 10 second interval are:
Isotope Containment Airborne Activity Released Krypton-85 23.84 Ci Xenon-133 865.3 Ci Per the equation in Section 8.4.1, these activity releases equate to the following activity concentrations at the control room outside air intake:
Kr-85:
23.84 Ci x 3.1E-4 ACi/(cc-Ci) = 7.39E-3 Ci/cc at intake Xe-133:
865.3 Ci x 3.IE-4 mCi/(cc-Ci) = 2.68E-l IsCi/cc at intake Per Section 8.4.2, within 8.5 seconds the plenum activity concentrations (at the detector location) are 90% of the outside air activity concentrations:
Kr-85:
0.9 x 7.39E-3 Ci/cc = 6.65E-3 uCi/cc at detector location Xe-133:
0.9 x 2.68E-l IuCi/cc = 2.4 1E-I Ci/cc at detector location
E&TS DEPARTMENT CALCULATION SHEET ICCN NOJ PRELIM. CCN NO.
PAGE _
OF _
Project or DCP/FCNIECP Caic. No. N072-003 CCN CONVERSION:
I CCN NO. CCN --
S.-IU..AIehi A--
Anrn IUA%
1i.A.,
0 ^r m n
flffeifg hfmene hartm Olf}
f 1
uU IJtJ~V I UI I amI ull it IF i.,uu i
DUII CmIa I mm mIll - _^lUI mu AU WWL1 1
u;uI II1; L U:;
L mu We Cs-i REV ORIGINATOR DATE IRE DATE REV ORIGINATOR DATEI IRE DATE RF1 R
E V A. R u s t e e
. R e m i c k I
I I
4 A. Rustee lT.
Remick l
_ I I I
I I
~~~~~
Per Design Input 4.5.2, the trip initiation time for either of these activity concentrations (including the time assumed for digital signal processing) is 60 seconds.
8.4.4 Total High Radiation Isolation Time. Based on the timing determined in the preceding sections, the total control room isolation time for a high radiation induced CRIS is:
transit time to the HVAC intake per Assumption 3.7 intake plenum fill time per Section 8.4.2 radiation monitor trip initiation time per Section 8.4.3 damper closure time per Assumption 3.8 total 0.0 seconds 8.5 seconds 60.0 seconds 6.0 seconds 74.5 seconds To provide margin in this analysis, and to not require a restrictive CRIS response time, this analysis will model a control room isolation time of 3 minutes for any high radiation induced CRIS.
8.4.5 Radiological Consequences of a CRIS Failure. Per Section 8.4.4, this analysis will model a control room isolation time of 3 minutes for any high radiation induced CRIS. In the event of a CRIS failure, the calculated control room doses will remain valid provided that manual Operator Action is taken to isolate the control room within this same 3 minute period, or manual Operator Action is taken to isolate the control room prior to any released radioactivity passing through the Control Room HVAC isolation dampers.
Operator Action within 3 minutes is reasonable due to the Units 2&3 Licensee Controlled Specification 3.9.102 (References 6.4e & 6.4f) requirement for direct communications between the Control Room and the Senior Reactor Operator supervising core alterations, and due to the requirements of Abnormal Operating Instruction S023-13-20 (Reference 6.5c). The AOI addresses Refueling Accidents. The AOl states that symptoms of a refueling accident include high radiation levels in the Containment, and bubbles emerging from a submerged fuel assembly which has been dropped or damaged. In the event of a fuel handling accident with high radiation levels, the AOI requires immediate initiation of CRIS. As such, the Control Room Operators can be made aware of a dropped fuel assembly, and the potential need to isolate the control room.
I I
I
E&TS DEPARTMENT CALCULATION SHEET ICCN NOJ PRELIM. CCN NO.
PAGE_ OF _
Project or DCPJFCNIECP Caic. No. N-4072-003 CCN CONVERSION:
I CCN NO. CCN -
Subject Fuel Handling Accident (FHA) Inside Containment - Control Room & Offsite Doses Sheet 91 of 241 REV ORIGINATOR DATE IRE DATE REV ORIGINATOR DATE IRE DATE R
4 A. Rustaey T. Remick I
I v
8.5 Modeling of Flow from Containment to CR HVAC Intake Filter Upon receipt of a Control Room Isolation Signal the control room HVAC system is automatically shifted to the emergency mode of operation. Transfer to the emergency mode includes starting the emergency air conditioning units, opening the outside air isolation damper to the emergency filtration trains, and starting the fans. Per Assumption 3.14, this analysis models two trains of emergency HVAC in operation during the first 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> of the accident.
Operator action is assumed to be taken within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> to deactivate one train of emergency intake and recirculation units.
As discussed in Section 5.4.3, the control room charcoal filter shine dose is dependent on the CR HVAC intake filters' instantaneous activity loading at various times during the FHA event.
Due to atmospheric dispersion, only a portion of the Containment activity release will become entrained in the Control Room HVAC intake flow. To analyze this dispersion and intake into the control room, the LOCADOSE Code will model a direct flowpath between the Containment Building (Region 2) and a Control Room HVAC Intake Filter (Region 3). An equivalent unfiltered flow rate from the Containment Building to the CR HVAC Intake Filter can be determined as follows:
CuriesCR Intake Filter = (Activity Conc at CR Outside Air Intake) x CR Intake Flowrate CuriescRfF = ([Activity Released From CtInt] x XlQcmH o CR) X CR Intake Flowrate Curiesc
=
Ctmi Activity x Ctmt Exhaust Flowrate] x XQ) x CR Intake Flowrate
~CRIF Ctmt Volume x
xC naeFort Since the LOCADOSE Code determines the Containment activity per Containment volume, the direct Containment to CR HVAC intake filter (Region 3) flowrate would be as follows:
CuriesCPJF = [Ctmt Activity] x [FlowrateCt., to Region 3]
Flowratemr to Region 3 = Ctmt Exhaust Flowrate x XIQC,., lo CR x CR Intake Flowrate Although atmospheric dispersion between the Containment and control room varies with time.
per Design Input 4.3.2 virtually all of the activity is released from the containment during the first 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> of the event. Therefore, a single Region 2 to 3 flowrate determined by using the 0 to 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> Control Room atmospheric dispersion factor from Design Input 4.7 will be applied for the 30 day analysis duration. In addition, although one of the two filtered control room intake flow paths will be isolated at 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, since virtually all of the activity is released from the containment during the first 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> of the event, the activity introduced and retained within each intake filter (i.e., the filter loading) will be identical.
E&TS DEPARTMENT IUUN NUJi CALCULATION SHEET PRELIM. CCN No Project or DCP/FCN/ECP Caic. No. N4072-003 cc cc SuiC.
44,o+Mul LHandlinn Adde4rn# (FHA% Insidem Co~nnainment-Co~ntrol CRoom & tlffeitm D-nes Sheet 92 f
A4
_Owuj~.U A ull I 1W *WJSfl' flt AI-W AL W..I*
A IS e
- I -
WI IA4 0t.O
- d. 1.iS.
W.;
Ad I
REV ORIGINATOR DATE IRE DATE REV ORIGINATOR DATE IRE DATE R
4 A. Rustae T. Remick I
7
iiiIii
=
4 Because the purpose of this flowrate is to determine the activity buildup in one Control Room HVAC intake filter, a flowrate of 2200 cfrn (per Design Input 4.4.3) will be used for the CR intake flow. Applying the 0 to 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> control room X/Q factor from Design Input 4.7, the equivalent unfiltered flowrate with a total Containment exhaust flowrate of 82000 cfin (Design Input 4.3.2) is as follows:
Flowratechn,,o Region 3 = 82000ft 3 x [3. WxO 3 sec min 0.02832 m 3 x 2200 ft3 min 60 sec ft 3 min FlowrateCm, o Region 3 264 cfm Modeling the flow into a Control Room HVAC intake filter (Region 3) in this manner slightly reduces the flow from the Containment (Region 2) to the environment from 82000 cftn to 81736 cfin, and thus slightly reduces the resulting offsite and control room inhalation and immersion doses. The dose reduction is equivalent to the reduction in the containment exhaust flow to the environment, or (264 cfin - 82000 cfin) x 100% = 0.3%. This dose impact is considered to be negligible since it represents only 0.2 rem of the 75 rem thyroid inhalation dose criterion.
Per assumption 3.12, the intake filter shine model also accounts for the recirculation filter shine.
Therefore, the normal intake flow (for the first 3 minutes) and the inleakage flow must also be considered.
Table 8.5-1 I
I I
I I
I Volumetric Flowrate (2 train)
Duration Total Volume CFM Minutes Cubic Feet Recirculation, 3
17,460 Filtered 5820 CR HVAC isolates on hi rad
- HVAC, 117 514,800 Filtered 4400 CR HVAC in operation Inleakage 10 120 1,200 Unfiltered end of EAB dose (2 scenarios) 1000 analysis 120,000 Total 10 CFM Unfiltered Inleakage 1000 CFM Unfiltered inleakage 533,460 652,260 L
E&TS DEPARTMENT CALCULATION SHEET ICCN NOJ PRELIM. CCN NO.
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OF _
Project or DCP/FCNIECP Caic. No. N4072-003 CCN CONVERSION:
I CCN NO. CCN -
Subject Fuel Handling Accident (FHAI Inside Containment - Control Room & Offsite Doses Sheet 93 of 241 REV ORIGINATOR I DATE IRE I
DATE REV ORIGINATOR DATE IRE DATE R
4 A.
Ruge T. RemickY I___________
___________ I K
I I
Therefore the single train flow rate for event duration based on Table 8.5-1 are; For CR unfiltered inleakage of 10 CFM:
533,460 CF - (2 x 120 min) = 2,223 CFM I
For CR unfiltered inleakage of 1000 CFM:
652,260 CF -(2 x 120 min) = 2,718 CFM l
Based on above the revised flowrate from containment to region 3 for CR unfiltered inleakage of I
10 CFM is; I
Flowrate~m, to Region 3 82000ft 3 3.1 xlO 3 sec min 0.02832 m3 2223 ft3 Cit tlo Region 3 min 3
60sec ft 3 min FlowrateCtmto Region 3 = 267 cfm Similarly, the flowrate from containment to region 3 for CR unfiltered inleakage of 1000 CFM l
is; Flowratecm, to Regio 3 82000ft 3 3 lxJO-3sec min 0.02832 m 3 2718 flt3 mm m
x60 sec ft 3 min FlowrateCtt to Region 3 = 327 cfm A correction factor for the normal intake and inleakage flows is applied in Section 8.9.
For CR unfiltered inleakage of 1000 CFM, modeling the flow into a Control Room HVAC I
intake filter (Region 3) in this manner slightly reduces the flow from the Containment (Region l
- 2) to the environment from 82000 cfm to 81673 cfln, and thus slightly reduces the resulting offsite and control room inhalation and immersion doses. The dose reduction is equivalent to the reduction in the containment exhaust flow to the environment, or (327 cfin - 82000 cfmn) x 100% = 0.4%. This dose impact is considered to be negligible since it represents only 0.3 rem of the 75 rem thyroid-inhalation dose criterion.
Since a Control Room HVAC intake filter (Region 3) does not retain noble gases, the LOCADOSE Code will model a "Region 3 recirculation filter" (with an arbitrary flowrate of 1,000,000 cfm and a noble gas filter efficiency of 100 percent) to remove the noble gases that enter Region 3. To maximize the filter shine dose, the LOCADOSE Code will not model a "Region 3 exhaust flow"; this effectively retains 100 percent (less radioactive decay) of the iodine and particulates entering Region 3.
E&TS DEPARTMENT CALCULATION SHEET ICCN NOJ PRELIM. CCN NO.
PAGE _ OF _
Project or DCP/FCN/ECP Calc. No. N-4072-003 CCN CONVERSION:
I CCN NO. CCN -
Subject Fuel Handling Accident (FHAI Inside Containment - Control Room & Offsite Doses Sheet 94 of 241 4 A. Rustaey T. Remick I
E 7 I
=
=
+~~~~~
8.6 LOCADOSE Code Time Steps The time steps entered into the LOCADOSE Code were chosen to model the times at which parameters important to the analysis are changed (e.g., HVAC changes). Although the release of radioactivity from the Containment Building ceases within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> (per Assumption 3.4), this FHA inside containment dose analysis will be evaluated for an event duration of 30 days, which will maximize the direct gamma radiation shine dose contribution from the activity retained on the control room HVAC filter.
Table 8.6-1 LOCADOSE Code Time Steps Time Step (hours after start Significance of the Time Step of FHA event) 0 hrs a)
Beginning of Fuel Handling Accident Inside Containment limiting fault 0.00278 hr Intermediate time step for release data information (10 seconds) 0.05 hr Control Room HVAC system automatically isolates on high radiation, and placed in two (3 minutes) train operation 0.1 hr Intermediate time step to facilitate shine dose analysis integration 0.2 hr Intermediate time step to facilitate shine dose analysis integration 0.5 hr Intermediate time step to facilitate shine dose analysis integration I hr Intermediate time step to facilitate shine dose analysis integration 2 hrs
> 99.9 percent of initial containment airborne activity has been released to the environment End of EAB dose analysis 4 hrs Intermediate time step to facilitate shine dose analysis integration 6 hrs Intermediate time step to facilitate shine dose analysis integration 8 hrs CR and LPZ X/Q change LPZ breathing rates changes Control Room HVAC placed in single train operation 10 hrs Intermediate time step to facilitate shine dose analysis integration 24 hrs CR and LPZ X/Q change LPZ breathing rates changes CR occupancy factor changes 96 hrs CR and LPZ X/Q change CR occupancy factor changes 720 hrs End of CR and LPZ dose analyses (a) 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of decay occur prior to the start of the accident (per Assumption 3.2)
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Subject Fuel Handling Accident (FHA) Inside Containment - Control Room & Offsite Doses Sheet 95 of 241 REV ORIGINATOR DATE I
IREI DATE IREV ORIGINATOR DATE I IRE DATE R
4 A. Rustaey lT.
Remick l
l I I
I l
=
=
Jo~~~~
8.7 Containment Shine Dose As discussed in Section 5.4.1, the containment shine dose at distant locations such as the EAB and LPZ are at least two orders of magnitude (a factor of 100) less than the offsite immersion doses due to the fuel handling accident inside containment (FHA-IC) activity release. As calculated in Section 8.10, the gamma immersion dose calculated at the EAB dose receptor is less than 0.3 rem, and the gamma immersion dose calculated at the LPZ dose receptor is less than 0.01 rem. As such, the direct gamma radiation shine dose from the containment to the EAB and LPZ will be less than 0.003 and 0.0001 rem respectively. These doses are negligible, and as such it is not necessary to calculate the containment shine dose at the offsite locations.
The contaminated air inside the Containment Building is a gamma radiation shine source for the Control Room dose receptor. The containment shine dose to the Control Room Operators is dependent on the containment cloud's activity concentration (i.e., the containment cloud's gamma source strengths) and the geometry that separates the cloud from the control room.
Containment shine doses are calculated in the control board area of the control room, at dose points 9, 10, 15 and 16 as shown in Figure 3-1 of Assumption 3.17. Per Assumption 3.17, the maximum dose at these four dose points represents the Control Room Operator dose.
Tables 10.1-1 through 10.1-4 present Quattro-Pro spreadsheet evaluations of the containment shine dose at these locations.
As discussed in Section 5.4.1, the SOURCE2 Code is used to determine the containment cloud instantaneous gamma source strengths. Input to the SOURCE2 Code consists of the BASE IO gamma energy structure, the containment cloud's instantaneous activity loading at various times during the FHA event, and a SOURCE2 Code Multiplier representing one over the containment air dilution volume. The BASE I 0 gamma energy structure is defined in Design Input 4.9. The containment cloud's instantaneous activity loadings (ci) at various times are presented in LOCADOSE Code File "fha-ic.to" (Section 9.2.3), but the loadings modeled in the SOURCE2 Code input file include all of the significant digits contained in LOCADOSE Code File "fha-ic.tm". Per Design Input 4.3. 1, the containment air dilution volume is 1.422e6 ft' with a corresponding inverse volume of 2.483e-I1 cm-3. The SOURCE2 Code input and output files are presented in Section 9.3. 1.
To address a fundamental limitation on the accuracy of the SOURCE2 Code answers (Reference 6.6c, Users Manual Section 2.2.1.2), containment cloud instantaneous gamma source strengths of less than e-25 MeV/cc-sec will be assigned values of 0.0 MeV/cc-sec. This simplification has no impact on the containment shine doses, since the containment shine dose multiplication factors (Design Input 4.10 DMFs) used to scale these instantaneous source strengths are all smaller than I e-9 Rem/hr per MeV/cc-sec. As such, the dose rates associated with the omitted
EUTS DEPARTMENT CALCULATION SHEET ICCN NO.1 PRELIM. CCN NO.
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Sublect Fuel Handlina Accident (FHAI Inside Containment - Control Room & Offsite Doses Sheet 96 of 241 REV ORIGINATOR DATE l
IRE DATE REV ORIGINATOR l DATE IRE DATE R
E 4
A. Rustae T. Remick v
_ _ _ _ _ __ I
_ V I containment cloud instantaneous gamma source strengths of less than e-25 MeV/cc-sec are less than e-34 Rem/hr (= e-25 MeV/cc-sec x e-9 Rem/hr per MeV/cc-sec).
In Tables 10.1-1 through 10.1-4 the containment cloud instantaneous gamma source strengths (MeV/cc-sec) are multiplied by the containment shine dose point 9, 10, 15 or 16 Dose Multiplication Factors (DMFs, Rem/hr per MeV/cc-sec) from Design Input 4.10 to calculate the containment shine contribution to the Control Room Operator whole body gamma dose rate (Rem/hr). As a sample calculation, the Table 0.1-1 containment shine dose rate at dose point 9 at time 10 seconds due to the failure of a high burnup fuel rod is:
Dose Rate = E Dose Rate at AEi
= A, S(AEJ) x DMF(AEi)
Dose Rate (DP 9, 10 sec) = [(2.47e3 MeV/cc-sec) x (1.00e-38 Rem/hr per MeV/cc-sec)]
+ [(4.45e2) x (9.96e-23)] + [(1.24e2) x (2.55e-18)] + [(6.98) x (1.46e-15)]
+ [(6.44e-1) x (2.50e-14)] + [(2.48e-2) x (2.92e-13)] + (2.58e-2) x (9.79e-13)]
+ (4.78e-7) x (2.28e-12)] + [(2.13e-7) x (2.30e-1 1)] + [(0.00) x (1.69e-10)]
Dose Rate (DP 9, 10 sec) = 2.47e-35 Rem/hr + 4.43e-20 + 3.16e-16 + 1.02e-14
+ 1.61e-14 + 7.24e-15 + 2.53e-14 + 1.09e-18 + 4.90e-18 + 0.00 Dose Rate (DP 9, 10 sec) = 5.91e-14 Rem/hr The preceding dose rates calculated with rounded-off input data compare favorably with the Section 10.1 dose rates calculated with all of the significant digits of a spreadsheet.
In Tables 10.1-1 through 10.1-4 the containment shine dose rates are time integrated using Section 5.4.4 methodology to determine the accident duration containment shine dose. As a sample calculation, the Table 10.1-1 containment shine dose at dose point 9 between times 10 seconds and 3 minutes due to the failure of a high burnup fuel rod is:
D = DR(t)dt = CROF (DR2 -DR)
(t2-1)
J I~~~n(DR 2IDR,)(2
)
tI t2 DR(DP 9, t = 10 seconds)
DR(DP 9, t = 3 minutes)
CROF(t < 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />)
= 10 seconds = 0.00278 hours
= 3 minutes = 0.05 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />
= 5.91e-14 Rem/hr
= S.01 e-14 Rern/hr
= 1.0 (per Table 10.1-1)
(per Table 10.1-1)
(per Design Input 4.16)
E&TS DEPARTMENT CALCULATION SHEET ICCN NOJ PRELIM. CCN NO.
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Riihicrt Fil Wondlinn Artnirlant XFWAI Insidel Cnntainment - Cnntrol Rnnm P. Offcitp Dlses Sheet 07 f 21 VJ---
l mi *Cl-
- v Abaus
- t l
ng*
@lWEE*,
Ad-
- tW
-=.
SV - l-- -,-Vw
_f REV ORIGINATOR DATE IRE DATE REV ORIGINATOR DATE IRE DATE R
4 A. Rustae I
T. Remick I
I:~~~~~~~~~~~~~~~~~~~~
D( 0 sec to 3 min) = [1.0 x 5[01e0-14 Rem.hr -
91e-14 Remlhr x 105 hr- 000278 h]
S.01e-14) 5.91e-14 D(DP 9, 10 sec to 3 min) = 2.57e -15 Rem The preceding dose calculated with rounded-off input data compares favorably with the Section 10.1 dose calculated with all of the significant digits of a spreadsheet.
Table 8.7-1 summarizes the 30-day Control Room Operator doses calculated in Tables 10.1-1 through 10.1-4. Per Table 8.7-1, the maximum containment shine dose within the control board area of the control room occurs at dose point 10. A review of Tables 10.1-1 through 10.14 reveals that virtually all of the containment shine dose to Control Room Operators occurs within the first I hour of the fuel handling accident (by which time the majority of the containment airborne source has been released to the environment).
TABLE 8.7-1 Containment Shine Doses Dose Point Containment Shine 30-day Control Room Dose (Rem) 9 1.69e-14 10 2.19e-14 15 2.02e-14 16 1.43e-14
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Subject Fuel Handling Accident (FHA) Inside Containment - Control Room & Offsite Doses Sheet 98 of 241 REV ORIGINATOR DATE IRE DATE REV ORIGINATOR DATE IRE I
R I+/-A.Iustae1 T.
Remick
______DT I
- H~~
8.8 Environmental Cloud Shine Dose The contaminated air in the environmental cloud outside of the control room envelope is a gamma radiation shine source for the Control Room dose receptor (the environmental cloud dose contributions to the offsite EAB and LPZ dose receptors are considered in the offsite immersion dose). The environmental cloud shine dose to the Control Room Operators is dependent on the environmental cloud's activity concentration (i.e., the environmental cloud's gamma source strengths) and the geometry that separates the cloud from the control room.
Environmental cloud shine doses are calculated in the control board area of the control room, at dose points 9, 10, 15 and 16 as shown in Figure 3-1 of Assumption 3.17. Per Assumption 3.17, the maximum dose at these four dose points represents the Control Room Operator dose.
Tables 10.2-1 through 10.2-4 present Quattro-Pro spreadsheet evaluations of the containment shine dose at these locations.
As discussed in Section 5.4.2, the SOURCE2 Code is used to determine the containment cloud instantaneous gamma source strengths. The SOURCE2 Code runs determining the containment cloud instantaneous gamma source strengths are the same runs described in the Section 8.7 evaluation of containment shine dose. The SOURCE2 Code input and output files are presented in Section 9.3.1.
To address a fundamental limitation on the accuracy of the SOURCE2 Code answers (Reference 6.6c, Users Manual Section 2.2.1.2), cloud instantaneous gamma source strengths of less than I e-25 MeV/cc-sec will be assigned values of 0.0 MeV/cc-sec. This simplification has no impact on the environmental cloud shine doses, since the environmental cloud shine dose multiplication factors (Design Input 4.11 DMFs) used to scale these instantaneous source strengths are all smaller than I e-3 Rem/hr per MeV/cc-sec. As such, the dose rates associated with the omitted environmental cloud instantaneous gamma source strengths of less than e-25 MeV/cc-sec are less than e-28 Rem./hr (= le-25 MeV/cc-sec x le-3 Rem/hr per MeV/cc-sec).
Tables 10.2-1 through 10.2-4 determine the environmental cloud instantaneous gamma source strengths by scaling the SOURCE2 Code containment cloud instantaneous gamma source strengths by the containment release rate (thereby determining the source tern rclcasc rate to the environment), and then by the control room atmospheric dispersion factor (thereby determining the environmental cloud instantaneous gamma source strengths at the control room HVAC intake).
NOTE: Tables 10.2-1 through 10.2-4 present instantaneous source strength data at 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, and 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> in two columns, one labeled with a negative sign ("-" representing just prior to the stated time), and one labeled with a positive sign ("+" representing just after the
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4 A. Rustaey T. Remick I
E I
stated time). Although the containment cloud instantaneous source strength data at each pair of times is identical, the environmental cloud instantaneous source strengths will vary at the stated time due to the control room atmospheric dispersion factor scaling.
As a sample calculation, the Table 10.2-1 environmental cloud instantaneous source strength for BASEI0 energy range of 0.0 to 0.1 MeV, at time '8-' hours is:
e()= Sc()x Fx X()
S, (at 8-hrs, 0.0 to 0.1 MeV) = 2.28e-9 MeV/cc-sec F
= 38.7 m3/sec X/Q(0 to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />)
= 3.1e-3 sec/r 3 (per "ss-c.so" [Section 9.3.2])
(per Design Input 4.3.2)
(per Design Input 4.7)
I I
S,(at 8-hrs, 0.0 to 0.1 MeV) = (2.28e-9 MeV/cc-sec) x (38.7 m 3/sec) x (3.1e-3 sec/m 3)
S,(at 8-hrs, 0.0 to 0.1 MeV) = 2.74e-10 MeV/cc-sec The preceding source strength calculated with rounded-off input data compares favorably with the Section 10.2 source strength calculated with all of the significant digits of a spreadsheet.
In Tables 10.2-1 through 10.2-4 the environmental cloud instantaneous gamma source strengths (MeV/cc-sec) are multiplied by the environmental cloud shine dose point 9, 10, 15 or 16 Dose Multiplication Factors (DMFs, Rem/hr per MeV/cc-sec) from Design Input 4.11 to calculate the envirom-nental cloud shine contribution to the Control Room Operator whole body gamma dose rate (Rem/hr). As a sample calculation, the Table 10.2-1 environmental cloud shine dose rate at dose point 9 at time 10 seconds is:
Dose Rate = I Dose Rate at AEi
= A, S(AE) x DMF(AEi)
Dose Rate (DP 9, 10 sec) = [(2.96e+2 MeV/cc-sec) x (5.22e-5 Rem/hr per MeV/cc-sec)]
+ [(5.34e+1) x (8.90e-5)] + [(1.48e+l) x (9.26e-5)] + [(8.37e-1) x (1.03e-4)]
+ [(7.73e-2) x (.lI e-4)] + (2.98e-3) x (1.21c-4)] + [(3.09e-3) x (1.27e-4)]
+ [(5.73e-8) x (1.32e-4)] + [(2.56e-8) x (1.48e-4)] + [(O.00) x (1.75e-4)]
Dose Rate (DP 9, 10 sec) = 1.55e-2 Rem/hr + 4.75e-3 + 1.37e-3 + 8.62e-5
+ 8.50e-6 + 3.61 e-7 + 3.92e-7 + 7.56e-12 + 3.79e-12 + 0.00 I
I I
Dose Rate (DP 9, 10 sec) = 2.17e-2 Rem/hr
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Subject Fuel Handling Accident (FHAI Inside Containment - Control Room & Offsite Doses Sheet 100 of 241 IREV ORIGINATOR DATE IRE DATE REV ORIGINATOR DATEI IRE DATEI RI 4 ARuste T. Remick j
Vm 4~~~~~~~
The preceding dose rates calculated with rounded-off input data compare favorably with the Section 10.2 dose rates calculated with all of the significant digits of a spreadsheet.
In Tables 10.2-1 through 10.24 the environmental cloud shine dose rates are time integrated using Section 5.4.4 methodology to determine the accident duration environmental cloud shine dose. As a sample calculation, the Table 10.2-1 envirornental cloud shine dose at dose point 9 between times 10 seconds and 3 minutes is:
= CROF X (DR2 -DR)
D = DRtdt CO 2_
J
~~~In(DR 2IDR1) ti t2 DR(DP 9, t, = 10 seconds)
DR(DP 9, t2 = 3 minutes)
CROF(t < 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />)
= 10 seconds = 0.00278 hours
= 3 minutes = 0.05 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />
= 2.17e-2 Rem/hr (per Table 10.2-1)
= 1.84e-2 Rem/hr (per Table 10.2-1)
= 1.0 (per Design Input 4.6)
D(10 sec to 3 min) = [1.01 x 184e -2 Rem*hr -2.17e 2 Rem/hrI x [0.05 hr -0.00278 hrl 1 84e -2 2.17e -2 D(DP 9, 10 sec to 3 miin) = 9.45e-4 Rem The preceding dose calculated with rounded-off input data compares favorably with the Section 10.2 dose calculated with all of the significant digits of a spreadsheet.
Table 8.8-1 summarizes the 30-day Control Room Operator doses calculated in Tables 10.2-1 through 10.2-4. Per Table 8.8-1, the maximum environmental cloud shine dose within the control board area of the control room occurs at dose point 10. A review of Tables 10.2-1 through 10.2-4 reveals that virtually all of the environmental cloud shine dose to Control Room Operators occurs within the first I hour of the fuel handling accident (by which time the majority of the containment airborne source has been released to the environniut).
E&TS DEPARTMENT CALCULATION SHEET PRELIM. CCN NC Project or DCP/FCN/ECP Caic. No. N
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Sheet 101 of 241 TABLE 8.8-1 Environmental Cloud Shine Doses Dose Point Environmental Cloud Shine 30-day Control Room Dose (Rem) 9 6.29e-03 10 7.75e-03 15 6.92e-03 16 6.Ole-03
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Subject Fuel Handling Accident (FHA Inside Containment - Control Room & Offsite Doses Sheet 102 of 241 REVI ORIGINATOR DATE IRE DATE REV ORIGINATOR DATE IRE DATE R
4 A. Rustaey T. Remick
___Ima I ii 8.9 Control Room HVAC Filter Shine Dose Upon receipt of a Control Room Isolation Signal the control room HVAC system is automatically shifted to the emergency mode of operation. Transfer to the emergency mode includes starting the emergency air conditioning units, opening the outside air isolation damper to the emergency filtration trains, and starting the fans. Per Design Input 4.4.3, this analysis models two trains of emergency HVAC in operation during the first 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> of the accident.
Operator action is assumed to be taken within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> to deactivate one train of emergency intake and recirculation units.
Per the LOCADOSE Code Activity Transport Output File (Section 9.2.3) virtually all of the containment airborne source is released from the containment during the first 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> of the event. Therefore, although one of the two filtered control room intake flow paths will be isolated at 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, the activity introduced and retained within each filter (i.e., the filter loading) will be identical.
The contaminated iodine and particulate isotopes retained on control room HVAC intake filters SA1510MA206 and SA1510MA207 (A206 and A207) and control room recirculation filters SA151OME418 and SA 15OME419 (E418 and E419) are gamma radiation shine sources for the Control Room dose receptor. As noted in Section 5.4.3, for the purpose of determining the shine dose contribution from these filters, intake filters A206 and A207 are assumed to be 100%
efficient at removing iodine and particulates from the incoming air. This maximizes the amount of iodine and particulates retained on A206 and A207, and thus maximizes the shine dose from these two filters. In reality, iodine and particulates that are not trapped on the intake filters will eventually be trapped on recirculation filters E418 and E419, which are located in the vicinity of the intake filters. However, the results of the filter shine dose calculation would not be significantly different, since the geometry of the direct shine pathways from E418 and E419 is similar to the geometry of the direct shine pathways from A206 and A207.
The control room HVAC filter shine dose to the Control Room Operators is dependent on the filter activity concentration (i.e., the filters' gamma source strengths) and the geometry that separates the filters from the control room.
Control room filter shine doses are calculated in the control board area of the control room, at dose points 9, 10, 15 and 16 as shown in Figure 3-1 of Assumption 3.17. Per Assumption 3.17, the maximum dose at these four dose points represents the Control Room Operator dose.
Tables 10.3-1 through 10.3-4 present Quattro-Pro spreadsheet evaluations of the control room filter shine dose at these locations.
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Q1 hhiatf Ca gl W:nnlin Aqitnf WF1ADI Ineield e'r.nntinmanf - C'^ntrrnl nnm &
fflcit ndsa Qhme ar nqf IA4 REV ORIGINATOR DATE IRE DATE REV ORIGINATOR DATE IRE DATE R
4 A. Rustae T. Remick E
As discussed in Section 5.4.3, the SOURCE2 Code is used to determine the control room HVAC filter instantaneous gamma source strengths. Input to the SOURCE2 Code consists of the BASElO gamma energy structure, a control room HVAC filter instantaneous activity lading at various times during the FHA event, and a SOURCE2 Code Multiplier representing one over the CR HVAC intake filter volume. The BASE10 gamma energy structure is defined in Design Input 4.9. Each filter's instantaneous activity loadings (ci) at various times are presented in LOCADOSE Code File "fha-ic.to" (Section 9.2.3), but the loadings modeled in the SOURCE2 Code input file include all of the significant digits contained in LOCADOSE Code File "fha-ic.trn". Since noble gas activity will not be collected on the CR HVAC intake filter, the SOURCE2 Code input file used to determine instantaneous gamma source strength spectra for the filter shine analysis will zero out the CR HVAC intake filter noble gas activity values. Per Assumption 3.13, the volume of each CR HVAC intake filter is 7.32 W3 with a corresponding inverse volume of 4.824e-6 cm-3. The SOURCE2 Code input and output files for the case with a high bumup fuel rod failure are presented in Section 9.3.2.
To address a fundamental limitation on the accuracy of the SOURCE2 Code answers (Reference 6.6c, Users Manual Section 2.2.1.2), filter instantaneous gamma source strengths of less than I e-25 MeV/cc-sec will be assigned values of 0.0 MeV/cc-sec. This simplification has no impact on the control room HVAC filter shine doses, since the control room filter shine dose multiplication factors (Design Input 4.12 DMFs) used to scale these instantaneous source strengths are all smaller than e-8 Rem/hr per MeV/cc-sec. As such, the dose rates associated with the omitted filter instantaneous gamma source strengths of less than e-25 MeV/cc-sec are less than 1 e-33 Rem/hr (= l e-25 MeV/cc-sec x e-8 Rem/hr per MeV/cc-sec).
In Tables 10.3-1 through 10.3-4 the control room HVAC filter instantaneous gamma source strengths (MeV/cc-sec) are multiplied by the control room HVAC filter A206 and A207 shine dose point 9, 10, 15 or 16 Dose Multiplication Factors (DMFs, Rem/hr per MeV/cc-sec) from Design Input 4.12 to calculate the control room HVAC filter A206 and A207 shine contributions to the Control Room Operator whole body gamma dose rate (Remn/hr). The filter A206 and A207 shine dose rates are summed together to determine the total control room HVAC filter shine dose rates. As a sample calculation, the Table 10.3-1 control room HVAC filter shine dose rate at dose point 9 at time 0.1 hours1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> is:
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Subject Fuel Handling Accident (FHA) Inside Containment - Control Room & Offsite Doses Sheet 104 of 241
_R V O R G N A_
REV ORIGINATOR DATE IRE DATE REV ORIGINATOR DATE IRE DATE R
4 A. Rustaey T. Remick IE
=
X~~~~~
FILTER A206 DOSE RATE (DP 9. 0.1 hours1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />)
Filter A206 Dose Rate = £ A206 Dose Rate at AEi = S S(AEi) x A206 DMF(AEj)
Filter A206 Dose Rate = [(8.67e+1 MeV/cc-sec) x (2.19e-10 Rern/hr per MeV/cc-sec)]
+ (1.31e+4) x (1.74e-9)] + [(7.83e+3) x (2.02e-9)] + [(5.83e+2) x (2.09e-9]
+ [(5.40e+1) x (2.05e-9] + (2.03) x (2.02e-9)] + [(2.06) x (1.99e-9)]
+ (9.29e-4) x (1.95e-9)] + [(4.14e-4) x (1.78e-9] + [(0.00) x (1.60e-9)]
Filter A206 Dose Rate = 1.90e-8 Rern/hr + 2.28e-5 + 1.58e-5 + 1.22e-6
+ 1.1 e-7 + 4.10e-9 + 4.lOe-9 + 1.81le-12 + 7.37e-13 + 0.00 Filter A206 Dose Rate = 4.00e-5 Renhr FILTER A207 DOSE RATE (DP 9. 0.1 hours1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />)
Filter A207 Dose Rate = F A207 Dose Rate at AEj = S S(AE1) x A207 DMF(AE)
Filter A207 Dose Rate = [(8.67e+1 MeVlcc-sec) x (1.1 le-1 0 Rem/hr per MeV/cc-sec)]
+ [(1.3 e+4) x (1.34e-9)] + [(7.83e+3) x (1.74e-9)] + [(5.83e+2) x (1.93e-9]
+ (5.40e+l) x (1.95e-91 + [(2.03) x (1.97e-9)) + [(2.06) x (1.96e-9)]
+ (9.29e-4) x (1.93e-9)] + (4.14e-4) x (1.80e-9] + [(0.00) x (1.64e-9)]
Filter A207 Dose Rate = 9.62e-9 Rem/hr + 1.76e-5 + 1.36e-5 + 1.13e-6
+ 1.05e-7 + 4.00e-9 + 4.04e-9 + 1.79e-12 + 7.45e-13 + 0.00 Filter A207 Dose Rate = 3.25e-5 Rem/hr TOTAL DOSE RATE (DP 9.0.1 hours1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />)
Total Dose Rate
= Filter A206 Dose Rate + Filter A207 Dose Rate
= 4.00e-S Resnhr + 3.25e-5 Rem/hr
= 7.25e-5 Rem/hr The preceding dose rates calculated with rounded-off input data compare favorably with the Section 10.3 dose rates calculated with all of the significant digits of a spreadsheet.
I
E&TS DEPARTMENT CALCULATION SHEET ICCN NO.J PRELIM. CCN NO.
PAGE _
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Project or DCP/FCN/ECP Caic. No. N4072-003 CCN CONVERSION:
CCN NO. CCN -
9uhipnt Fe Hnnrilinn Accident FHAI Insde Cnntainment - Cnntrnl Rnnm. (MffitR DATes I Sqheet i0 os f 241 REVI ORIGINATOR DAT E
I DATE l REVI ORIGINATOR DARl IRE l
DATEl R TEJ 4 A. Rustae T. Remick I
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=
{
In Tables 10.3-1 through 10.3-4 the control room HVAC filter shine dose rates are time integrated using Section 5.4.4 methodology to determine the accident duration control room HVAC filter shine dose. As a sample calculation, the Table 10.3-1 control room HVAC filter shine dose at dose point 9 between times 0.1 hours1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and 0.2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />s:
D = DR(t)dt = CROF (DR. -DR)
(t2-11)
I t,
t2 DR(DP 9, t, = 0.1 hours1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />)
DR(DP 9, t2 = 0.2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />)
CROF(t < 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />)
= 0.1 hours1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />
= 0.2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />
= 7.22e-5 Rem/hr
= 1.84e-4 Rem/hr
=1.0 (per Table 10.3-1)
(per Table 10.3-1)
(per Design Input 4.16)
D(O.1 hr to 0.2 hr) = 11.0 X 1.84e-4 Remlhr - 7.22e-5 Rem)hr x [0.2 hr -0.1 hr) 1 84e-4 7.22e-5 D(DP 9, 0.1 hr to 0.2 hr) = 1.20e-5 Rem The preceding dose calculated with rounded-off input data compares favorably with the Section 10.3 dose calculated with all of the significant digits of a spreadsheet.
Table 8.9-1 summarizes the 30-day Control Room Operator doses calculated in Tables 10.3-1 through 10.3-4. Per Table 8.9-1, the maximum control room HVAC filter shine dose within the control board area of the control room occurs at dose point 10. A review of Tables 10.3-1 through 10.34 reveals that the control room HVAC filter shine dose contributions to Control Room Operators occur continuously during the 30 day event duration (primarily due to the long-lived isotopes that are retained on the filters).
E&TS DEPARTMENT ICUN NOJ CALCULATION SHEET PRELIM. CCN N Project or DCPFCNECP Caic. No. N4072-003 CC Subject Fuel Handling Accident (FHA) Inside Containment - Control Room & Offsite Doses Sheet 106 of 241 REV ORIGINATOR DATE l
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E 4 A. Rustaey T. Remick 1V
_____L..mi j
im{i...m~j I TABLE 8.9-1 Control Room HVAC Filter Shine Doses Dose Point Control Room HVAC Filter Shine 30-day Control Room Dose (Rem) 9 4.42e-02 10 6.28e-02 15 4.20e-02 16 4.83e-02 I
Note that the filter shine values given in the above table only included HVAC 2 train filtered flow of 4400 cfin for 117 minutes but did not include 0-3 minute recirculation flow and any unfiltered control room inleakage flow. Therefore the above numbers may be scaled up using ratio of values given in Table 8.5-1 without the need for reanalysis; For 10 CFM unfiltered CR inleakage:
For 1000 CFM unfiltered CR inleakage:
533,460/514,800 = 1.04 652,260/514,800 = 1.27 TABLE 8.9-la Revised Control Room HVAC Filter Shine Doses 30-day Control Room Dose (Rem)
Normal Intake Normal Intake Dose Point CR HVAC only 10 CFM CR inleakage 1000 CFMCR inleakage 9
4.42e-02 4.60e-02 5.6le-02 10 6.28e-02 6.53e-02 7.98e-02 15 4.20e-02 4.37e-02 5.33e-02 16 4.83e-02 5.02e-02 6.13e-02 I
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E&TS DEPARTMENT CALCULATION SHEET IPRELIM. CCN N1 Project or DCP/FCNIECP Calc. No. N-4072-003 cc Subiect Fuel Handlina Accident (FHA) Inside Containment - Control Room & Offsite Doses Sheet 107 of 241 REV ORIGINATOR DATE l
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DATE REV ORIGINATOR DATE IRE DATE 4 A. Rustay T. Remick
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4 8.10 Summation of Offsite and Control Room Doses Table 8.10-1 sums the offsite and control room doses due to a design basis Fuel Handling Accident Inside Containment (FHA-IC). Per Design Input 4.1.3, in the event of an FHA-IC a total of 226 fuel rods will fail, representing 16 fuel rods in the dropped fuel bundle, and 210 fuel rods in one or two impacted fuel bundles. The dropped and impacted fuel assemblies may contain either once burned fuel or high bumup fuel. Per Design Input 4. 1, once burned fuel has been irradiated for a single fuel cycle, and high burnup fuel has been irradiated for more than one fuel cycle.
The doses presented in Table 8.10-1 include:
The offsite and control room inhalation and immersion doses that are extracted from the LOCADOSE code output of Section 9.2.5.
The containment gamma radiation shine dose to control room personnel as summarized in Table 8.7-1 for dose point 10. Per Section 8.7, the containment shine to offsite dose receptors is negligible.
The environmental cloud gamma radiation shine dose to control room personnel as summarized in Table 8.8-1 for dose point 10. Per Section 8.8, the environmental cloud shine to offsite dose receptors is considered in the offsite immersion dose.
The control room HVAC filter gamma radiation shine dose to control room personnel as summarized in Table 8.9-1 for dose point 10.
Of note is that:
(1) the containment direct shine dose to both the control room and offsite dose receptors is negligible (2)
The sum of the environmental cloud gamma radiation shine dose and the control room HVAC filter gamma radiation shine dose is approximately 50 percent of the gamma immersion dose [actually 50.4 percent = (7.75e-3 + 7.98e-2) / 0.1736 ].
I As such, this fuel handling accident calculation could have been conservatively simplified if it had been assumed that the sum of the environmental cloud gamma radiation shine dose and the control room HVAC filter gamma radiation shine dose was equal to the gamma immersion dose.
E&TS DEPARTMENT CALCULATION SHEET ICCN NOJ PRELIM. CCN NO.
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Siihipt Fel 14.qnrilinn Acrident (FHAI Incir Cnntainment - ContrnI Rnnm & ffsitp Dne Sheet 108 nf 21 V-Jv-T-
- I V-#MA w
f --
- -. TV REV
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-1 REV ORIGINATOR DATE IRE DATE REV ORIGINATOR DATE I
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II Table 8.10-1 FHA Inside Containment Doses ()
Dose Contributor Event Duration 2 Hour Event Duration (Calculation Section)
Control Room Doses EAB Doses LPZ Doses
[CR isolated at 3 min] (Rem)
(Rem)
(Rem)
IOCFM IOO CFM Inleakage Inleakage THYROID DOSE:
Inhalation Dose (9.2.5) 1.483e+0I 2.534e+01 S.640e+01 1.602e+00 BETA SKIN DOSE:
Immersion Dose (9.2.5) 3.616e+00 3.591e+00 3.117e-01 8.856e-03 WHOLE BODY GAMMA DOSE:
Immersion Dose (9.2.5) 1.736e-01 1.726e-01 2.694e-01 7.655e-03 Containment Direct Shine Dose (8.7) 2.19e-14 2.19e-14 Negligible Negligible Environmental Cloud Shine Dose (8.8) 7.75e-03 7.75e-03 Included in Included in Immersion Dose Immersion Dose CR HVAC Filter Shine Dose (8.9) 6.53e-02 7.98e-02 Not ApDplicable Not Avvlicable TOTAL WHOLE BODY GAMMA DOSE 2.470e-01 2.610e-01 2.694e-01 7.655e-03 (a)
The doses presented in this table are based on an FHA-IC in which a total of 226 fuel rods will fail, representing 16 fuel rods in the dropped fuel bundle, and 210 fuel rods in one or two impacted fuel bundles.
The analysis models the 16 fuel rod failure portion with a 12 percent high burnup fuel rod iodine gap inventory and a radial peaking factor of 1.71. The analysis models the 210 fuel rod failure portion with a 12 percent high bumup fuel rod iodine gap inventory and an assembly averaged relative power density of 1.37.
In the event that the FHA-IC is defined by a different fuel rod damage scenario, then the results presented in this table must be reviewed for applicability.