RS-09-078, Response to Request for Additional Information Related to Technical Specification 3.3.6.1, Primary Containment and Drywell Isolation Instrumentation

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Response to Request for Additional Information Related to Technical Specification 3.3.6.1, Primary Containment and Drywell Isolation Instrumentation
ML091740100
Person / Time
Site: Clinton Constellation icon.png
Issue date: 06/23/2009
From: Hansen J
Exelon Generation Co, Exelon Nuclear
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
RS-09-078, TAC ME1499 C-023, Rev 2
Download: ML091740100 (44)
v • d • e


Text

Exelon Nuclear wwwexeloncorp.com 4300 Winfield Road Nuclear Warrenville, IL 60555 10 CFR 50.90 RS-09-078 June 23,2009 Attn: Document Control Desk U. S. Nuclear Regulatory Commission Washington, DC 20555-0001 Clinton Power Station, Unit 1 Facility Operating License No. NPF-62 NRC Docket No. 50-461

Subject:

Response to Request for Additional Information Related to Technical Specification 3.3.6.1, "Primary Containment and Drywell Isolation Instrumentation" (TAC No. ME1499)

References:

1. Letter from P. R. Simpson (Exelon Generation Company, LLC) to U. S. NRC, "Request for Amendment to Technical Specifications Section 3.3.6.1, 'Primary Containment and Drywellisolation Instrumentation'to Eliminate Requirements for Main Steam Line Isolation on High Turbine Building Temperature," dated June 15, 2009
2. Letter from J. L. Hansen (Exelon Generation Company, LLC) to U. S. NRC, "Request for Processing an Amendment to Technical Specifications Section 3.3.6.1, 'Primary Containment and Drywell Isolation Instrumentation,' to Eliminate Requirements for Main Steam Line Isolation on High Turbine Building Temperature on an Emergency Basis," dated June 20,2009
3. Email from Marshall David (U. S. NRC) to Mitchel A. Mathews (Exelon Generation Company, LLC), "Clinton Emergency LAR RAI (TAC ME1499), dated June 22,2009 In the Reference 1, Exelon Generation Company, LLC (EGC), requested a change to the Technical Specifications (TS) of Facility Operating License No. NPF-62 for Clinton Power Station (CPS), Unit 1. The proposed change was requested to eliminate the requirement for main steam line (MSL) isolation on high Turbine Building temperatures from TS Section 3.3.6.1, "Primary Containment and Drywellisolation Instrumentation,"

Table 3.3.6.1-1 (l.e., Function 1.f).

In Reference 2, on June 20,2009, EGC requested the review of Reference 1 be performed on an emergency basis.

June 23, 2009 U. S. Nuclear Regulatory Commission Page 2 During the NRC's review of the Reference 1, the NRC found that additional information was required to support its review. This request was received by EGC in Reference 3 on June 22,2009. The requested information is provided in Attachment 1 to this letter.

The information provided in this letter does not affect the No Significant Hazards Consideration, or the Environmental Consideration provided in Attachment 1 of the original license amendment request as described in the referenced submittal.

In accordance with 10 CFR 50.91(b), "State consultation," EGC is providing the State of Illinois with a copy of this letter and its attachment to the designated State Official.

There are no regulatory commitments contained within this letter. Should you have any questions concerning this letter, or require additional information, please contact Mr.

Mitchel Mathews at (630) 657-2819.

I declare under penalty of perjury that the foregoing is true and correct. Executed on the 23rd day of June, 2009.

Respectfully, Jeff y L. Hansen Manager - Licensing Exelon Generation Company, LLC Attachments:

1. Response to NRC Requests for Additional Information
2. Calculation C-023, Revision 2, "Re-Analysis of Main Steam Line Break (MSLB) using Alternative Source Terms"

Attachment 1 Response to NRC Requests for Additional Information Request No.1.

By letter dated June 15, 2009, (Agencywide Documents Access and Management System (ADAMS) Accession No. ML0916605802), supplemented by letter dated June 20, 2009, Exelon Generation Company, LLC (EGC), submitted a license amendment request regarding proposed changes to the technical specifications (TSs). The proposed amendment would eliminate the requirement for main steam line isolations on high Turbine Building temperatures from TS Section 3.3.6.1, "Primary Containment and Drywellisolation Instrumentation," Table 3.3.6 .1-1,

{i.e., Function 1.t). The Nuclear Regulatory Commission (NRC) staff is reviewing the submittal and has determined that the following additional information is needed to complete its review.

The proposed changes in the letter dated June 15, 2009, state that "No credit is taken in the transient or accident analysis for the automatic isolation of the MSIVs by these Turbine Building area temperature switches. Thus, the Turbine Building temperature switches are not assumed to function to mitigate any accident described in Chapters 6 or 15 of the CPS USAR. " However, the purpose of the limiting condition for operation (LCD) for primary containment isolation instrumentation is to satisfy Title 10 of the Code of Federal Regulations (10 CFR)

Section 50.36, criterion 2 and 3, which establish, in part, an operating restriction that is an initial condition of a design basis accident (DBA) or a structure which is part of the primary success path and which functions or actuates to mitigate a design basis accident.

The primary containment and drywell isolation instrumentation automatically initiates closure of appropriate primary containment isolation valves (PCIVs) and drywell isolation valves. The function of the PCIVs, in combination with other accident mitigation systems, is to limit fission product release during and following postulated DBAs. Primary containment isolation within the time limits specified for those isolation valves designed to close automatically ensures that the release of radioactive material to the environment will be consistent with the assumptions used in the analyses for a DBA. The proposed changes in the letter dated June 15, 2009 state that "Ambient Temperature-High is provided to detect a leak in the RCPB, and provides diversity to the high flow instrumentation. The isolation occurs when a very small leak has occurred. If the small leak is allowed to continue without isolation, offsite dose limits may be reached." The monitoring of the turbine building temperatures serves to insure a certain level of boundary integrity consistent with that assumed in the licensing bases accident analyses. Levels of leakage from the accident boundary beyond those currently controlled by Function 1.f are not modeled in typical accident analyses. In fact, no leakage is typically assumed, with the exception of MSIV leakage, for SSE (safe shutdown earthquake) qualified pipe and condensers.

Page 1 of 9

Attachment 1 Response to NRC Requests for Additional Information Per the USNRC staff's letter entitled, IISafety Evaluation Input for Alternative Source Term Full Implementation and Related Technical Specification Changes for Clinton Power Station, Unit 1 (TAC NO. MB8365)," dated September 19, 2005 it appears that no leakage, other than MSIV leakage is assumed for the MSIV leakage pathway. For the loss of coolant accident, integrity of piping up to the turbine/auxiliary building (secondary containment) wall is credited. For the control rod drop accident, piping integrity up to and including the condenser is credited.

Please provide a complete assessment of all design basis accidents in the Clinton licensing bases impacted by the proposed change that assume integrity of the MSIV leakage pathway boundary (for example those DBAs that assume the integrity of the MSIV leakage pathway as: 1) an initial condition of a DBA or 2) a structure which is part of the primary success path, or which functions or actuates to mitigate a design basis accident. In the assessment, provide each design basis accident impacted by the proposed change and a justification why the currently assumed level of integrity assumed in the current licensing bases is no longer required.

Response to Request No.1 At Clinton Power Station (CPS), the outboard Main Steam Isolation Valves (MSIVs) are located within the Auxiliary Building steam tunnel. The temperatures in this area will continue to be monitored by the instruments used in maintaining the function described in CPS (Technical Specifications (TS) Section 3.3.6.1, "Primary Containment and Drywellisolation Instrumentation," Table 3.3.6.1-1, (i.e., Function 1.e, "Main Steam Tunnel Temperature"). EGC is not proposing to modify the main steam line isolation associated with these instruments, which are located in the Auxiliary Building Steam tunnel.

The three bounding accidents evaluated for CPS using Alternative Source Term methodology per Regulatory Guide 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors," and involving a release of radioactivity through the steam tunnel are:

1. Main Steam Line Break outside containment,
2. Design Basis Accident (DBA) Loss of Coolant Accident (LOCA) (Reactor Recirculation line break inside containment), and
3. Control Rod Drop Accident.

For the Steam Line Break outside containment, leakage occurs outside primary containment at a location downstream of the outermost MSIV. Closure of the MSIVs terminates the leakage when full closure is reached (i.e., within 5.5 seconds). The Page 2 of 9

Attachment 1 Response to NRC Requests for Additional Information MSLB release assumptions on the amount and pathway are unchanged by the level of integrity of the Main Steam piping downstream of the MSIVs.

For the DBA LOCA, integrity of the safety and non-safety piping to the secondary containment boundary (Le., at the wall between the Auxiliary Building and Turbine Building) is assumed. Main steam piping downstream of the MSIVs in the Auxiliary Building steam tunnel is credited for aerosol removal. That piping is capable of performing its safety function during and following a Safe Shutdown Earthquake. No credit is taken for hold-up or dilution for piping downstream of this, in the condenser, or Turbine Building. Any lack of integrity during the DBA LOCA will be detected by the temperature instruments located within the Auxiliary Building steam tunnel which are not being modified by the proposed change.

Following a Control Rod Drop Accident (CRDA), radioisotopes postulated to be released will be transported through the MSL directly to the main steam condenser.

Two cases are analyzed:

For Case 1, the activity is assumed to leak from the condenser at a rate of 1% per day, in accordance with RG 1.183. Releases from the Turbine/Condenser are assumed to be from the plant vent without credit for dilution or holdup in the Turbine Building.

For Case 2, with the Steam Jet Air Ejector (SJAE) in operation, the activity is released through a system of Charcoal Delay Beds, where Iodine and particulates are effectively removed. Therefore, the only contributor to dose resulting from this pathway is the delayed release of noble gas nuclides.

The Mechanical Vacuum Pump is assumed to shut down before any unfiltered material from the condenser can be released through its pathway due to system isolation initiated by the MSL Radiation Monitor. No changes to this assumption or any other additional unmonitored pathways are introduced as a results of this request. All leakage from the main steam turbine condenser leaks to the atmosphere at a vent release rate of 1% per day, for a period of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. A smaller, 25 gpm leak at the operating pressure of 962 psia does not constitute an additional open pathway at post accident conditions in the Main Condenser since the CRDA is terminated by the Average Power Range Monitor (APRM) 120% power signal scram.

In summary, for the design basis accidents potentially impacted by the proposed change, the current licensing basis analyses will remain valid following the implementation of the proposed change including assumptions regarding levels of integrity.

Page 3 of 9

Attachment 1 Response to NRC Requests for Additional Information Request No.2.

Your letter dated June 15, 2009, states that "No credit is taken in the transient or accident analysis for the automatic isolation of the MSIVs by these Turbine Building area temperature switches. Thus, the Turbine Building temperature switches are not assumed to function to mitigate any accident described in Chapters 6 or 15 of the CPS USAR." The letter also states that liAs stated in the CPS TS Bases, credit is not taken for the Turbine Building temperature instruments in any transient or accident analysis in the USAR, since bounding analyses are performed for large breaks such [as] a MSLBs which are isolated by other MSL leakage detection methods as discussed in Section 3.0. Moreover, this trip is not assumed to mitigate the consequences of a design basis accident (DBA) or transient, and is not input into the assumptions for any DBA analysis."

While the bounding analyses were performed for large breaks, such as a MSLB, the current analyses should bound the spectrum of steam line breaks that extend the boundaries of the MSL. With the deletion of the automatic isolation of the MSIVs by these Turbine Building area temperature switches, justify why the large break MSLB will continue to be limiting and provide an assessment of the spectrum high energy line breaks, as defined by your Updated Final Safety Analysis Report, orjustify why such an assessment is not necessary. Provide necessary information so that the DBA can be modeled by the NRC and justification for the assumptions used.

Additional questions:

Are the remaining instruments safety related?

How they are credited to mitigate the consequences of a steam break?

What is the timing of the credited systems?

Are the credited systems automatic or manual?

What is the release from the steam line prior to isolation?

What are the radiological consequences?

Response to Request No.2:

The dose calculation for the bounding Main Steam Line Break (MSLB) outside containment is calculation C-023 Revision 2, "Re-Analysis of Main Steam Une Break (MSLB) using Alternative Source Terms" (Le., Attachment 2). This calculation takes no credit for the presence of the Page 4 of 9

Attachment 1 Response to NRC Requests for Additional Information Turbine Building, but releases all discharge immediately to the atmosphere. Any line break smaller than the bounding break of a 24 inch steam line is discharged into the Turbine Building and processed through the Turbine Building (VT) exhaust to the Common Station HVAC Stack (Stack).

To verify that the dose consequences remain bounded for a smaller steam release the following evaluation was developed. Calculation C-023, Revision 2 analyzed a 140,000 pound mass coolant release with either a 0.2 uCi/gm or 4.0 uCi/gram of 1-131 dose equivalent activity per TS 3.4.8. This is a more conservative value than the actual USAR Section 15.6.4 total of 96,250 lb.

Total dose consequences are calculated based on concentrations at the receptor locations in the steam plume for the Control Room or based on the release times for the applicable XlQ for the Exclusion Area Boundary (EAB) and Low Population Zone (LPZ). Doses are calculated for inhalation (l.e., rem Committed Effective Dose Equivalent (CEDE)) and plume submersion (Le.,

Effective Dose Equivalent (EDE)) and totaled to yield rem Total Effective Dose Equivalent (TEDE). The activity of the cloud is based on the total mass of water released from the break.

This assumption is conservative because it considers the maximum release of fission products.

For comparative purposes, a 25 gpm water equivalent steam leak (l.e., current automatic isolation setpoint) was compared to the dose consequences of the MSLB to determine a time when similar dose consequences could be observed for a leak of that magnitude. Such a comparison would allow for determining when the dose for a small break of 25 gpm, which was the basis of the previous Turbine Building temperature trip, will produce a dose equivalent to the bounding MSLB accident.

In the EGC setpoint analysis, a 25 gpm water equivalent steam leak equates to 3.44 Ibm/second. This is equal to 12,384 Ibm/hour. Dividing the currently assumed 140,000 pound mass coolant release by 12,384 Ibm/hour results in reaching the dose equal to the bounding MSLB accident in 11.3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> with a small break of 25 gpm. With no iodine spike, this time is over 200 hours0.00231 days <br />0.0556 hours <br />3.306878e-4 weeks <br />7.61e-5 months <br />. These times are adequate to allow for operator action.

This duration (Le., 11.3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />) is based on the conservative assumption that the leak is primary coolant, and that the leak is treated in the same manner that a MSLB leak is treated, with assumptions for coolant activity, iodine spiking, and ground level release. A 25 gpm leak is not expected to yield spiking, as the leakage transient would not be large enough to produce such an effect. Additionally, a leak so small would most likely be all steam. Section 12.2.4 of the CPS USAR outlines that reactor steam becomes radioactive through the same sources as does reactor water. The concentrations of isotopes are different from those in reactor water and depend upon the carryover factors from liquid to vapor phase. The carryover factor for halogens (fission products) is less than 2%, and that for other fission and non-coolant activation products is less than 0.1%. Comparing the resultant carryover values associated with steam alone yield only a fraction of the currently analyzed consequences.

Similarly, for an intermediate steam leak of 100 gpm, approximately 2.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> is required to reach the dose for the bounding MSLB scenario. With no iodine spike, the time is approximately 50 hours5.787037e-4 days <br />0.0139 hours <br />8.267196e-5 weeks <br />1.9025e-5 months <br />. This size steam leak significantly bounds the conditions under which Operations could continue to run the plant since best-case winter capacity for all four Plant Chilled Water (WO) system chillers is exceeded by the heat from the steam alone under these conditions. Turbine Building heater bay temperatures would be expected to approach 200 Page 5 of 9

Attachment 1 Response to NRC Requests for Additional Information degrees F and the flows into the radwaste system would substantially exceed typical operating conditions.

The Stack is monitored by redundant normal range radiation monitors. Offsite Dose Calculation Manual (ODCM) Section 3.4.1 provides the Operational Requirement at the site boundary (500 mrem per year) with surveillance requirements per ODCM Sections 3.4.1.1 and 3.4.1.2 on sample and analysis frequency per ODCM Table 3.4-1. Additionally either of these two monitors provides high alarms allowing operator response in accordance with the alarm response procedures. These alarm response procedures note the automatic action to initiate the high range radiation monitor and the manual action to enter the off normal procedure for abnormal release of airborne radioactivity. This off normal procedure requires that operators isolate the source of the radioactive leak prior to reaching the detector high alarm up to and including closure of MSIVs. This alarm is intended to ensure that ODCM limits are not exceeded. The off normal procedure also refers to the CPS Emergency Plan Annex. If Alert limits (200 times ODCM limit, approximately 10 mrem/hr at the site boundary) are exceeded, Operator actions would be taken in accordance with emergency operating procedures to isolate primary and secondary pathways not needed for Emergency Operating Procedure (EOP) actions.

There is no significant safety related equipment in the Turbine Building, which can be impacted by a steam leak. Furthermore, there is no impact of small breaks in the Turbine Building on primary containment and drywell design parameters as there is for small breaks in the drywell.

For a small break in the Turbine Building, for which the make up system maintains reactor water level, fuel damage is not credible. With no moisture carryover in a small break, iodine transfer is minimal. Consequently, the delay and monitoring designed into the VT system and Stack provides appropriate control of any release such that an analysis of the entire spectrum of Steam Line breaks below the bounding break is not warranted.

A comparison of the MSL break analysis to the dose consequences for a smaller steam leak indicates that substantial time is available for existing operator actions to mitigate the expected spectrum of smaller steam leaks.

Response to Request No.2 Additional Questions:

Are the remaining instruments safety related?

The remaining MSL break detection instruments are reactor level 1, MSL pressure low, MSL flow high, and Auxiliary Building main steam tunnel temperature high. These are safety related instruments.

How they are credited to mitigate the consequences of a steam break?

The first instrument to detect the break is main steam line flow high. This is the only instrument credited in the analysis to initiate MSIV closure. All other instruments provide diverse initiation of isolation, and are not credited in the analysis. (Reference USAR Table 15.6.4-1)

Page 6 of 9

Attachment 1 Response to NRC Requests for Additional Information What is the timing of the detection credited systems?

Detection occurs (from main steam line flow high) and the isolation is initiated approximately 0.5 seconds after the break. Isolation completes within 5.5 seconds after the break. (Reference USAR Table 15.6.4-1)

Are the credited systems automatic or manual?

These detection instruments initiate automatic isolation.

What is the release from the steam line prior to isolation?

The mass release through a main steam line guillotine break is 96,250 lb. However, for conservatism in dose analysis only, a bounding value for current Boiling Water Reactor (BWR) plants of 140,000 Ib is used, based on Standard Review Plan (SRP) 15.6.2.4 for a General Electric Stride Safety Analysis Report (GESSAR) 251 plant.

What are the radiological consequences?

Accident dose consequence for the main steam line break are:

Location Case 1 (normal equilibrium limit of 0.2 uCi) Dose Case 2 (iodine spike (rem TEDE) limit of 4.0 uCi) dose (rem TEDE)

Limits CR: 5.0; EAB and LPZ: 2.5, RG 1.183 CR: 5.0; EAB and LPZ:

25, 10 CFR 50.67 EAB 0.0835 1.67 LPZ 0.0232 0.463 CR 0.193 3.85 (Reference calculation C-023 Revision 2)

The bounding dose consequences for a MSLB occur for an Iodine spiking case in the Control Room (CR).

Page 7 of 9

Attachment 1 Response to NRC Requests for Additional Information Request No.3.

Please explain the original basis for inclusion of the Turbine Building area temperature switches (used for main steam line isolations on high Turbine Building temperatures from TS Section 3.3.6.1, "Primary Containment and Drywell Isolation Instrumentation," Table 3.3.6 .1-1, (i.e., Function 1.f)) in your current technical specifications and design basis.

Response to Request #3 The Turbine Building High Temperature MSL isolations were part of the original design basis of CPS. The original basis of the isolation was to provide a main steam line isolation on an ambient temperature increase in the Turbine Building caused by a steam line leak equivalent to 25 gpm after two minutes. This isolation setpoint was revised for one of the five modules to allow an equivalent leakage of 25 gpm for five minutes. This change was associated with License Amendment 9, dated September 2, 1988. It was determined that radiological considerations were not included in the determination of this setpoint since it was not the bounding concern and that radiation monitors are installed in the exhaust ductwork from the Turbine Building and in the common station HVAC stack to which the exhaust is directed.

The Turbine Building High Temperature MSL isolation was maintained during the conversion to Improved Technical Specifications (ITS) although the standard ITS did not identify any requirements for this trip. At the time of the ITS conversion, the Main Steam Tunnel differential temperature instruments were deleted; however, no credit for the Turbine Building High Temperature MSL isolations was taken in the justification for the deletion.

The CPS Final Safety Analysis Report (FSAR) originally stated that the diversity of trip initiation signals for a main steam line leak is provided by main steam line tunnel ambient temperature, high differential temperature, which was deleted during ITS conversion, and main steam line high flow instrumentation. An increase in ambient temperature, differential temperature, or main steam line flow, will initiate main steam line and main steam line drain valve isolation. Diverse means of measuring excess leakage external to containment (e.g., process line break outside containment) are detected by low reactor water level, high process line flow, and high ambient temperatures in the steam tunnel and turbine building (to be deleted with this license amendment request), and high differential temperatures (deleted during ITS conversion) in the steam tunnel.

Based on the above, diverse means of MSL isolation will continue to be provided and the Turbine Building High Temperature isolation may be eliminated.

Page 8 of 9

Attachment 1 Response to NRC Requests for Additional Information Request No.4.

Your letter dated June 15, 2009, states that if the trip function for the temperature detectors in the Turbine Building main steam tunnel is removed from the Technical Specifications, " an alarm is intended to be maintained to alert the operator in the Control Room of a high temperature condition in the Turbine Building main steam tunnel." Please describe how this alarm will be maintained and used.

Response to Request No.4 The existing alarm function for two divisions of the Turbine Building temperature sensors will be maintained. As part of the design change, the trip function will be eliminated, and the temperature sensors will be used solely to provide alarm and monitoring functions. Five sensors from one division will actuate one alarm, and five sensors from a second division will input into an alarm and a recorder. The instrumentation will continue to be powered from a Class 1E power supply, and a description of these alarms will be provided in the CPS USAR. The alarm setpoints will be retained, and calibration of the instruments will continue to be performed on a 24-month frequency through CPS Predefined Maintenance Requirements (PMRQs).

Confirmation of a leak may be obtained by a second alarm on high temperature, plant stack exhaust high radiation alarms, or visual observation. Once leakage has been verified, operators will take action, up to and including actions to initiate closure of the MSIVs by manual means from the Main Control Room, or entry into Emergency or Abnormal Operating Procedures, if necessary.

Page 9 of 9

Attachment 2 Calculation C-023, Revision 2 Re-analysis of Main Steam Line Break (MSLB) Accident Using Alternative Source Terms

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  • I CALCULATION NO. C-023 t REV. NO. 2 PAGE NO. 2 of16 ~

CALCULAnON TABLE OF CONTENTS 1.0 PURPOSEIOBJECTIVE 3 2.0 METHODOLOGYAND ACCEPTANCE CRITERIA 3 2.1 General Description 3 2.2 Source Term Model 4 2.3 Release Model 4 2.4 Dispersion Model  ; 4 2.4.1 EAB and LPZ 4 2.4.2 Control Room 5 2.5 Dose Model  ; 5 2.5.1 EAB and LPZ 5 2.5.2 Control Room 5 2.6 Acceptance Criteria 6 3.0 ASSUMPTIONS 9 3.1 Activity Release and Transport 9

.4~g*2 DE~~~~~~:*:::::*:~:*::::::::::::::~:::~::::*:*:::::::::::::~:::::::.:::::::::::::.::::::::::::::::::::::::::::::::::::::~:::::::.::::::::~:::-~ ...

4.1 Mass Release Data ~ 9 4.2 Iodine and Noble Gas Activity Release 10 4.3 Control Room Data 10 4.4 EAB and LPZ Data 10

5.0 REFERENCES

11 6.0 CALCULATIONS 12 6.1 Cloud Volumes, Masses, and Control Room Intake Transit Times 12 6.2 Dispersion for Offsite Dose Assessment 13 6.3 Release Isotopics and Quantification 13 6.4 Dose Assessment 14 7.0

SUMMARY

AND CONCLUSIONS 15 8.0 OWNER'S ACCEPTANCE REVIEW CHECKLIST FOR EXTERNAL DESIGN ANALYSIS.16 Attachments:

A. Spreadsheet Performing Cesium Molar Fraction and Total MSLB Dose Assessment, With Formula Sheets [pages AI-AI5]

B. Computer Disclosure Sheet [pages B1-B1J

I CALCULATJON NO. C.Q23 I REV. NO. 2 PAGE NO. 3 of 16 I 1.0 PURPOSE/OBJECTIVE The purpose of this calculation is to determine the Control Room (CR), Exclusion Area Boundary (EAB), and Low Population Zone (LPZ) doses following a Main Steam Line Break (MSLB) Accident. This calculation is performed in accordance with Regulatory Guide (RG) 1.183 [Reference 6] as described herein.

The principal attributes of this analysis compared to those performed previously for this event under Standard Review Plan 15.6.4 guidance and 10CFR100 and 10CFR50, General Design Criterion 19 requirements are:

1. Doses are evaluated in terms of Total Effective Dose Equivalent (TEDE) and evaluated against 10CFR50.67 limits as modified by RG 1.183.
2. Noble gas releases are as previously analyzed and are not impacted by AST application.

.- _... '3~' Hlstorlcallydetermlne-llquld reactor coolant and steam release' continue to be the basis--*-*----*

for the determination of that no fuel damage results from an MSLB.

4. A simplified and more conservative basis is used for the determination of radionuclide releases based on a bounding reactor coolant blowdown value.
5. Iodine releases are based on reactor coolant 1-131 equivalent limitations in Clinton Power Station (CPS) Technical Specifications rather than the historical assumption of 3 times design basis values.
6. Cesium releases, as cesium iodide, are now considered in addition to iodine and noble gas release that have been historically assumed.

As per CPS USAR [Ref. 2) Section 15.6.4, this event involves the postulation that the largest steam line instantaneously and circumferentially breaks outside the primary containment at a location downstream of the outermost isolation valve, with this event representing the envelope evaluation of steam line failures outside primary containment. Closure of the Main Steam Isolation Valves (MSIVs) terminates the reactor coolant mass loss when the full closure is reached. No operator actions are assumed to be taken during the accident, so the normal air intake into the Control Room would continue unfiltered during the duration of the event.

The mass of coolant released durinq the MSLB is taken as a bounding value for all current Boiling Water Reactor (BWR) plants and for dose analysis purposes only of 140,000 pounds, as provided in Standard Review Plan 15.6.4. Paragraph 1II.2.afor a GESSAR-251 plant. This ensures that the discharge quantity and dose consequences are maximized.

2.0 METHODOLOGY AND ACCEPTANCE CRITERIA 2.1 General Description The radiological consequences resulting from a design basis MSLB accident to a person at the EAB; to a person at the LPZ; and to an operator in the Control Room following an MSLB accident were performed using a Microsoft EXCEL spreadsheet, provided as Attachment A.

I CALCULAnON NO. C*023 PAGE NO. 4 of 16 I 2.2 Source Term Model No fuel damage is expected to result from a MSlB. Therefore, the activity available for release from the break is that present in the reactor coolant and steam lines prior to the break, with two cases analyzed. Case 1 is for continued full power operation with a maximum equilibrium coolant concentration of 0.2 uCl/gm dose equivalent 1-131 [Ref. 8]. Case 2 is for a maximum coolant concentration of 4.0 uCllgm dose equivalent 1-131, based on a pre-accident iodine spike caused by power changes. This accident source term basis meets the guidance in RG 1.183 for analysis of this event.

Inhalation Committed Effective Dose Equivalent (CEDE) Dose Conversion Factors (DCFs) from Federal Guidance Report (FGR) No. 11 [Ref. 3] and External Dose Equivalent (EDE) DCFs from FGR No. 12 [Ref. 4] are used.

2.3 Release Model Noble gas releases are those historically determined to correspond to 100 uCl/sec-MWt after 30 minutes delay. These release quantities are taken from Reference 1.

Iodine releases are determined based a release of 140,000 Ibs of reactor coolant with either 0.2 uCilgm or 4.0 uCi/gm of 1-131 dose equivalent activity per Technical Specification 3.4.8 [Ref. 8].

The iodine species released from the main steam line are assumed to be 95% Csi as an aerosol, 4.85% elemental, and 0.15% organic. Therefore, 95% of iodine releases have an atom equivalent cesium release. Cesium isotopic abundance is determined based on source terms developed for pH control for longer lived or stable isotope [Ref. 13], and from ANSI/ANS-18.1-1999 [Ref. 10] for shorter lived isotopes.

Releases are assumed to be instantaneous and no credit Is taken for dilution in turbine building air.

2.4 Dispersion Model 2.4.1 EAB and lPZ EAB and lPZ XlQ's are determined using the methodology in RG 1.5 [Ref. 5], that is also cited as one basis for evaluation in the CPS-USAR. Specifically:

t 0.0133

-=--

Q OyU where o y = horizontal standarddeviationof the plume(meters) u = wind velocity(meters/second)

I CALCULATJON NO. C-023 I REV. NO.2 PAGE NO. 50f16 Horizontal standard deviations are taken from the PAVAN outputs for the EAB and LPZ included in Rev. 1 of Calculation C-015 [Ref. 9]. Per RG 1.5, F stability and a 1 meter/sec wind speed are used.

2.4.2 Control Room For control room dose calculations, the plume was modeled as a hemispherical volume, the dimensions of which are determined based on the portion of the liquid reactor coolant release that flashed to steam. The activity of the cloud is based on the total mass of water released from the break. This assumption is conservative because it considers the maximum release of fission products.

Activity release is conservatively assumed to effectively occur at the Control Room intake elevation and, again conservatively, no credit is taken for plume buoyancy.

2.5 Dose Model Dose models for both onsite and offsite are simplified and meet RG 1.183 [Ref. 6] requirements, providing results in units of Total Effective Dose Equivalent (TEDE). Dose conversion factors are based on Federal Guidance Reports 11 and 12 [Refs 3 & 4].

2.5.1 EAB and LPZ Doses at the EAB and LPZ for the MSLB are based on the following formulas:

DoseCEOE (rem) =Release(Curies)* ~ (sec/m")

  • Breathing Rate(m?/sec)
  • Inhalation DCF(remCEDE ICi inhaled) and DoseEPE (rem) = Release(Curies)*1. (sec/rn? ). Submersion DCF(remeoe -m? / Ci -sec)

Q and finally, Dose TEOE (rem) =Dose CEOE (rem) + Dose EOE (rem) 2.5.2 Control Room CR operator doses are determined somewhat differently. Steam cloud concentrations are used, rather that XJQ times a curie release rate. No CR filter credit is taken and, therefore, for inhalation, a dose for a location outside of the CR is used. For cloud submersion, a geometry factor is used to credit the reduced plume size seen in the CR. This is a conservative implementation of RG 1.183 guidance. The formulas used are:

DoseCEDE (rem) = PlumeConcentration (Ci/m")

  • TransitDuration(sec)
  • Breathing Rate(m' /sec)
  • Inhalation DCF(remCEDE/Ci inhaled) and

I CALCULATION NO. C-023 I REV. NO. 2 PAGE NO. 6ofl6 I DoseEDE (rem) =Plume Concentration (CiJm)* Transit Duration (sec). Submersion DCF(remEOE - m 3/ Ci *sec) and finally, DoseTEDE (rem) =DoseCEDE (rem) + DoseEDE (rem) 2.6 Acceptance Criteria Dose acceptance criteria are per 10CFR50.67 [Ref. 7] and RG 1.183 [Ref. 6), Table 6 guidance.

Table 2.1 lists the regulatory limits for accidental dose to 1) a control room operator, 2) a person at the EAB, and 3) a person at the LPZ boundary.

Table 2.1. Regulatory Dose Limits (Rem TEDE) per Refs. 7 and 6.

1-131 Dose CR EAB LPZ Equivalent (30 days) (2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />) (30 days Normal Eauilibrlum 5 2.5 2.5 Iodine Spike 5 25 25 Direct conformance with the Assumptions in Regulatory Guide 1.183 Appendix D "Assumptions for Evaluating the Radiological Consequences of a BWR Main Steam line Break Accident" is provided by this analysis, as shown in the Conforrilance MatrixTable 2.2.

I CALCULATION NO. C-023 I REV. NO. 2 PAGE NO. 70fld o'-RG SectiOQ 1 Assumptions acceptable to the NRC staff regarding core inventory and Not Applicable No fuel damage. release the release of radionuclides from the fuel are provided in Regulatory estimate based on coolant Position 3 of this guide. The release from the breached fuel is based on activity.

Regulatory Position 32 of this guide and the estimate of the number of fuel rods breached.

2 If no or minimal fuel damage is postulated for the limiting event. the Conforms See below released activity should be the maximum coolant activity allowed by technical specification. The iodine concentration in the primary coolant is assumed to correspond to the following two cases in the nuclear steam supply system vendor's standard technical specifications.

2.1 The concentration that is the maximum value (typically 4.0 IJCilgm DE 1- Conforms 4.0 uCilgm DE 1-131 is a limit 131) permitted and corresponds to the conditions of an assumed pre- in Technical Specification accident spike. and (TS) 3.4.8 [Ref. 8] and is u~

in this analysis.

2.2 The concentration that is the maximum eqUilibrium value (typically 0.2 Conforms 0.2 uCi/gm DE 1-131 is a limit IJCi/gm DE 1-131) permitted for continued full power operation. in TS 3.4.8 [Ref. 81 and is I used in this analysis.

3 The activity released from the fuel should be assumed to mix Conforms No fuel damage.

instantaneously and homogeneously in the reactor coolant. Noble gases should be assumed to enter the steam phase instantaneously..

4.1 The main steam line isolation valves (MSN) should be assumed to close Conforms An MSIV closure time of in the maximum time allowed by technical specifications. 5.5 seconds was assumed in the analysis. This is the I Technical Specification SR 3.6.1.3.6 maximum allowed MSIV closure time of 5 seconds plus 0.5 seconds for instrument response.

4.2 The total mass of coolant released should be assumed to be that Conforms A bounding value of 140.000

I CALCULAnON NO. C-023 I REV. NO. 2 PAGE NO. 800d

'. T~ble i2:'~ C(mform.an~~~~:R.Gj1.:183~#~~di~P~(Miit~*:Steaa.;U"~:~r,e~_~f*  :~~,~.:: :;,;:.;.:;,~ ,.:,"f)t,~+~

, .. . .. .--.- . .:, .. , .... ' ':X ". ; tr . ... .... ***~DresdenJQuad*'~*

~".f~: :~~;~aIY~i~:'; S*~**:-:***<:* ::~;~:~~~:,~}~~~::;:~~~~

.~

RG , - - .. , . -' ,.

.*$eclioo* - ,

RG Positioo : - . ::':;:

amount in the steam line and connecting lines at the time of the break Ibs or reactor coolant is used plus the amount that passes through the valves prior to closure. for dose assessment 4.3 All the radioactivity in the released coolant should be assumed to be Conforms Release is assumed at released to the atmosphere instantaneously as a ground-level release. ground level, with no credit .

No credit should be assumed for plateout, holdup, or dilution within taken for plateout, holdup.or facility bUildings. dilution within facility bUildings.

4.4 The iodine species released from the main steam line should be Conforms The subject values are used.

assumed to be 95% Csi as an aerosol, 4.85% elemental, and 0.15%

organic.

I

I CALCULAnON NO. C-023 PAGE* NO.9 006 I 3.0 ASSUMPTIONS 3.1 Activity Release and Transpolt (I Iodine coolant activity isotopic distributions and Noble Gas activity releases are taken from Appendix A of the Extended Power Uprate Task T0901 Accident Radiological Analysis [Reference 1], consistent with USAR Section 15.6.4.5 [Reference 2].

  • Release from the break to the environment is assumed instantaneous. No holdup in the Turbine Building or dilution by mixing with Turbine Building air volume is credited.
  • The steam cloud is assumed to consist of the portion of the liquid reactor coolant release that flashed to steam.

e The activity of the cloud is based on the total mass of water released from the break.

This assumption is conservative because it considers the maximum release of fission products.

(I Buoyancy effect of the cloud was conservatively ignored.

iii For the control room dose calculations,

,. The plume was modeled as a hemispherical volume. This is consistent with the assumption of no Turbine BUilding credit. It is also reasonable for the more likely release paths through multiple large heat and smoke vents situated In the Turbine Building Roof.

,. Dispersion of the activity of the plume was conservatively ignored.

The cloud was assumed to be carried away by a wind of speed 1 m/s. Credit is not taken for decay.

3.2 Control Room

  • Inhalation doses are determined based on concentrations at the intake, and exposures for the duration of plume traverse.
  • External exposure doses are determined based on concentrations at the intake, exposures for the duration of plume traverse, and a geometry factor credit (Equation 1 of Ref. 6) based on the control room envelope volume of 324,000 cubic feet (Reference 12).

4.0 DESIGN INPUT 4.1 Mass Release Data Per Reference 1, the mass release following a MSLB is the USAR Section 15.6.4 total of 96,250 lb of the primary coolant [Ref. 1 and 2]. As stated in USAR Section 15.6.4.3.2, there is no fuel damage as a consequence of this accident [Ref. 2]. For this dose analysis, a more conservative 140,000 Ibs of primary coolant are assumed to be released.

I======,;;;,,;;;;,

CALCULATION NO. .............

C-023;,;;;;o;,, I...R..E..V,;,,;.N_O_._2 ....._ _..;,;,,;,;;;;,;;,,;,,,;,;;;,,;,,,,;,,;;,,,,;;,:.~

PAGE NO. 10 ofI6 4.2 Iodine and Noble Gas Activity Release The following design basis MSlB noble gas release quantities are provided In EPU-T0901 Appendix A: Activity Release to the Environment Following a Design Basis Accident [Ref. 1].

Noble Gas Activity Release Isotopes (CI)

Kr-83M 0.0695 Kr-85M 0.122 Kr-85 0.000475 Kr-87 0.379 Kr-88 0.389 Kr-89 1.62 Xe-131M 0.000388 Xe-133M 0.00581 Xe-133 0.163 Xe-135M 0.476 Xe-135 0.439 Xe-137 2.14 Xe-138 1.62 Reactor coolant iodine activity distributions are taken from Ref. 1 and USAR [Ref. 2] Sections 15.6.4.5.1.1 and 15.6.4.5.2.2, as follows:

Iodine Activity Distribution Isotopes (uCIIgram) 1-131 0.015 1-132 0.150 1-133 0.100 1-134 0.300 1-135 0.150 Release activities are calculated in Attachment A.

4.3 Control Room Data

  • =

Control Room Envelope 324,000 ft3. [Ref. 12]

" No Emergency Filtration Credit taken.

4.4 EA8 and LPZ Data

" EAB Distance from Release. m 975 [Ref. 11]

II LPZ Distance from Release. m 4018 [Ref. 11]

  • EAB Gy. m 36.1 fRef. 9]
  • LPZ G y, m 129.8 [Ref. 9]

I CALCULATION NO. C-023 I REV. NO. 2 PAGE NO. II of 16

5.0 REFERENCES

1. Extended Power Uprate Task T0901: Accident Radiological Analysis, EPU-T0901 Rev.

O.

2. Clinton Power Station USAR Rev. 11.
3. Federal Guidance Report No. 11, "limiting Values of Radlonuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion",

1988.

4. Federal Guidance Report No. 12, "External Exposure to Radionuclldes in Air, Water, and Soli", 1993.
5. RegUlatory Guides 1.5. "Assumptions Used for Evaluating the Potential Radiological Consequences of a Steam Line Break Accidents for Boiling Water Reactors," 3/10/71.
6. Regulatory Guide 1.183, "Alternative Radiological Source Terms For Evaluating Design Basis Accidents At Nuclear Power Reactors", July 2000.
7. 10 CFR Part 50.67, "Accident source term", January 1, 2001.
8. Clinton Technical Specification, RCS Specific Activity. 3.4.8.
9. Calculation C-015, Rev. 1 "Calculation of Alternative Source Term (AST) Onsite and Offsite XlQ Values".
10. American Nuclear Society Standard (ANS) 18.1-1999 "Radioactive Source Terms For Normal Operation of Light Water Reactors", Table 5.
11. Clinton Power Station Site Development Drawing No. M01-1101, Rev. J.
12. Calculation C-022, Rev. 0 "Site Boundary and Control Room Dose Following a FHA in Containment Using Alternate Source Terms".
13. PBAPS Calculation PM-1056, Rev. 1, "Suppression Pool pH Calculation for Alternative Source Terms".
14. ASME Steam Table Data from "STEAM TABLES* PROPERTIES OF STEAM AND WATER", Copyright 1996 by Sclentech, Inc.

I CALCULATION NO. C-023 I REV. NO. 2 PAGE NO. 12 of 16 I 6.0 CALCULATIONS No fuel damage Is expected for the limiting MSLB. As discussed in Section 2, two iodine concentrations are used (0.2 IJCi/g and 4.0 IJCi/g) [per Ref. 6] when determining the consequences of the main steam line break. All of the radioactivity In the released coolant Is assumed to be released to the atmosphere instantaneously as a ground-level release. No credit is taken for plateout, holdup, or dilution within facility buildings.

The spreadsheets in Attachment A perform this analysis using data and formulations discussed above and shown in Attachment A. The following summarizes parameters and their treatment in the spreadsheet.

6.1 Cloud Volumes, Masses, and Control Room Intake Transit Times The cloud Is assumed to consist of the portion of the conservatively bounding liquid reactor coolant release that flashes to steam. The flashing fraction (FF) is derived as follows:

FF x (steam enthalpy at 212 F) + (1-FF) x (liquid enthalpy at 212 F) =

(liquid enthalpy at temperature of steam entering the turbine)

A 539.9 F turbine inlet temperature is used, from the Valves Wide Open Heat Balance at 105% Turbine Rated Steam Flow of USAR [Ref. 2J Figure 10.1-2, with liquid enthalpy of 536.6 BTU/lb.

At 212 F, a steam enthalpy of 1150.5 BTUIlb and a liquid enthalpy of 180.17 BTUllb are used (these enthalples are taken from the ASME Steam Tables, Reference 14).

Substituting, FF = (536.6 - 180.17) / [(1150.5 - 180.17)] = .367 For conservatism, a value of .40 or 40% is used.

As stated in Section 3.1, the cloud Is assumed to consist of the initial steam blowdown and that portion of the liquid reactor coolant release that flashed to steam.

The mass liquid water released = 140,000 Ib Flashing fraction for calculating cloud volume =40%

The mass of water carrying activity into the cloud = 140,000 Ib

=(140,000 Ib)(453.59 glib)

= 6.350E7 g The mass of steam in the cloud =40%*140,OOOlb

= 56,000 Ib The release is assumed to be a hemisphere with a uniform concentration. The cloud

=

dimensions (based on 56,000 Ib of steam at 14.7 psi and 212 OF, vg 26.799 ft3/1b) were calculated as follows:

I CALCULATION NO. C*023 PAGE NO. 13006 ~

Volume = (56,000 Ib)(26.799 ft3/1b)

=1,500,744 tr

= (1,500,744 fe)/(35.3 trlm3 )

= 42,514 m3 The volume of a hemisphere Is 1t d3 /12. Thus, the diameter of the hemispherical cloud is 54.6 meters.

The period of time required for the cloud to pass over the control room intake, assuming a wind speed of 1 mls is 54.6 s (=(54.6 m)/(1 mls>>. Therefore, at a wind speed of 1 mIs, the base of the hemispherical cloud will pass over the control room intake in 54.6 seconds.

6.2 Dispersion for Offslte Dose Assessment As discussed in Section 2.4.1 the following formulation was used for Offsite Dose XlQ assessment, with F Pasquill Stability and a 1 m/sec wind speed.

x, 0.0133 Q ayu where a y = horizontal standarddeviationof the plume(meters) u = wind velocity(meters/second)

As calculated in the PAVAN run in Reference 9, at the 975-meter EAB distance a y Is 36.1, and at the 4018-meter LPZ distance cry is 129.8. The resulting EAB and LPZ X/Os are 3.68E-4 and 1.02E-Q4 sec/m 3 , respectively.

6.3 Release Isotoplcs and Quantification The iodine, noble gas and cesium activity releases are given in Attachment A, which also determines reSUlting doses.

Noble gas releases are taken from the input in Section 4.2.

Iodine releases are based on reactor coolant isotopic distributions from Section 4.2, which are normalized based on FGR-11 CEDE dose conversion factors to obtain coolant concentrations corresponding to Case 1: 0.2 uC/gm, and Case 2 4.0 uCilgm, which are the TS 3.4.8 values.

The resulting concentrations were multiplied by the 140,000 Ibs of release converted to grams.

Cesium releases are based on the fact that a single cesium atom will accompany 95% of the released iodine atoms. For Cs-133, CS-134, Cs-135, and Cs-137, isotopic data for end of cycle conditions from Reference 13 were used. For shorter lived isotopes such as Cs-136 and Cs-138, the ratio of their concentration values in Reactor Water to that of Cs-137 in Reference 10 is used to predict their relative concentrations. Releases reflect this distribution, with the molar

I CALCULATION NO. C-023 I REV. NO.2 PAGENO."14 of 16 fractions converted to curie quantities based on the isotope's decay constant. Cs-133, representing about 38% of the cesium, is stable.

6.4 Dose Assessment Doses at the EABand LPZ distances, and in the Control Room are calculated in Attachment A using the formulas in Section 2.5. Concentrations at the receptor locations are that In the steam plumefor the Control Room or based on the release times the applicable XlQ for the EAB and LPZ.

Doses are calculated for inhalation (rem CEDE) and plume submersion (rem EDE) and totaled to yield rem TEDE. The breathing rate of 3.47E-04 m3/sec is per RG 1.183 guidance without the round-off.

The resulting calculated doses are in the spreadsheet and in the Summary and Conclusions Section below.

I CALCULATION NO. C-023 I REV. NO. 2 PAGE NO. 15 of 16 7.0

SUMMARY

AND CONCLUSIONS Accident doses from a design basis MSLB were calculated for the control room operator, a person at the EAB, and a person at the LPZ. The results are summarized in the Table below. The doses at the Control Room, EAB, and LPZ resulting from a postulated design basis MSlB do not exceed, and are a small fraction of, the regulatory limits.

Location Case 1 Case 2 (normal equilibrium. (Iodine spike limit of 0.2IJ.CI) limit of 4.0 IJ.CI)

Dose (rem TEDE) Dose (rem TEDE)

LIMITS CR:5.0* EAB&LPZ: 2.5 CR: 5.0; EAB&LPZ:25 EAB 0.0835 1.67 LPZ 0.0232 0.463 CR 0.193 3.85

I CALCULATION NO. C-023 I REV. NO.2 PAGE NO. 16ofI6 8.0 OWNER'S ACCEPTANCE REVIEW CHECKLIST FOR EXTERNAL DESIGN ANALYSIS DESIGN ANALYSIS NO. C-023 REV: 2 if~

N/A I. Do assumptions have sufficient rationale? 0

2. Are assumptions compatible with the way the plant is operated and with the licensing basis? ~ 0 0

~ 0

3. Do the design inputs have sufficient rationale? 0
4. Are design inputs correct and reasonable? 0 0 S. Are design inputs compatible with the way the plant is operated and with the licensing basis? ~ 0 0

~

6. Are EngineeringJudgments clearly documentedand justified? 0 0 Are Engineering Judgments compatible with the way the plant is operated and 7.

with the licensing basis? .0 0 8.

Do the results and conclusions satisfy the purpose and objective of the Design Analysis? rtf 0 0 9.

Are the results and conclusions compatible with the way the plant is operated and with the licensing basis? ~ 0 0

10. Does the Design Analysis include the applicable design basis documentation? ~ 0 0 II.

Have any limitationson the use of the results been identifiedand transmitted to the appropriate organizations?

Il(( 0 0

12. Are there any unverified assumptions? 0 ~ 0 Do all unverified assumptions have a trackingand closure mechanism in
13. place? 0 0 ~

14.

Have all affected design analyses been documented on the Affected Documentslist (ADL) for the associated ConfigurationChange? ~ 0 0 Do the sources of inputs and analysis methodology used meet current

~

technical requirementsand regulatory commitments? Of the input sources or

15. analysis methodology are based on an out-of-datemethodologyor code, 0 0 additional reconciliation may be required if the site bas since committed to a more recent code) 16.

Have vendor supporting technical documents and references(including GE DRFs) been reviewed when necessary? ~ 0 0 EXELON REVIEWER: - DATE: 34,w,

A I 8 I C I 0 E F G I H I I I J I K l M 1 Clinton MSLB Dose Spreadsheet ease 1: Reactor Coolantat maximum value (DE 1-131 of 0.2 uCilg) permitted for 2 I continued full power operation 3 42514 Volume of cloUd (cubic meters) ease 2: Reactor Coolant at maximum value permitl8d (DE 1-131 of 4.0 uCilg) 4 6.35E+07 Mass of water In reactor coolant release (grams) corresponding to an assumad pre-accident spike 5 54.6 seconds for cloud to pass over CR intake for wind sPell< of 1 mlsecond 6 324000 Volume of Control Room Envelope (cubic feel) - maximum used for conservatism 7 140.000 Mass of liQuid Water Released (Ibl 8 40% Flashing Fraction I 9 56000 Mass of Steam in the Cloud lib) 10 26.799 VQ (ft'nbl (based on 14.7 osi and 212F 11 Reactor coolant iodine distribution is assumed to be a 1 am/ex: soed IicQravitv 12 Case 1 Casa2 13 Release Release 14 NormaUzed Case 1 Case 2 Casa 1 Casa2 Cloud Cloud Case 1 Casa2 15 Isotope Activity FGR 11 1*131 DE Normalized NormaliZlld Activity Activity Cancenlration Concanlralion Decav Activity A~

16 DIstribution DCF ActMtv Activitv ActIvity Release Release Constant Release Release 17 uCVgm Retneeoe/Cl uCVgm uCilgm uCilgm Ci Ci CVm3 CVm3 1/seconds moles moles 18 1*131 0.015 3.29E+04 1.50E-G2 7.26E-G2 1.45E+OO 4.61E+OO 9.22E+Ol 1.08E..Q4 2.17E-G3 9.98E-G7 2.84E-G7 5.68E-06 19 1-132 0.15 3.81E+02 1.74E-G3 8.41E-G3 1.68E-G1 4.61E+01 9.22E+02 1.08E-G3 2.17E-G2 8.37E-G5 3.38E-oB 6.77E-G7 20 1-133 0.1 5.85E+03 1.78E-G2 8.60E-G2 1.72E+OO 3.07E+Ol 6.15E+02 7.23E..Q4 1.45E-G2 9.26E-G6 2.04E-G7 4.08E-G6 21 1-134 0.3 1.31E+02 1.19E-G3 5.78E-G3 1.16E-G1 922E+01 1.84E+03 2.17E-G3 4.34E-G2 2.20E..Q4 2.58E-G8 5.16E-G7 22 '*135 0.15 1.23E+03 5.61E-G3 2.72E-G2 5.43E-G1 4.61E+01 9.22E+02 1.08E-G3 2.17E.()2 2.91E-G5 9.73E-oB 1.95E-06 23 Totals 4.13E-G2 2.00E-G1 4.00E+OO Totals 6.45E-G7 l.29E-G5 24 "non-&Diked" '&Diked" 25 26 Case 1 Case 2 27 Case 1 Casa2 Release Release 28 ActMty ActMty CloUd Cloud 29 Release Release Concentration Concentration 30 Cl Ci CVm3 Ci1m3 31 Kr-83M 6.95E-G2 6.95E-G2 1.63E-G6 1.63E-G6 Case 1 Case 2 Case 1 Casa2 32 Kr-85M 1.22E-G1 1.22E-G1 2.87E-G6 2.87E-G6 ActMty Actlvitv Decay Activity Activity 33 Kr~ 4.75E..Q4 4.75E-G4 1.12E-oB 1.12E-G8 Release Release Constant Release Release 34 Kr-87 3.79E-<l1 3.79E-G1 8.91E-G6 8.91E-G6 MoIarFrae. moles moles 1/seconds curies curies 35 Kr-88 3.89E-G1 3.89E-G1 9.15E-G6 9.15E-G6 Cs-134 4.4317% 2.71E-oB S.43E-G7 1.07E-oB 4.71E-G3 9.42E-G2 36 Kr-89 1.62E+OO 1.62E+OO 3.81E-GS 3.81E-G5 Cs-135 17.4506% 1.07E-G7 2.14E-G6 9.SSE-15 1.66E-G8 3.32E-G7 37 Xe-131M 3.88E..Q4 3.88E..Q4 9.13E-G9 9.13E-G9 Cs-136 0.0120% 7.35E-11 1.47E-G9 6.10E-G7 7.29E-G4 1.46E-G2 38 Xe-133M 5.81E-G3 5.81E-G3 1.37E-G7 1.37E-G7 C8-137 40.17% 2.46E-G7 4.92E-G6 7.28E*10 2.92E-G3 S.83E-G2 39 Xe-133 1.63E-G1 1.63E-G1 3.83E-G6 3.83E-G6 Cs-138 0.0102% 6.24E-ll 1.25E-G9 3.59E-G4 3.65E-G1 7.29E+OO 40 Xe-l35M 4.76E-G1 4.76E-G1 1.12E-G5 1.12E-G5 Tolals 62.08% 3.80E-G7 7.61E-G6 41 Xe-135 , 4.39E-G1 4.39E-G1 1.03E-G5 1.03E-G5 Balance Is stable C&-133 42 Xe-137 2.14E+OO 2.14E+OO 5.03E-GS S.03E-G5 I I 43 Xe-138 1.62E+OO 1.62E+OO 3.81E-G5 3.81E-G5 I I 44 I I GaIClNlion C-023.Rev.2 AtlachmentA

  • DoseC8IaJJation PageA10f A1S

A B C J 0 E F J G J H I J I K l tot 45 Curies Released Case 1 Dose (rem CeDE) casE 2 Dose (rem CEDE) 46 to the Environment (Inhalation) (Inhalation) 47 Isotope Case 1 Case 2 DeF' CR EAB LPZ CR EAS LPZ 48 1-131 4.61E+OO 9.22E+Ol 3.29E+04 6.76E-02 1.94E-02 5.39E-03 1.35E+OO 3.88E-Ol 1.08E-Ol 49 1-132 4.61E+Ol 9.22E+02 3.81E+02 7.82E-03 2.25E-03 6.25E-04 1.56E-Ol 4.49E"'()2 1.25E"'()2 50 1-133 3.07E+Ol 6.15E+02 5.85E+03 8.ooE-02 2.3OE-02 6.39E"'()3 1.6OE+OO 4.6OE-Ol 1.28E-Ol 51 1*134 9.22E+Ol 1.84E+03 1.31E+02 5.38E-03 1.54E-03 4.3OE-04 1.08E"'()1 3.09E-02 8.59E..()3 52 1-135 4.61E-+Ol 9.22E-+02 1.23E+03 2.53E...()2 7.25E-03 2.02E.()3 5.05E-Cl 1.45E-C1 -4.03f...()2 53 54 Cs-134 4.71E-03 9.42E"'()2 4.63E+04 9.70E-05 2.78E-05 7.74E-06 l.94E-03 5.57E-04 1.S5E...()4 55 Cs-135 1.66E-08 3.32E"'()7 4.55E+03 3.37E-l1 9.67E-12 2.69E-12 6.74E-l0 1.93E-l0 5.38E-ll 56 Cs-l36 7.29E-04 1.46E-02 7.33E+03 2.38E-06 6.83E",()7 1.90E...()7 4.76E",()5 1.37E-05 3.8OE-06 57 C5-137 2.92E-03 5.83E-02 3.19E+04 4.15E-05 1.19E-OS 3.31E..()6 8.29E-04 2.38E-G4 6.62E"'()5 58 Cs-l38 3.65E-Ol 7.29E+OO 1.01E+02 1.65E-OS 4.73E-06 1.31E..()6 3.29E-04 9.45E-05 2.63E-05 59 Sub-total (rem CEDE) 1.86E-01 5.3SE-02 1.49E-02 3.72£+00 1.07E+OO 2.97E-01 60 61 Curies Released Case 1 Dose (rem EOE) Case 2 Dose (rem EOEl 62 to the environment (External) (External) 63 Isotope Case 1 ease 2 ocP CR EAB LPZ CR fAS LPZ 64 1-131 4.61E+OO 9.22E+Ol 6.73E-02 2.48E-05 1.14E-04 3. 18E-05 4.95E-04 2.29E-03 6.36E-04 65 1-132 4.61E+Ol 9.22E+02 4. 14E-Ol 1.52E-03 7.04E-03 1.96E..()3 3.05E-02 1.41E-Ol 3.92E-02 56 1-133 3.07E+Ol 6.15E+02 1.09E-Ol 2.67E-04 1.23E-03 3.43E-04 5.33E-03 2.46E-02 6.85E"'()3 67 1-134 9.22E+Ol 1.84E+03 4.81E-Ol 3.54E-03 1.63E...()2 4.55E-03 7.0TE-02 3.27E-Ol 9.09E-02 68 1-135 4.61E+01 9.22£+02 2.9SE...()1 1.09E-03 5.02E-03 1.40E..()3 2.1TE-02 1.ooE-Ql 2.79E"'()2 59 70 Cs-l34 4.71E"'()3 9.42E"'()2 2.BOE-01 1.05E-07 4.86E-07 1.35E-07 2. 1OE-06 9.72E-06 2.70E..()6 71 Cs-l35 1.56E"'()8 3.32E"'()7 2.09E-06 2.nE-18 1.26E-17 3.56E-18 5.54E-17 2.56E*16 7.12E-17 72 Cs-l36 7.29E-04 1.46E-02 3.92E-Ol 2.28E-08 1.05E-07 2.93E-08 4.56E-07 2.11E-06 5.86E"'()7 73 C5-137 2.92E-03 5.83E-02 2.86E-05 5.66E*12 3.08E-l1 8.56E-12 1.33E-l0 6.15E-l0 1.71E-l0 74 Cs-l38 3.65E-Ol 7.29E+OO 4.48E-Ql 1.3OE-05 6.01E-05 1.67E-05 2.60E-04 1.20E-03 3.35E-04 75 76 Sub-total (rem EOE 6.45E-03 2.98E-02 8.29E-03 l.29E-O 5.96E-01 1.66E-01 rr Iodine and Cesium Total (rem TEDE 1.93E-01 8.33E.o2 2.32E-02 3.85£+00 1.67E+OO 4.63E.o1 78 Cur\e$ Released Case 1 Dose rem ccE Case 2 Dose (rem EOE) 79 to the environment (External) (External) 80 Case 1 Case 2 DCF2 CR EA8 LPZ CR EAB LPZ 81 Kr-83M 6.95E-02 6.95E-02 5.55E..()6 3.08E-11 1.42E-l0 3.95E-l1 3.08E-l1 1.42E-l0 3.95E-ll 82 Kr-85M 1.22E-01 1.22E-Ol 2.nE-02 2.69E-07 1.24E-D6 3.46E-07 2.69E-07 1.24E-06 3.46E-07 83 Kr-85 4.75E-04 4.75E-04 4.40E-04 1.67E-11 7.71E-11 2.14E-11 1.67E-11 7.71E-l1 2.14E-l1 84 Kr-87 3.79E-Ol 3.79E-01 1.52E-01 4.61E-06 2.13E-05 5.92E-06 4.61E-06 2. 13E-05 5.92E..()6 85 Kr-88 3.89E-01 3.89E-Ol 3.77E-Ol 1.17E-05 5.41E-05 1.50E"'()5 1.17E-OS 5.41E-05 1.50E-OS 86 Kr-89 1.62E+OO 1.62E+OO O.OOE+OO O.ooE+OO O.ooE+OO O.OOE+OO O.ooE+OO O.OOE+OO O.OOE+OO 87 Xe-131M 3.88E-04 3.88E-04 1.44E.()3 4.45E-11 2.06E-l0 5.72E-ll 4.4SE-11 2.06E-l0 5.72E-11 88 Xe-l33M 5.81E-03 5.81E-03 5.07E-03 2.35E-09 1.09E-08 3.02E-09 2.35E-09 1.09E-08 3.02E-09 89 Xe-133 1.63E-Ol 1.63E-01 5.nE-03 7.50E-08 3.47E-07 9.64E-08 7.50E-08 3.47E-Q7 9.64E-08 90 Xe-l35M 4.76E-Dl 4.76E-Ol 7.S5E-02 2.86E-06 1.32E-05 3.68E-06 2.86E-06 1.32E-05 3.68E-06 91 Xe-135 4.39E",()1 4.39E-Ol 4.4OE-02 1.54E-06 7.12E-06 1.98E-06 1.54E-06 7.12E-06 1.98E-06 92 Xe-137 2. 14E+OO 2.14E+OO O.OOE+OO O.ooE+OO O.OOE+OO O.ooE+OO O.ooE+OO O.ooE+OO O.OOE+OO 93 Xe-l38 1.62E+OO 1.62E+OO 2.13E-Ol 2.76E-05 1.27E-04 3.54E-OS 2.76E-OS 1.27E-04 3.54E-05 94 Noble Gas SuIHotal rem EDEl 4.86E.05 2.2SE-04 6.25E-OS 4.l1tiE-OS 2.25E-04 6.25E-G5 95 I 96 Overall Totallrem TeDEl U3E-01 8.3SE-02 2.32E-G2 3.85E+OQ 1.611:"tQU 4.631:041 CalQllation C-023, Rev. 2 AtIadunent A- Oo&e Calculation PageA20f A15

A B I C I 0 I E. I F I G I H I J K L M 97 I I I I I 98

  • Dose CorMlrslon Faclor IrsmICutlel from Federal Guldance Report (FGR) 11 Il8r Reg. GuldB 1.183 99 > Dose eon-sIon Fader (1lIf1HIl'/CurIe-seccnd) from FGR 12 per Reg. GI.Ide 1.183 100 3.47E-04 Brealhlng rate (m'lsecond) per Regulelory GuIde 1.183 (without I'OUIllkJIf) 101 621E-02 ~RoomlieometryFactor per Reg. Gulde 1.183, Poslllon 42.7 102 3.61E'"01 EAB a, (meI8rII) for F stabiiIy. (laken from PAVAH NIla in Ref. 9) 103 1.298E'"02 LPZ a, (meters) for F alabIIiIy. (taken from PAVAH runs In Ref. 9) 104 1.00E+OO WInd Speed (mIs) I I I 105 3.68E-04 XIQ (secondsIm') at EA BoundaJy 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> based on RG 1.5 methodology 106 1.02E-04 XIQ (secondaIm°) at LowPopulatIon Zone* 0-2 based on RG 1.5 CaIQl\aIion C-023. Rev. 2 Allachment A* Dose Calcula1ion PageA30f A15

A B C 0 E F 1 C'inton MSLS Dose; Case 1:

2 3 =(A9"Al0)l35.3 Volume of clOUd(cubl Case 2:

4 :Ar453.59 Mass of water in react 5 =(A3"12IPIOY'(113) seconds for cloud to D 6 324000 Volume of Control ROt 7 140000 Mass of Llauid Water 8 0.4 Flashing Fraction 9 :ArAB Mass of Steam in the 10 26.799 Vg (tt'IIb) (based on 1 11 Reactor coolantlodine OIS 12 13 14 Normalized Case 1 Casa2 15 isotoPe Aclivitv FGR 11 1-131 DE Normalized NoImalized 16 Distribution DCF Activity Activltv Activity 17 uCVgrn RlIItlcEDEICl uCilgm uCilgm uCilgm 18 1-131 0.015 32900 =C18OS181C$18 :018"0.2JD$23 =E18*2O 19 1-132 0.15 381 =C19"Bl9/CS18 -019"0.210523 =£19"20 20 1-133 0.1 5846 CC20OS201C$18 =020"0.210$23 =£20"20 21 1*134 0.3 131 =C21"B21/C$18 =021"0.210$23 =E21"2O 22 1-135 0.15 1230 aC22"B22IC$18 =022"0.2ID$23 =E22"2O 23 Totals =SUM(018:022) =SUM(E18:E22) =SUM(F18:F22) 24 "$Diked" 25 26 Case 1 ease 2 27 ease 1 ease 2 Release Release 28 ActMlv ActMtv Cloud Cloud 29 Release Release Conl:entration Concentration 30 Ci Ci CIfm3 CUm3 31 Kr-83M 0.0695 0.0695 =C311$A$3 =D311SA$3 32 Kr-85M 0.122 0.122 =C321$A$3 :D32/$A$3 33 Kr-85 0.000475 0.000475 =C331SAS3 =0331$A$3 34 Kr-87 0.379 0.379 :C341SAS3 =D341$A$3 35 Kr-88 0.389 0.389 =C3SI$AS3 =D35I$A$3 36 Kr-89 1.62 1.62 =C36J$A$3 =D361SAS3 37 Xe-131M 0.000388 0.000388 =C371SAS3 =D371$A$3 38 Xe-l33M 0.00581 0.00581 =C38/SAS3 =D381$A$3 39 Xe-133 0.163 0.163 =C391$A$3 =0391$A$3 40 Xe-135M 0.476 0.476 =C4OI$A$3 :D4OISAS3 41 Xe-135 0.439 0.439 =C411$A$3 =D411$A$3 42 Xe-137 2.14 2.14 =C421SAS3 =D421SAS3 43 Xe-l38 1.62 1.62 =C43I$AS3 =D43I$A$3 44 45 Curies Released 46 to the Environment 47 isotoPe Case 1 ease 2 DCF1 CR 48 1-131 =G18 =H18 32900 =11ME48"$A$1 00*$A$S 49 1-132 =G19 =H19 381 =119"$E49"$A$1 OO"$A$5 50 1*133 =G2O =H20 5846 =120"$ESO"$A$100"$A$5 51 1-134 =G21 =H21 131 =121OSE51"$A$ l000SASS 52 1-135 =G22 "'H22 1230 =12T$E52"$A$100"$A$S CaJculallon e-023. Rev. 2 AlIachmentA

  • Dose Calculalioo PageA40f A15

A B' C 0 E F 53 54 Cs-134 =lJ5 =M35 =(37000oooooo '0.0oo00125 - SCS4I$A$3 *$E54'$A$l00'$A$5 55 Cs-135 =136 =M36 - 3700000000000 '0.00ooooo123 - 'SC55I$A$J *$ESS*SA$l00*$A$S 56 Cs-l36 =l37 =M37 - 370ooo00ooo *0.00ooooo198 - SC56I$A$J *$E56*$A$l00*$A$S S7 CS-137 "L38 =M311 = 37oooooooo0 *0.0oo000863 - $C571$AS3 *$ES7*$A$1 00*SA$5 56 Cs-l38 =l39 =M39 = 3700000000000 *0.0oo0oo274 =(SC58I$A$3)*SE58*$A$100*$A$5 59 Sub-totallrem ceDE -SUMIF48:F58) 60 61 Curies Released 62 to the Environment 63 tsotooe Case 1 ease 2 DCP CR 64 1-131 =C48 =048 0,06734 =118*$E64*$A$101*$ASS 65 1-132 =C49 =049 0.4144 =119*$E65*$A$101*SASS 66 1-133 =C50 =050 0.10878 ='20*$E66*$ASl 01 *SAS5 67 1-134 =C51 =051 0.481 =121*SE67*$A$1 01*$A$5 68 1-135 =C52 =052 0.29526 =122*$E68*$A$101*$A$5 69 70 Cs-134 =l35 =M35 - 3700ooooo000 *0.OOOOOOOOOOOOO757 =rsC701$A$J *SE70*SA$101*$A$5 71 Cs-135 =L36 =M36 - 3700000000000 *5.65E-19 - SC711$A$3 *$E71 *$A$101*$A$5 72 Cs-136 =L37 =M37 - 370000ooooo0 *O.OOOOOOOOOOOOl06 = SC721$A$3 *SE72*SA$101*SA$5 73 CS-137 =L38 =M38 I- 3700000000000 *7.74E-18 = SC731$AS3 *$E73*$A$101*$A$5 74 Cs-138 =l39 =M39 = 3700000000000 *O.OOOOOOOOOOOO121 = $C74J$A$J *$E74*$A$101*$A$5 75 76 Sub-totallrem EOE) =SUM7F64:F74) rr IOdine and Cesium Total -SUMlf59+F76) 78 Curies Released 79 to the Environment 80 Case 1 Casa2 ocP' CR 81 Kr-83M =CJl =031 0.OO0555 =E31 *$E81 *SAS101*$A$5 82 Kr-85M 0.122 0.122 0.027676 -E32*$E82*sA$ 101*SASS 83 Kr-85 0.000475 0.000475 0.0004403 =E33*SE83*SAS101*SA$5 84 Kr-87 0.379 0.379 0.15244 =E34*$E84*SA$ 101*$A$5 as Kr-88 0.389 0.389 0.3n4 =E35*$E8S*$A$101 *$A$5 86 Kr-89 1.62 1.62 0 =E36*$E86*$A$101 *$A$5 87 Xe-131M 0.000388 0.000388 0.0014393 =E37*SE87*SA$101*$A$5 88 Xe-133M 0.00581 0.00581 0.005069 =E38*$E88*SA$101 *$ASS 89 Xe-133 0.163 0.163 noosrra =E39*SE89*$A$101*$A$5 90 Xe-135M 0.476 0.476 0.07548 =E40*SEOO*$A$101*$A$S 91 Xe-l35 0.439 0.439 0.04403 =E41*$E91*$A$101*$A$S 92 Xe-137 2.14 2.14 0 =E42*SE~101*$A$5 93 Xe-l38 1.62 1.62 021349 =E43*$E93*$A$101"SA$5 94 It Gas Sub-total (rem =s0,,",81:F93f 95 96 Overall Total (rem TEDE =SUIIfF77+f'M) 97 98

  • Dose Converslon FllClor 99 2 Dose ConvenlionFllClor 100 O.000J.41 Brealhlng rsI8 (m'l8econ 101 =($A$8"O.338J/1113 ConlnlI Room 102 36.1 EAB .... (meters) for F sta 103 129.8 LPZ .... (rnelln) for F alai 1041 Wmd Soeed (mIs\

Calculation c-ozs. Rev. 2 Allachment A- Dose C8IaJIation Page AS of A15

A c o E F 105 =O.01331A$102lAS104 1 =0.01331AS103lAS104 calculation C-023. Rev. 2 AlIachment A . Dose Calculation PageA60f A15

G I H I I I J I K u, Reactor Coolant at maximum value (DE 1-131of 0.2 uCUg) permi1ted for CXIfltinued fuU poww operation 2

Reactor Coolant at maximum value permitted (DE 1*131 of 4.0 uCVg) corresponding to an assumed pr&;ilCCident spike

~4 5

6 7

6 ..

9 10 11 12 Case 1 Casa2 13 Release Release 14 Case 1 Casa2 Cloud Cloud 15 Activity AclMty Concentration Concentration Decav 16 Release Release Constant 17 Cl CI CIlm3 ClIm3 1/seconds 18 = SC$181C18 'E18~10CKk)00 =G1S'20 =G181$A$3 =H181$A$3 =lN2 8.04"86400) 19 = $C$181C19 'E19'$A$4/1000000 =G19"20 =G191$A$3 =H191$A$3 =lN2 2.3"3600) 20 = SC$181C20 'E20"$A$4/1000000 ==G20'2O =G201$A$3 =H201$A$3 =lN2 20.8"3600) 21 == $C$181C21 "E21 "$A$4I1000000 =G21"20 =G211$A$3 =H211$A$3 =lN2 52.6"60) 22 :: '$C$181C22 "E22"$A$4ll000000 =G22"20 =G22J$A$3 =H221$AS3 =lN2 6.61"36(0) 23 Totals 24 25 26 27 26 29 30 31 Case 1 Case 2 32 Activity AclMty Decay 33 Release Release Constant 34 MoIarFrac. moles moles l/seconds 35 Cs-l34 0.044317152955112 =O.95*SH35"LS23 =O.95"SH35*M$23 =lN2 2.062"86400"365.25) 36 Cs-135 0.174506296053598 =O.95*SH36"LS23 =O.95"$H36"MS23 =LN2 2300000"86400"365.25) 37 Cs-l36 0.000119942189253291 =O.95*SH3rLS23 =0.95"SH37"MS23 =LN 2 13.16"86400) 36 C5-137 0.401736793048373 =O.9S"SH36*LS23 =O.95"$H38"M$23 cLN2 30.17"86400"365.25) 39 Cs-l38 0.000101901239392202 =O.95*SH39"L$23 =O.95"$H39"MS23 =lN2 32.2"60) 40 Totals =SUM(H35:H39) =SUM(l35:I39) =SUM(J35:J39) 41 Balance Is stable Cs-133 42 43 44 45 Case 1 Dose (rem CEDE Case 2 Dose (rem CEDE\

46 (Inhalation) (Inhalation) 47 EAB LPZ CR EAB LPZ 46 =C46*SE46"$A$100*SA$10S =C48"SE46"$A$100*SA$106 =J18*SE48*$A$100*$A$5 =D48"SE48"$A$l OO*$A$l 05 =048"$E46"$A$l00"$A$l06 49 =C49*SE49"SA$1 OO"SASl 05 =C49"SE49"$A$100'$A$106 =J19"$E49*SAS100'SASS =D49"SE49*SAS1OO*$A$105 =D49*SE49'$AS100"$A$106 SO =CSO*SESO'$A$100*$A$10S =C5O"$ESO"$A$100*SAS106 =J20*$ESO*SAS100*SAS5 =DSO"SESO*SAS1OO*SA$l05 =DSO*SESO*$AS100"SA$106 51 =C51*SE51'SAS100*$A$105 =C51*$E51"$A$100*SAS106 =J21*SE51 *SASl00*SASS =D51"$ES1*SAS100*$A$105 =051 *$E51*SA$100*$A$106 52 =C52*SE52*$A$100*$A$105 =C52*$E52*W100*SA$106 =J22*$E52*SAS100*$A$5 =D52*$E52*$AS10lrSAS105 1OO"SAS1 06 Calculation c-oza, Rev.2 AIlachmentA

  • DoseC8IcuIalIon PageA70f A15

G H t J K 53 54 =C54'SE54'SA$100'SAS10S =C54'SE54-sA$100'$A$106 - SD54/$A$3 'SE54'$A$loo-sA$5 =D54'SE54'SA$100'SA$105 -054'SE54'SA$1OO'$A$106 SS =CSS'SESS'SA$ loo'$AS10S =C5S'SESS'$A$100~106 - SD551$A$3 'SESS'$A$l00'SAS5 =DSS'SE55'$A$1OO"sA$1OS =055'SE55'$A$100'SA$106 56 =CS6'SE56'$A$l OO'$ASI OS =C56"SE56'$A$1 OO'SA$l06 = S056/$A$3 'SE56'$A$100"SAS5 =0560SE56'$A$1 OO"SAS lOS =056'SE56'SAS100'W106 S7 =CS7'SESrSA$1 OO'SASI OS =CS7"SES7'SA$I00"$AS106 = SD571$A$3 'SES7'SA$100'SASS =OSnEsrSAS100"SAS t 05 =OSrSESnAS l00"SA$106 SS -CS8"SESS'SA$100'$AS10S =Csa'SE58'$A$100'$A$106 = SD58I$A$3 'SESS'$A$l00-sASS =DS8'SESS'SA$1 OO'$A$l05 =OSS'SESS'SA$100'$AS106 59 -SUMIG48:GS8} -SUMlH48:H58) =SUMII48:I58) -SUMI.uB:J58) -SUMIK48:K58) 60 61 Case 1 Dose (rem EDE) Case 2 Dose lrem EDE) 62 IExternal) (External) 63 EAB lPZ CR EAB lPZ 64 =C64'SE64'SA$10S =C64'SE64'SA$l06 =J18'SE64'SA$101'$A$S =D64'SE64'$AS105 -CH>4"SE64'SAS106 65 =C6S'$E6S'$A$10S =C65'SE6S'$A$106 =J19'SE6S'SA$101'SA$S =D65'SE65'$A$l05 =D65'$E6S'$A$106 66 =C66'SE66'SAS10S -e66"SE66'$A$l06 =J20'SE66'$A$101'SA$S -D66'$E66'$A$I05 =D66'SE66'$A$106 67 =C67'SE67'SA$10S =C6NE67'SA$106 =J21'SE6rSA$101'SASS =D6rSE6rSAS105 =067"SE67~106 68 "'C68'$E68'$A$105 =C68'$E68'$A$l06 =J22'SE68'$A$101'$A$5 =D68'$E68'$A$1 05 =068'$E68'$A$106 69

[ 70 =C70'SE70'SA$10S =C70'SE70"$A$106 = S070ISA$3 'SE70'SAS101'SA$S =D70'SE70'$A$I05 =D70'SE70'$A$106 71 =C71'SE71'SA$10S =C71'SE11'SA$106 = $0111$A$3 'SE11'SA$101'$A$5 =Dn'SE11'$AS105 =D71'SE71'$A$106 72 =C72'$E72'SA$10S =Cn'SE72'$A$l06 = SDl2i$A$3 'SE 12'SAS 101'SASS =D72'SE12'$AS105 =D72'$E12'$A$l06 13 =C13'$E13'SA$10S =C13'SE13'SAS106 = S0731$A$3 'SE13'SA$l 01'$ASS =073"SE73'$A$105 =073'SE73'SA$106 74 =C74'$E74'$A$105 =C74'$E74'$A$106 1= S0741$A$3 -sE74'$AS101'$A$5 =074'$E74"$A$105 =D74'$E74"$A$106 7S 76 =SUM(G64:G741 =SUMIH64:H74) -SUM(I64:174) tsSUM(J64:J74l -SUMIK64:K74\

17 -SUM(G59->G76) -SUM(H59+H76\ -SUM(I59+t76) -SUMIJ59+J761 =SUMlK59+K761 78 Case 1 Dose (rem EOE) Case 2 Dose (rem EOEI 79 (External\ (Externall 80 EAB lPZ CR EAB lPZ 81 =CSI 'SE81'SASI 05 =C81'SE81 'SA$106 =F31'SE81 '$A$101'$A$5 =D81'SE81'SA$105 =081 'SE81'SASI 06 82 =CB2'$E82'$A$10S =C82'SE82'SAS106 =F32'SE82'SA$101'$ASS =D8T$E82'$A$105 =D82'SE8TSA$106 83 =CS3'SES3'SA$10S =C83'SE83'$AS106 -F33'SE83'$ASI01 '$A$5 =083'SE83'$AS10S =D83'SE83"SAS106 84 =C84"SE84'$A$10S -e84'$E84 '$A$106 =F34'SE84'$A$101'$A$S =D84'$E84'$AS105 =D84"SE84'$A$106 8S =CBS'$E8S'$A$10S =C85'$E8S'SA$106 =F3S'SESS'SAS101'SASS =D8S'$ESS*$A$10S =08S'SE85'$A$106 86 =CB6'$E86'$A$105 =C86'SE86'SA$106 =F36'SE86'$A$101'$ASS =D8nE86'SAS105 =D86"$E86'$A$l06 87 =CBrSESr$A$10S =C87"SE87'$A$106 =F37"SE87'$A$1 01'$ASS =D8nES7'$AS10S =D87"SE8rsA$106 88 =C88'SE88'SA$l05 =C88'SE88'$A$l06 =F38'SE88'$ASI 01'$ASS =088'SE88'$AS105 =D68'SE88'SAS106 89 =CB9'SE8a-$A$10S -C89'SE89'$A$l06 =F39"SE89'$A$101'$ASS =D89'SE89-sA$105 -089"$E89"$A$106 90 =C90'SE90'SAS10S =C90"SE90'SA$106 =F40'SE90'$A$101 '$ASS =D90'$E90'$A$105 -090'SE90'$A$106 91 =C91'SE91 '$A$1 05 =C91'SE91'SA$106 =F41'SE91'$A$101'$ASS =D91'$E91'$AS105 -091'SE91"$A$106 92 "C92'SE92'$AS10S =C92'$E92'SA$106 =F4TSE92'$AS101"SAS5 =09T$E92'$A$105 :D92'SE92"SAS106 93 =C93'$E93"sA$10S =C93'SE93"SA$106 =F43"SE93'SAS101~ =D93'$E93'$A$105 =093"SE93"$AS106 94 =SUMiG81:G93) -SUM(HB1:H93) 1:193) =SUIIIJa1:J93f =SUMlK81 :JU3l 95 96 -SUM/G17+G94) aSUMlH17+H94) I&SUM017+t94) -SUMlm+.J94} =S(JMlK77+K94) 97 98 99 100 101 102 103 104 CaIcuIa1ion C-023. Rev. 2 AlIaChment A* Dose Cak:ulation PageAS Of A15

iE-,---------~----------+--'-------------------------------

G H I. J K CalaJ1ation e-023. Rev. 2 AItachment A

  • Dose CalalIation Page A9 of A15

L M 1

2 3

4 5

6 7

8 9

10 11 12 13 14 Case 1 Case 2 i 15 Activity Activity 16 Release Release 17 moles moles 18 =G18*37oo0000000l$K1816.023E+23 =H18*37000000000/$Kl816.023E+23 19 =G19*37oo0000000/$Kl916.023E+23 =H19*37000000000J$K19J6.023E+23 20 =G20*37000000000/$K2016.023E+23 =H2O*37000000000l$K20!6.023E+23 21 =G21*370000000001$K2116.023E+23 =H21*37000000000/$K21J6.023E+23 22 =G22*37000000000/SK221.023E+23 =H22*37000000000l$K22l6.023E+23 23 =SUM(L18:L22) =SUM(M18:M22) 24 25 26 27 28 29 30 31 Case 1 Case 2 32 Activitv Activitv 33 Release Release 34 curies curies 35 =135*6.023E+23*$K35I37000000000 =J35*6.023E+23*$K35137000000000 36 =l36*6.023E+23*$K36/3700c00OOOO =J36*6.023E+23"$K36I37000000000 37 =137*6.023E+23*$K37137000000000 =J37*6.023E+23*$1<37137oo0000000 38 =138*6.023E+23*$K38137000000000 =J38*6.023E+23*$K38137oo0000000 39 al39*6.023E+23*SK39137000000000 ....39*6.023E+23*$K39137000000000 40 41 42 43 44 45 46 47 48 49 50 51 52 Calculation e-023. Rev. 2 Attachment A

  • Dose Calculation Page A10 of A15

L M 53 54 55 56 57 58 59 60 61 62 63 64 65 66 67 68 69 70 71 72 73 74 75 76 n

78 79 SO 81 82 83 84 85 86 87 88 89 90 91 92 93 94 95 96 97 98 99 100 101 102 103 104 CaJculatIon e-023. Rev. 2 AlIachmenl A* DoseCaIallalion Page A11 of A15

.~---~-- ----

L M if--------------+-------------

Calculallon e-023, Rev. 2 Allachment A - Dose Calculallon PageA12 of A15

A I B I C I 0 I E I F I G H I I I J K L 1 Peach Bottom Beginning of Core ute 1100 Effec:tIYeFuU ' " - Days. and End of Cycle EOC) Cesium Isotope Qu8ntItie8 2 !IU58d for General Ca Molar Frllc:tion Detennination for ASn 3 Decay 4 100EFPO EOC 100 EFPO EOC Constant 100EFPO EOC 5 (grams) (grams) At. Mass (gnHnOlea) 1/sec:onda Ci Ci 6 Cs-133 1.025E+05 1.678E+05 Cs-133 132.9054 7.712E+02 1.263E+03 O.OOOE+OO O.OOOE+OO O.OOOE+OO 7 Cs-l34 1.031E+04 1.977E+04 Cs-134 133.9067 7.699E+01 1.476E+02 1.07E-oa 1.33SE+07 2.559E+07 8 Cs-135 4.502E+04 7.841E+04 Cs-135 134.9059 3.337E+02 5.812E+02 9.SSE-15 5.188E+01 9.03SE+01 9 Cs-137 1.087E+05 1.832E+05 Cs-137 138.9071 7.940E+02 1.338E+03 7.28E-10 9.410E+06 1.586E+07 10 Cs-138 2.37E-01 3.99E-01 6.10E-oT 2.352E+06 3.964E+06 11 Cs-138 2.01E-01 3.39E-01 3.59E-04 1.176E+W 1.982E+W 12 Total 2.&65E+05 ....92£+05 U76E+03 3.331E+03 13 14 ANSI/ANS-18.1-1999 Relative Abundances In Reactor Water Molar FI'IlCtiOll 15 uCUaramof moIesIQI'3ITl of ratio to Cs-l33 39.0219% 37.9218%

16 ReactorCool< ReactorCoolan Cs-13T Cs-l34 3.8956% *.4317%

17 Cs-l34 3.00E-05 1.04E+08 2.56E-02 Cs-.135 16.8848% 17.4506%

18 Cs-136 2.00E-Q5 1.21E+06 2.99E-04 Cs-137 40.1755% 40.1737%

19 Cs-137 8.00E-QS 4.07E+09 1.00E+OO Cs-136 0.0120% 0.0120%

20 Cs-l38 1.00E-02 1.03E+06 2.54E-04 Cs-138 0.0102% 0.0102%

Calculation c-023. Rev. 2 AttachmentA - DoseCalculation Page A13 of A15

A B C 0 E F G H I 1 Peac:hBot 2 Used for 3

4 100EFPO EOC 100EFPO eoc 5 (grams) 1_1 ALMau (allHllOlN) 6 C&-133 102500 167800 C&-133 132.llO54 n1.2 1263 7 Cs-134 10310 19nO C&-134 133.9067 76.99 147.6 8 C&-135 45020 78410 C&-135 134.9059 333.7 581.2 9 Cs-137 108700 183200 C&-137 136.9071 794 1338 10 C&-136 :f<10*370000000001$J1OJ6.023E+23 =-l10*37oo0000000J$J10J6.023E+23 11 C&-138 :f<11*37000000000J$J1116.023E+23 "l11*370000000001$J1116.023E+23 12 Total =SUM(B6:B9) -SUM/C8:C9) -suM(H6:H11) -slJM(16:t11) 13 14 ANSVANS-15 uCUgramof molesIwam of ratio to 16 Reactor CooIar Reactor Coolant C&-137 17 C&-134 0.OO3 =817"37000JJ7 =C171C$19 18 Cs-136 0.00002 =B18*37000JJ10 =C181CS19 19 Cs-137 0.00008 =819"37000lJ9 =C19JCS19 20 C&-138 0.01 =B20*370001J11 =C201C$19 calculatione-023. Rev. 2 AIIadunent A* DoseCalculation PageA14 of A15

J K L 1

2 3 Decal' 4 Constant 100 EFPD EOC 5 1/sec:onda CI CI 6 0 =H6*$J6'6.023E+23137000000000 =16*$J6*6.023E+23137000000000 7 =LN 2'1' 2.062*86400*365.25) =H7*$J7*6.023E+23137000000000 =l7*$J7*6.023E+23137000000000 8 =LN 2'1 2300000*86400*365.25) =H8*$J8*6.023E+23137000000000 =18*$J8*6.023E+23137000000000 9 =LN 2\J 30.17*86400*365.25) =H9*$J9*6.023E+23137000000000 =19*$J9*6.023E+23137000000000 10 =LN 2)1 13.16*864(0) =K$9*$B$1~19 =L$9"$B$181$B$19 11 =LN 2)1 32.2*60) =K$9*$B$2OJ$B$19 =1.$9"$8$201$8$19 12 13 14 Molar Frae::tion 15 Cs-133 =H6IH$12 =1611$12 16 Cs-134 =H7/H$12 =1711$12 17 Cs-135 =H8IH$12 -1811$12 18 Cs-137 =H9IH$12 =19/1$12 19 Cs-136 =H101H$12 sll()(1$12 20 Cs-138 =Hll/H$12 =11111$12 CalaJIaUon C-023. Rev. 2 Attachment A* Dose Calculation PageA15 of A1S

I=====:::::::::::::::::::=====:=:::::::::====================-====-_I....

CALCULAnON NO. C-023 I PAGE NO. Bl of al j EV._NO,;"";;;,,;,,,._2_ _......=.,;;",;",;;;,--.,;,;.....;;;.;;;"".,;;;,,;~

R.....

Computer Disclosure Sheet Discipline Nuclear Client:: Exelon Corporation! Amergen Date: January 2006 Project: Clinton Power Station MSL8 AST Job No.

Program(s) used Rev No. Rev Date Calcu~ation Set No.: C-OZ3. Rev. 2 Attachment A spreadsheet NJA NlA .

Status [ ) Prelim.

[X] Rnal

[ ] Void WGI Prequalification [ ] Yes rX] No Run No.

Description:

Analysis

Description:

Spreadsheet used to perform dose assessment for MSlB. as described in calcuJation.

The attached computer output has been reviewed, the inp~ data checked, And the results approved for release. Input criteria for this analysis were established.

By: On: January 2006 Run by: H. Rothstein 7l~

Checked by: P. Reichert /.I?A-Jdo ~

Separate cell*by-<:elllndependenl check by: A. Boatright ~~

Approved by: H. Rothstein if-~

Remarks: WGI Form for CompUter Software Control This spreadsheet is relatively straight-forward and w~s nand checkiKt. Attachment includes the spreadsheet in both normal and formula dlsotav mode and so is comolelelV documented. A seD8rate:~Y~Jl indeDendent check was also oedonned.

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