ML042020294

License Amendment Request NPF-38-256, Alternate Source Term

ML042020294
Person / Time
Site:	Waterford
Issue date:	07/15/2004
From:	Venable J Entergy Nuclear South
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
W3F11-2004-0053
Download: ML042020294 (77)

Text

Entergy Nuclear South Entergy Operations, Inc.

17265 River Road Killona, LA 70066

'-Entergy Tel 504 739 6660 Fax 504 739 6678 ivenabllentergy.com Joseph E. Venable Vice President, Operatons Waterford 3 W3Fl-2004-0053 July 15, 2004 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555

SUBJECT:

License Amendment Request NPF-38-256 Alternate Source Term Waterford Steam Electric Station, Unit 3 Docket No. 50-382 License No. NPF-38

REFERENCES:

1. Entergy LetterW3Fl-2003-0074, "License Amendment Request NPF-38-249, Extended Power Uprate," November 13, 2003

2. Entergy Letter W3Fl-2004-0035, Supplement to Amendment Request NPF-38-249, Extended Power Uprate," dated May 7, 2004

Dear Sir or Madam:

Pursuant to 10 CFR 50.90, Entergy Operations, Inc. (Entergy) hereby requests approval of an amendment for Waterford Steam Electric Station, Unit 3 (Waterford 3). Entergy intends to implement an Alternate Source Term (AST) as permitted by 10 CFR 50.67 for calculating accident offsite doses and doses to control room personnel. Relevant information is provided below and in attachments to this letter. Dose consequence analyses have been performed using the AST; the analyses and results are described in Attachment 2. The AST and methodology described in NUREG-1465, "Accident Source Terms for Light-Water Nuclear Power Plants," and in Regulatory Guide 1.183, "Alternative Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors," are employed in the dose analyses.

Reference 1 requested a license amendment for Waterford 3 to increase the unit's rated thermal power from 3441 MWt to 3716 MWt. Attachment 5 of Reference 1 is the Extended Power Uprate Report (PUR) which contains evaluations that support operation at the increased power level. Consistent with the existing Waterford 3 licensing basis, the PUR presents doses to control room personnel for only the Large Break Loss of Coolant Accident (LBLOCA) and Fuel Handling Accident (FHA). PUR Section 2.5.3.1, Control Room Habitability, states that control room doses for these and other accidents would be addressed as necessary in conjunction with the response to Generic Letter 2003-01, Control Room Habitability. Reference 2 reiterated our plan to submit this information in response to an NRC question. Entergy has decided to implement an AST for calculation of accident doses to control room personnel and, in addition, to apply the AST to the calculation of offsite accident doses.

W3F1I-2004-0053 Page 2 of 4 This submittal presents the calculated dose results for the following seven events that are expected to produce the most limiting dose consequences for Waterford 3:

• Large Break LOCA

• Small Break LOCA

• Main Steam Line Break Inside Containment

• Steam Generator Tube Rupture

• Outside Containment Main Steam Line Break / Feedwater Line Break

• Control Element Assembly Ejection

• Fuel Handling Accident A second supplemental AST submittal will present calculated dose results for the following events:

• Reactor Coolant Pump Seized Rotor / Sheared Shaft

• Inadvertent Atmospheric Dump Valve Opening

• Excess Main Steam Flow with Loss of Offsite Power

• Letdown Line Break The proposed change has been evaluated in accordance with 10 CFR 50.91 (a)(1) using criteria in 10 CFR 50.92(c) and it has been determined that this change involves no significant hazards considerations. The bases for this determination are provided by attached evaluations of the seven events identified above.

Attachment I describes the proposed application of the AST and includes the No Significant Hazards determination. Attachment 2 provides a description of the analyses and dose consequences of the above identified seven events.

The proposed change includes a new commitment as summarized in Attachment 3. AST has been approved for other nuclear plants. This submittal is modeled after the DC Cook submittal that was approved by the NRC on November 14, 2002.

Entergy has been informed by the NRC that resolution of the control room habitability in-leakage issue should be considered in the acceptability of the Extended Power Uprate. Waterford 3 has performed a tracer gas test to determine the unfiltered in-leakage. The analyses using the AST described herein also incorporate the assumption of the increased in-leakage and demonstrate acceptable results. Entergy requests approval of this amendment request by April 1, 2005, so that it can be implemented prior to restart from refueling outage 13 in the spring of 2005 in order to update the design assumption regarding in-leakage, resolve concerns identified in Generic Letter 2003-01, and support the power uprate implementation.

W3F11-2004-0053 Page 3 of 4 If you have any questions or require additional information, please contact Jerry Burford at 601-368-5755.

I declare under penalty of perjury that the foregoing is true and correct. Executed on July 15, 2004 Sincerely, JEV/fgb Attachments:

1. Analysis of Proposed Change

2. Licensing Report for the Radiological Consequences of Accidents for the Waterford Steam Electric Station, Unit 3 Using Alternative Source Term Methodology

3. List of Regulatory Commitments

W3F11-2004-0053 Page 4 of 4 cc: Dr. Bruce S. Mallett U. S. Nuclear Regulatory Commission Region IV 611 Ryan Plaza Drive, Suite 400 Arlington, TX 76011 NRC Senior Resident Inspector Waterford 3 P.O. Box 822 Killona, LA 70066-0751 U.S. Nuclear Regulatory Commission Attn: Mr. Nageswaran Kalyanam MS 0-07D1 Washington, DC 20555-0001 Wise, Carter, Child & Caraway Attn: J. Smith P.O. Box 651 Jackson, MS 39205 Winston & Strawn Attn: N.S. Reynolds 1400 L Street, NW Washington, DC 20005-3502 Louisiana Department of Environmental Quality Office of Environmental Compliance Surveillance Division P. 0. Box 4312 Baton Rouge, LA 70821-4312 American Nuclear Insurers Attn: Library Town Center Suite 300S 29th S. Main Street West Hartford, CT 06107-2445

Attachment I W3F1 -2004-0053 Analysis of Proposed Change to W3F1-2004-0053 Page 1 of 5

1.0 DESCRIPTION

This letter is a request to amend Operating License NPF-38 for Waterford Steam Electric Station, Unit 3 (Waterford 3).

The proposed change will implement an alternative source term (AST) for determining accident offsite doses and accident doses to control room personnel. The AST is being adopted principally as part of Entergy's response to Generic Letter 2003-01, Control Room Habitability. of this submittal presents the results of dose analyses for the following events that are expected to produce the most limiting dose consequences:

• Large Break LOCA (LBLOCA)

• Small Break LOCA (SBLOCA)

• Main Steam Line Break Inside Containment

• Steam Generator Tube Rupture

• Outside Containment Main Steam Line Break / Feedwater Line Break

• Control Element Assembly Ejection

• Fuel Handling Accident A second AST submittal will be made to supplement this submittal. It will present the dose consequences of the events below. The dose consequences of these events are expected to be bounded by the results of those analyses described in Attachment 2.

• Reactor Coolant Pump Seized Rotor / Sheared Shaft

• Inadvertent Atmospheric Dump Valve Opening

• Excess Main Steam Flow with Loss of Offsite Power

• Letdown Line Break

2.0 PROPOSED CHANGE

This change in the Waterford 3 licensing basis involves the adoption of an AST for calculating accident doses to control room personnel and offsite receptors.

Although this request includes no Technical Specification revisions, the evaluation of the SBLOCA assumes a primary-to-secondary leakage of 75 gallons per day (gpd) per steam generator. The current Technical Specification limit is 720 gpd through any one steam generator. Reference 1 proposed revising that value to 540 gpd. Following discussions with the NRC staff, Entergy committed to reduce this value to 150 gpd but now plans to further reduce it to the 75 gpd value used in the analysis. This new value is proposed in Reference 4 which revises the proposed change submitted in Reference 1.

3.0 BACKGROUND

Entergy submitted a license amendment request to increase the rated thermal power of Waterford 3 from 3441 MWt to 3716 MWt (Reference 1). Attachment 5 of Reference 1 is the Extended Power Uprate Report (PUR) which contains evaluations that support operation at the to W3Fl-2004-0053 Page 2 of 5 increased power level. Control Room doses for the LBLOCA and FHA events, which are the only events in the current Waterford 3 licensing basis for which control room doses are currently documented in the UFSAR, were addressed in the PUR. PUR Section 2.5.3.1, Control Room Habitability, states that control room doses for these and other accidents would be addressed as necessary in conjunction with the response to Generic Letter 2003-01, Control Room Habitability.

Control room tracer gas testing in response to Generic Letter 2003-01 has now been performed.

Based on the results, Entergy has decided to implement an AST in the calculation of accident doses to control room personnel. The revised dose consequence analyses have been performed for several event scenarios, including the LBLOCA and FHA events described above in the PUR. The AST will also be adopted for calculating offsite accident dose consequences.

4.0 TECHNICAL ANALYSIS

The AST and methodology described in NUREG-1465, "Accident Source Terms for Light-Water Nuclear Power Plants" (Reference 2) and in Regulatory Guide 1.183, "Alternative Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors" (Reference 3), provide regulatory guidance for the implementation of the AST. Revision of a plant's licensing basis from the TID-14844 source term to an alternative source term involves the preparation of dose consequence analyses. Demonstration that the results satisfy the regulatory acceptance criteria and NRC approval of the requested change establishes the acceptability of the use of the AST.

Entergy has prepared the dose consequence analyses of the seven event scenarios outlined in Section 1.0. These analyses are summarized in Attachment 2. The analyses included evaluation of the worst 2-hour Exclusion Area Boundary (EAB) offsite doses, the duration Low Population Zone (LPZ) offsite doses, and control room doses. The results for all of the analyzed events meet the regulatory acceptance criteria of 10 CFR 50.67 and Regulatory Guide 1.183.

5.0 REGULATORY ANALYSIS

5.1 Applicable Regulatory Requirements/Criteria The proposed changes have been evaluated to determine whether applicable regulations and requirements continue to be met.

Entergy has determined that the proposed changes do not require any exemptions or relief from regulatory requirements, other than the existing Technical Specification steam generator leakage limit of 720 gpd for any one steam generator. As discussed in Section 2.0 above, Entergy now plans to change that limit to the value of 75 gpd per steam generator assumed in the SBLOCA dose consequence analysis presented in Attachment 2 to this submittal.

Compliance with General Design Criterion 19, Control Room, is demonstrated for the proposed change based on meeting the dose limit to control room personnel of 5 Rem TEDE. The original licensing basis had been established based on the whole body, thyroid, and skin dose limits now described in the Updated Final Safety Analysis Report (UFSAR) and the PUR. As stated in Regulatory Guide 1.183, the applicable acceptance criteria to establish compliance with GDC 19 for facilities licensed to use AST is the 5 Rem TEDE criterion of 10 CFR 50.67(b)(2)(iii). The proposed change is an operating license amendment to implement to W3Fl-2004-0053 Page 3 of 5 the new AST licensing basis and analysis demonstrates that Waterford 3 complies with this 5 rem TEDE requirement.

Similarly, the acceptance criteria for offsite doses are now proposed based on TEDE dose limits specified in 10 CFR 50.67 and Regulatory Guide 1.183.

5.2 No Significant Hazards Consideration As provided by 10 CFR 50.67, Entergy is implementing the AST and dose calculation methodology described in NUREG-1465 and Regulatory Guide 1.183 (References 2 and 3) to calculate accident doses to control room and offsite personnel following postulated events that result in the release of radioactive material from the reactor fuel. The AST and associated methodology define the amount, isotopic composition, physical and chemical characteristics, and timing of radioactive material releases following postulated events. Transport of the material to the control room and offsite is modeled, and the resulting Total Effective Dose Equivalent (TEDE) is determined. Regulatory acceptance criteria account for the sum of the deep-dose equivalent (for external exposures) and the committed effective dose equivalent (for internal exposures).

Entergy has evaluated whether a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92, Issuance of Amendment," as discussed below:

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

Response: No.

The use of an alternative source term is recognized in the NRC regulation 10 CFR 50.67; guidance for its implementation is provided in Regulatory Guide 1.183. The AST involves quantities, isotopic composition, chemical and physical characteristics, and release timing of radioactive material for use as inputs to accident dose analyses. As such, the AST cannot affect the probability of occurrence of a previously evaluated accident. No facility equipment, procedure, or process changes are required in conjunction with implementing the AST that could increase the likelihood of a previously analyzed accident. The proposed changes in the source term and the methodology for the dose consequence analyses generally follow the guidance of Regulatory Guide 1.183. As a result, there is no increase in the likelihood of existing event initiators.

Regarding consequences, the results of accident dose analyses using the AST are compared to TEDE acceptance criteria that account for the sum of deep dose equivalent (for external exposure) and committed effective dose equivalent (for internal exposure).

Dose results were previously compared to separate limits on whole body, thyroid, and skin doses as appropriate for the particular accident analyzed. The results of the revised dose consequences analyses demonstrate that the regulatory acceptance criteria are met for each analyzed event. Implementing the AST, however, involves no facility equipment, procedure, or process changes that could affect the radioactive material actually released during an event. Consequently, no conditions have been created that could significantly increase the consequences of any of the events being evaluated.

to W3Fl-2004-0053 Page 4 of 5 Therefore, the proposed change does not involve a significant increase in the probability or consequences of any of the events being evaluated.

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

Response: No.

The AST involves quantities, isotopic composition, chemical and physical characteristics, and release timing of radioactive material for use as inputs to accident dose analyses.

As such, the AST cannot create the possibility of a new or different kind of accident. No facility equipment, procedure, or process changes have been made in conjunction with implementing the AST that could initiate or substantially alter the progression of an accident.

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

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

Response: No.

Implementing the AST is relevant only to calculated accident dose consequences. The AST involves quantities, isotopic composition, chemical and physical characteristics, and release timing of radioactive material for use as inputs to accident dose analyses. The results of the revised dose consequences analyses demonstrate that the regulatory acceptance criteria are met for each analyzed event. No facility equipment, procedure, or process changes are required in conjunction with implementing the AST that could increase the exposure of control room or offsite individuals to radioactive material. The AST does not affect the transient behavior of non-radiological parameters (e.g., RCS pressure, containment pressure) that are pertinent to margin of safety.

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

Based on the above, Entergy concludes that the proposed amendment(s) present no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of 'no significant hazards consideration" is justified.

5.3 Environmental Considerations The proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure.

Section 5.2 above demonstrates that use of the AST does not involve a significant hazards consideration.

to W3FI-2004-0053 Page 5 of 5 Concerning the types and amounts of effluents that may be released offsite, the AST involves some changes in assumed quantities and characteristics of radioactive material that are inputs to offsite accident dose calculations. These are changes to calculation assumptions only. No facility equipment, procedure, or process changes are associated with use of the AST that affect actual releases Consequently, implementation of the AST will not increase the quantities or alter the types of radioactive material actually released if an event were to occur.

Implementation of the AST also has no effect on the actual or calculated effluents arising from normal operation.

With respect to occupational doses, the AST is, again, only a change in dose calculation inputs and methodology. Calculated doses meet TEDE criteria. No aspect of implementing the AST involves facility equipment, procedure, or process changes that would increase actual onsite doses if an event were to occur. The AST does not result in actual or calculated changes in the normal radiation levels in the facility, or in the type or quantity of radioactive materials processed during normal operation.

Accordingly, the proposed amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

6.0 PRECEDENCE Several plants have been approved for the use of the alternative source term, among them Grand Gulf Nuclear Station in March 2001, D. C. Cook in November 2002, and River Bend Station in March 2003. Those submittals had included various Technical Specification changes which were justified by the change to the AST. This Waterford 3 license amendment request does not involve any changes to the Technical Specifications; it does credit Technical Specification changes being proposed with the Extended Power Uprate license amendment request.

7.0 REFERENCES

1. Entergy Letter W3F1-2003-0074, 'License Amendment Request NPF-38-249, Extended Power Uprate," November 13, 2003

2. NUREG-1465, "Accident Source Terms for Light Water Nuclear Power Plants," L. Soffer, et. al., February 1995

3. Regulatory Guide 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors," USNRC, July 2000

4. Entergy Letter W3F1-2003-0052, "Supplement to Amendment Request NPF-38-249, Extended Power Uprate," July 14, 2004

Attachment 2 W3FI -2004-0053 Licensing Report for the Radiological Consequences of Accidents for the Waterford Steam Electric Station, Unit 3 Using Alternative Source Term Methodology

Licensing Report for the Radiological Consequences of Accidents for the Waterford Steam Electric Station, Unit 3 Using Alternative Source Term Methodology July 14, 2004 I

TABLE OF CONTENTS 1.0 RADIOLOGICAL CONSEQUENCES UTILIZING NUREG-1465 SOURCE TERMS ............... 3 1.1. Introduction ................................................................ 3 1.2. Common Analysis Inputs and Assumptions ................................................................ 3 1.3. Control Room Air Conditioning System and Control Room Ventilation Model ........... ............. 7

2.0 CONCLUSION

S ............................................................... 22

3.0 REFERENCES

............................................................... 23 4.0 POST-LOSS OF COOLANT ACCIDENT PH EVALUATION

SUMMARY

.............................. 24 5.0 LARGE BREAK LOSS OF COOLANT ACCIDENT (LBLOCA) .............................................. 26 5.1. Input Parameters and Assumptions ....................................................  ; 26 5.2. Results ................................................... 29 6.0 SMALL BREAK LOSS OF COOLANT ACCIDENT (SBLOCA) .............................................. 32 6.1. Input Parameters and Assumptions ................................................... 32 6.2. Results ................................................... 34 7.0 INSIDE CONTAINMENT MAIN STEAM LINE BREAK (IC-MSLB) ...................................... 38 7.1. Input Parameters and Assumptions ................................................... 39 7.2. Results ................................................... 40 8.0 STEAM GENERATOR TUBE RUPTURE (SGTR) ................................................... 44 8.1. Input Parameters and Assumptions ................................................... 45 8.2. Results ................................................... 49 9.0 OUTSIDE CONTAINMENT MAIN STEAM LINE BREAK/FEEDWATER LINE BREAK (OC-MSLB/FWLB) ........................................ 52 9.1. Input Parameters and Assumptions ........................................ 52 9.2. Results ........................................ 54 10.0 CONTROL ELEMENT ASSEMBLY (CEA) EJECTION ........................................ 57 10.1. Input Parameters and Assumptions .. 57 10.2. Results .. 59 11.0 FUEL HANDLING ACCIDENT (FHA) .62 11.1. Input Parameters and Assumptions .. 62 11.2.

. . Results....................................................................................................................................... 63 2

1.0 RADIOLOGICAL CONSEQUENCES UTILIZING NUREG-1465 SOURCE TERMS 1.1. Introduction The Waterford Steam Electric Station, Unit 3 (Waterford 3) licensing basis for the radiological consequences analyses for Chapter 15 of the Updated Final Safety Analysis Report (UFSAR) is currently based on methodologies and assumptions that are derived from Technical Information Document (TID) 14844 (Reference 1) and other early guidance.

NUREG-1465 (Reference 2) provides a postulated fission product source term that is based on the current understanding of light-water reactor (LWR) accidents and fission product behavior.

This is also referred to as Alternate Source Term (AST). Reference 2 is applicable to LWR designs and forms the basis for the development of Regulatory Guide (RG) 1.183 (Reference 3).

The new source terms as described in NUREG-1465 and RG 1.183 are being used to calculate the control room and off-site radiological consequences for the Waterford 3 plant to support the Extended Power Uprate (EPU) program. RG 1.183 guidance is used as the basis for the assumed source terms for the various events. To support the EPU, the following UFSAR Chapter 15 radiological consequences analyses will be analyzed or evaluated: Post-Loss of Coolant Accident (LOCA) pH Evaluation, Large Break LOCA (LBLOCA), Small Break LOCA (SBLOCA), Inside Containment Main Steam Line Break (IC-MSLB), Steam Generator Tube Rupture (SGTR),

Outside Containment Main Steam Line Break/Feedwater Line Break (OC-MSLB/FWLB),

Control Element Assembly (CEA) Ejection, and Fuel Handling Accident (FHA). Each accident and the specific input assumptions are described in detail in subsequent sections in this report. A supplemental Licensing Report will be prepared and submitted to address non-limiting events (Reactor Coolant Pump Seized Rotor/Sheared Shaft, Inadvertent Atmospheric Dump Valve Opening, Excess Main Steam Flow with Loss of Off-site Power, and Letdown Line Break) and to address aspects other than off-site and control room dose associated with the adoption of a RG 1.183 AST as the licensing basis for Waterford 3.

1.2. Common Analysis Inputs and Assumptions The analysis of design basis events for the 3716 MWt power uprate using AST included evaluation of the worst two hour Exclusion Area Boundary (EAB) and the duration 30 day Low Population Zone (LPZ) off-site doses, and evaluation of Main Control Room (MCR) doses. The Total Effective Dose Equivalent (TEDE) doses are determined for off-site dose and for control room personnel. The dose conversion factors used are from NUREG/CR-6604. The TEDE dose is equivalent to the Committed Effective Dose Equivalent (CEDE) dose plus the acute dose for the duration of exposure to the cloud.

The RADTRAD radiological computer code, described in NUREG/CR-6604, has been used to perform the analyses herein.

Because tracer gas testing to determine control room in-leakage had not been conducted, results necessary to quantify control room in-leakage were not available at the time the EPU Licensing Amendment Request was compiled. Therefore, only the MCR doses for LOCA and FHA, which are currently documented in the UFSAR, were addressed in the EPU Licensing Amendment Request (Reference 4). Subsequent performance of the control room tracer gas testing in response to Generic Letter 2003-01 (Reference 5) has resulted in the adoption of an assumed maximum unfiltered in-leakage of 100 CFM (recirculation mode) and 65 CFM (pressurized mode) for the Waterford 3 control room. The radiological analyses assume an unfiltered in-3

leakage of at least 100 CFM for the duration of the event, with the exception of SBLOCA. The SBLOCA analysis credits the reduced in-leakage associated with the pressurized mode of operation at two hours into the event.

This report documents off-site and control room dose consequences for the SBLOCA event. This event had not been previously documented in the Waterford 3 licensing basis. Consideration of a different release pathway applicable to the SBLOCA, that of secondary steaming via Atmospheric Dump Valves (ADVs), has led to the inclusion of this event in the Waterford 3 licensing basis.

Important inputs to the radiological consequence evaluations are provided in Tables 1-1 through 1-6. The Dose Conversion Factors (DCFs) used are per International Commission on Radiological Protection, Publication 30 (Reference 6) and are per NUREG/CR-6604 (Reference 7).

Technical Specification DEI-131 activity limits of 1.0 11Ci/gm for the Reactor Coolant System (RCS) and of 0.1 [tCi/gm for secondary system are assumed. Thyroid dose conversion factors per NUREG/CR-6604 are used in assessing the DEI-131 equivalent activity for the distribution of iodine isotopes. The Technical Specification activity limit of 100/E gCi/gm is assumed for noble gas activities. For releases that do not involve fuel failure, the RCS activity distribution of Table 1-4, adjusted to Technical Specification activity limits, is assumed.

The STGR analysis and the OC-MSLB/FWLB analysis calculate doses assuming a pre-existing iodine spike (PIS) and an accident generated iodine spike (GIS). The analyses assume a PIS spiking factor of 60 per Technical Specifications. The GIS spiking factor for SGTR is assumed to be 500, conservatively higher than the 335 factor specified in Appendix F of RG 1.183. The STGR analysis bases the GIS spike on a purification flow of 128 gpm, which corresponds to three charging pumps operating. The OC-MSLB/FWLB analysis bases the GIS spike on an assumed purification flow of 44 gpm, which corresponds to the normal operating plant condition of one charging pump operating. The acceptability of assuming one charging pump operating is discussed in the NRC Safety Evaluation Report (SER) concerning the letdown line break analysis (Reference 16). In the SER, the NRC noted that the assumption of three charging pumps operating is overly conservative for determining the equilibrium iodine appearance rate at Waterford 3. Therefore, the analysis assumed only one charging pump operating for determining the iodine appearance rate, which corresponds to normal plant operation.

Table 1-1 provides the assumed core inventories used in the LBLOCA analyses. This same source term was used for other cases involving the Reactor Building Pathway, including the releases to containment for CEA Ejection, SBLOCA, and for the IC-MSLB. Two separate source terms were computed using the ORIGEN2 code (Reference 14). One calculation was performed to determine the gap fission product activities in peak power rods. A second calculation was performed to determine the core-wide fission product inventory. There was generally good agreement between these two calculations, with their slightly different biases. A LOCA source term (Table 1-1) was constructed using the more conservative (larger) value of core inventory from the two sources. Several isotopes are modeled in RADTRAD for which inventories were not calculated in the ORIGEN calculations. For those isotopes, the default Pressurized Water Reactor (PWR) core inventories (on a CiIMWt basis) from NUREG/CR-6604 (Reference 7) were assumed. These source terms are valid for enrichments up to 5.0 w/o U-235, for fuel management such that the maximum bumup of any rod within 10% of the limiting radial peaking factor (F1) is less than 40,000 MWD/MTU, and for the standard CE 16x 16 fuel assembly design.

The values are multiplied by the appropriate 3-pin power F, required for the event analyzed. The 4

new source term development included the performance of sensitivity studies to identify the enrichment and burnup that provide worst case values.

For events that involved fuel failure and steaming releases, the source term of Table 1-1 .A was utilized. This is the source term developed to determine the gap fission product activities in peak power fuel rods. The source term for the Fuel Handling Accident is discussed in Section 11.0 of this report.

As discussed in the EPU Licensing Amendment Request, new sets of off-site and control room atmospheric dispersion factors (x/Qs) were calculated for Waterford 3 using the PAVAN and ARCON96 computer codes (References 10 and 11) and using five years (1997 through 2001) of current meteorological data. Additional information on the X/Q's was provided in a response to a Request for Additional Information on EPU, forwarded by Reference 17. The meteorological data was obtained from the storage data module from each of the meteorological towers and was verified and validated to be of good quality. The control room X/Q values for both control room emergency intakes (east and west) were calculated for the following potential release locations:

• Main Steam Safety Valves (MSSVs), east and west sides

• ADVs, east and west sides

• Plant stack

• Main steam lines, east and west sides

• Fuel building, truck bay door and personnel door

• Containment equipment hatch The control room X/Q calculation assumed a straight line from the release point to the control room intakes. This provides a conservative distance from the release points to the control room intakes. Also, in all cases, intervening structures were ignored for calculational simplicity, thereby underestimating the true distance to the control room intakes. Release point height above grade was considered. The release point to intake directions were calculated relative to plant north and were then converted to true north.

In the calculation of new base X/Q values, for all the release paths, the release type is assumed to be a "ground level" release with zero vent flow and zero vent velocity. For the plant stack release, the vent velocity is also assumed to be zero in accordance with RG 1.194 (Reference 8).

These X/Q values are presented in Table 1-3.

For events which experience violation of the Specified Acceptable Fuel Design Limits (SAFDLs),

the amount of fuel failure which results in doses equal to or less than the regulatory acceptance limits were calculated based upon the release path applicable to the event scenario. Cycle specific analyses performed during the reload core design process ensure these fuel failure limits are not exceeded.

For the accident sequences where the source term dose (off-site and MCR) is dominated by the secondary side steaming release pathway (e.g., CEA Ejection, SBLOCA, IC-MSLB), these analyses assume a constant primary-to-secondary side leak rate. The mass release rate corresponding to the primary-to-secondary leak rate is conservatively evaluated at 3501F conditions, corresponding to Shutdown Cooling entry conditions.

For events where the SG is not in a faulted condition, the pressure differential between the RCS and the secondary side across the SG tubes is about the same or less than that under normal operating conditions. Thus, for non-faulted SGs, when there is no large primary-to-secondary 5

pressure differential, the operational leakage value which will be incorporated into Technical Specifications is the appropriate primary-to-secondary leakage value to assume in radiological analyses. The actual leak rate during these transients would be the same or, because of reduced differential pressure between primary and secondary sides, less than that for normal operating conditions. Under certain conditions, secondary pressure could exceed primary pressure, resulting in backflow from the SG to the RCS. Each of the above scenarios have slightly different assumptions pertaining to the actual leak rate modeled, however at nominal RCS conditions a value of 150 gpd is assumed, except for SBLOCA where a proposed Technical Specification change rate of 75 gpd is assumed. For the MSLB or FWLB cases where a significant AP exists across the SG tubes, a larger (540 gpd) leakage rate is assumed. For more information on the fission product release mechanisms for the individual sequences, see Sections 6.0 through 10.0.

For secondary steaming cases, releases of iodines and alkali metals are based on the activity released to the environment due to steam releases from the ADVs. Appropriate Partition Factors (PFs) are applied to these releases, to account for partitioning effects as these isotopes pass through the secondary inventory of the SGs. The concentration of the release considers the effects of the primary-to-secondary leakage mixing with the SG secondary inventory. Noble gas activity transferred to the secondary side by primary-to-secondary leakage is assumed to be immediately available for release.

The events presented herein that involve assumed releases from the ADVs include SBLOCA, SGTR, CEA Ejection, IC-MSLB, and the combined OC-MSLBIFWLB scenario. For secondary steaming release scenarios other than SGTR and OC-MSLB/FWLB, it is assumed that operators act at 30 minutes into the event to begin plant cooldown to shutdown cooling entry conditions of 350 'F. This assumption is consistent with current licensing bases, as documented in Chapter 15 of the Waterford 3 UFSAR. The time to initiate cooldown for the SGTR scenario is described in Section 8.0. A maximum cooldown rate of 100 0F/hr has been assumed. This is the maximum allowed cooldown rate for forced circulation conditions and conservatively bounds the plant procedural limits of 50 0F/hr for natural circulation conditions. A rapid cooldown increases the releases early in the event, thus increasing the releases subject to the higher X/Qs assumed for the initial two hours of the event. Additionally, this maximizes the releases prior to the assumed operator action at two hours into the event to select the preferred air intake for control room pressurization flow.

The event duration for secondary steaming cases is assumed to be at least 7.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />. This would bound the time required for shutdown cooling initiation (and thus for terminating steaming releases from the SG for decay heat removal) based on a nominal cooldown rate of 40 'F/hr for natural circulation conditions. Thus, it is assumed that the plant is rapidly cooled to shutdown cooling entry conditions, but that plant conditions are maintained by steaming at those conditions for up to 7.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> into the event. These assumptions ensure that the steaming release bounds that required for decay heat removal as well as cooldown of the RCS, including structural latent heat.

The cooldown scenario for the OC-MSLB/FWLB is described in Section 10.0. It determines the releases from the intact SG for this scenario. The intact SG is a relatively small contributor to the releases for that event. As described in Section 10.0, the cooldown for that event is assumed to start at 4.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> and shutdown cooling initiation is assumed at 8.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> into the event.

The steam releases for each secondary steaming event are presented within the input table in each individual event section. Note that because similar cooldown scenarios are used, results depend 6

on whether or not a Loss of Off-site Power (LOOP) is assumed or on how many reactor coolant pumps are assumed operating during the cooldown.

Table 1-6 contains half-life and disintegration energy data for iodine and noble gas isotopes. This information was used to calculate individual isotope concentrations in the primary system at the Technical Specification limit of 10/WE.

Changes to the input parameters used in the analyses include:

• Previously the design pressurization flow rate of 200 CFM was used for the MCR. For the analyses herein, a maximum pressurization flow of 225 CFM is specified as the inflow, to conservatively increase intake to the MCR. A minimum pressurization flow of 50 CFM is specified, to minimize the flushing out of activity due to out-leakage. This assumption was adopted based on industry operating experience.

• The unfiltered in-leakage values assumed for the MCR have been increased to bound the results of the tracer gas testing conducted April 2004.

• SG primary-to-secondary leakage rate assumptions have been revised to assume an operational leakage value for intact SGs.

• In the LBLOCA analysis, Shield Building Ventilation System (SBVS) exhaust flow is now based on a maximum flow of 11,000 CFM; previously this was based on a nominal flow of 10,000 CFM.

• The sprayed fraction of containment has been conservatively reduced from 85% to 80%.

Also, the mixing rate between sprayed and unsprayed portions of containment has been reduced to that recommended in RG 1.183.

• Iodine spiking assumptions for the OC-MSLB/FNVLB now assume that normal iodine concentration is based on one charging pump in service, vice all three. This is the same assumption as used in the Letdown Line break analysis recently approved by NRC.

1.3. Control Room Air Conditioning System and Control Room Ventilation Model The final results for unfiltered leakage of the MCR (LAT w/o 3335 dated May 31, 2004) are summarized below for each of the three modes of control room ventilation operation:

• 79 CFM Recirculation mode

• 59 CFM Isolation mode

• 36 CFM Pressurized mode Note per RGI.197, Section 1.4, that in-leakage rates below IOOCFM do not require consideration of measurement uncertainty. As shown in the table below, the results of the various dose analyses in terms of the assumed control room in-leakage readily exceed the values derived from the tracer gas testing.

7

Sequence Type Control Room In-leakage Modeled LBLOCA 200 cfm constant unfiltered SBLOCA 100 cfm unfiltered (0-2 hrs), 65 cfm in pressurized mode (> 2 hrs)

CEA Ejection 150 cfm constant unfiltered SGTR 250 cfm constant unfiltered FHA 200 cfm constant unfiltered IC-MSLB 100 cfm constant unfiltered OC-MSLBIFWLB 100 cfm constant unfiltered A study was performed to verify the assumptions made in a majority of the AST source term calculations pertaining to ability of the control room to isolate on a high radiation signal before any radiation enters the control room envelope. These calculations assumed that the normal mode of control ventilation was operating at its nominal operating conditions. Results from this study determined that the control room boundary will isolate within roughly 2.7 seconds and that it takes approximately 3.1 seconds for the contaminated air to travel from the radiation detectors to the first isolation valve/damper. Therefore, it was confirmed that the control room boundary will be isolated prior to contamination in the form of fission products entering the ductwork and reaching the control room envelope.

1.3.1. Control Room Air Conditioning System The Control Room Air Conditioning System is described in Waterford 3 UFSAR, Sections 6.4 and 9.4.1. No changes to the design or operational modes of the system have been made in conjunction with implementing the AST.

The system maintains a suitable environment for personnel and equipment during normal operation and accident conditions, and protects control room personnel from the effects of airborne radioactive material, toxic gas, and smoke. Only its features and operation with respect to mitigating doses due to airborne radioactive material are discussed here. The brief, high level discussion of the system is intended to provide background information for system features and operator actions that are cited in various dose consequence analyses.

The system provided for Waterford 3 has a single normal air intake and two widely separated air inlets for emergency operation. The system is a zone isolation system with filtered recirculated air and provisions for maintaining positive pressure in the control room envelope under accident conditions. Intake air for pressurization is filtered before entering the control room envelope.

During normal operation, air is drawn into the system through the single normal air intake. Air handling units condition the air for distribution into the control room envelope, and exhaust fans discharge air from the control room envelope to the environment.

The Control Room Air Conditioning System is automatically isolated on a Safety Injection Actuation Signal (SIAS) or by high radiation detected outside the normal air intake. Automatic isolation involves shutting the normal outside air intake and exhaust isolation valves, stopping the normally operating exhaust fans, opening recirculation dampers, and starting both emergency 8

filtration units in the recirculation mode. Operation of only one emergency filtration unit is required. The control room operators may then initiate the pressurization mode of operation by opening a valve in one of the emergency air intakes. The intake air is ducted to the operating filtration unit before being discharged to the control room envelope. The emergency filtration unit continues to recirculate air within the control room envelope while providing filtered pressurization air flow.

1.3.2. Control Room Ventilation Model Various control room and Control Room Air Conditioning System parameters are input to the control room dose analyses. The assumed control room envelope volume is 220,000 ft3.

Recirculation flow through the filter of the operating emergency filtration unit is assumed to be 3,800 Cubic Feet per Minute (CFM). Consistent with Technical Specification limits, filter efficiency of 99% is assumed for aerosol, elemental, and organic iodine. A 0% efficiency is assumed for noble gases. The pressurization flow rate is assumed to be 225 CFM (300 CFM assumed for SGTR).

Individual dose analyses specify the assumed unfiltered in-leakage rates, the time at which operators initiate the pressurization mode of operation (if such action is assumed), and the time at which operators select the more favorable of the two emergency air intakes (if such action is assumed).

Unfiltered in-leakage to the control room envelope distributes through the volume and is processed through the emergency filtration unit with filter efficiencies assumed as above.

Control room out-leakage is assumed to be the sum of the unfiltered in-leakage and the minimum pressurization flow. For purposes of determining out-leakage, the minimum pressurization flow in the pressurization mode of operation is assumed to be 50 CFM. As an example, if unfiltered in-leakage is assumed to be 100 CFM for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> prior to initiating pressurization and 65 CFM after initiating pressurization, then the out-leakage would be 100 CFM for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> (100 CFM in-leakage and 0 CFM pressurization flow) and 115 CFM thereafter (65 CFM in-leakage and 50 CFM pressurization flow).

1.4. Exceptions to Reg. Guide 1.183 For the purpose of the Waterford 3 Licensing Report for the Radiological Consequences of Accidents for the Waterford Steam Electric Station, Unit 3 Using Alternate Source Term Methodology, the guidelines within Reg. Guide 1.183 were followed with the following exceptions:

• The breathing rates assumed for the control room and off-site doses use a more accurate value of 3.47E-4 m3 /s versus the rounded-up 3.5E-4 m3 /s specified in the regulatory guidance.

• Section 5.1 of Appendix E states "For facilities with traditional generator specifications (both per generator and total of all generators), the leakage should be apportioned between the unaffected and affected SGs in such a manner that the calculated dose is maximized." For the accident sequences analyzed as part of this submittal, the primary-to-secondary side SG tube leakage is specified on a per SG basis.

• Appendix F states that for SG tube rupture sequences, the analysis must address an accident-initiated iodine spike equivalent to 335 times the iodine activity. Waterford 3 has 9

conservatively modeled the accident-initiated iodine spike equivalent to 500 times the iodine activity (same as f6r MSLB).

Regulatory Guide 1.183 states (Appendices E & F) that a PF of 100 can be used for sequences that involve releases due to secondary side steaming (CEA Ejection, MSLB, SGTR, and implied for SBLOCA). For these sequences a conservatively smaller value (on the order of 10 (CEA Ejection, SBLOCA) or I (MSLB, SGTR)) was used early in the event sequences to the account for uncertainties in SG performance and for potential uncovery of the top of the SG U-tubes due to the event transient.

10

TABLE 1-1 LOCA CORE INVENTORY Isotope CiMlVt Isotope Ci/MWt Isotope Ci/MWt Co-58 2.553E+02 Ru-103 4.509E+04 Cs-136 2.11OE+03 Co-60 1.953E+02 Ru-105 2.340E+04 Cs-137 4.273E+03 Kr-85 3.562E+02 Ru-106 1.719E+04 Ba-139 4.976E+04 Kr-85m 1.052E+04 Rh-105 1.621E+04 Ba-140 5.088E+04 Kr-87 2.125E+04 Sb-127 2.208E+03 La-140 5.285E+04 Kr-88 3.003E+04 Sb-129 9.305E+03 La-141 4.615E+04 Rb-86 1.496E+01 Te-127 2.132E+03 La-142 4.449E+04 Sr-89 2.941E+04 Te-127m 2.823E+02 Ce-141 4.476E+04 Sr-90 2.867E+03 Te-129 9.162E+03 Ce-143 4.585E+04 Sr-91 3.927E+04 Te-129m 1.360E+03 Ce-144 3.421E+04 Sr-92 3.805E+04 Te-131m 4.161E+03 Pr-143 4.450E+04 Y-90 3.022E+03 Te-132 4.059E+04 Nd-147 1.911E+04 Y-91 3.683E+04 1-131 2.853E+04 Np-239 5.120E+05 Y-92 3.819E+04 1-132 4.127E+04 Pu-238 2.902E+01 Y-93 4.320E+04 1-133 5.769E+04 Pu-239 6.545E+00 Zr-95 4.785E+04 1-134 6.418E+04 Pu-240 8.254E+00 Zr-97 4.562E+04 1-135 5.412E+04 Pu-241 1.390E+03 Nb-95 4.782E+04 Xe-133 5.642E+04 Am-241 9.181 E-01 Mo-99 5.173E+04 Xe-135 1.659E+04 Cm-242 3.514E+02 Tc-99m 4.532E+04 Cs-134 8.023E+03 Cm-244 2.056E+01 11

TABLE 1-l.A CORE INVENTORY FOR STEAMING EVENTS Isotope Curics Kr-83m 1.68E+07 Kr-85 1.17E+06 Kr-85m 3.93E+07 Kr-87 7.94E+07 Kr-88 1.12E+08 1-131 1.04E+08 1-132 1.50E+08 1-133 2.13E+08 1-134 2.40E+08 1-135 1.98E+08 Xe-133 2.05E+08 Xe-135 6.20E+07 Xe-131m 1.17E+06 Xe-133m 6.56E+06 Xe-135m 4.17E+07 Xe-138 1.95E+08 12

TABLE 1-2 GENERAL RADIOLOGICAL ANALYSIS ASSUMPTIONS Core Power Level: 3735 MWt Maximum Radial Peaking Factor:

LOCA 1.0 (100% fuel failure)

FHA, Non-LOCA Transients 1.65 Containment Leak Rate: 0.50 % volume/day (0-24 hours) 0.25 % volume/day (24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> - 30 days)

Containment Volume: 2,568,000 ft3 Sprayed Volume Fraction: 80%

Unsprayed Volume Fraction: 20%

Spray Fission Product Removal (LBLOCA):

Elemental 20/hr (maximum PF = 200)

Organic 0 Particulate/Aerosol 3.596/hr (until PF = 50) 0.3596/hr (once PF > 50)

Natural Deposition:

Elemental 0.40/hr Organic 0 Particulate/Aerosol Powers 10% Aerosol Decontamination Factor Containment Mixing Rate Between Sprayed and 17,122 CFM Unsprayed Regions:

Maximum Spray Delay Time: 60 seconds Containment Leakage Pathway:

Controlled Ventilation Area System (CVAS)

Filtration (Reactor Auxiliary Building) 54%

Shield Building 40%

Unfiltered Direct Bypass 6%

Shield Building Volume: 550,000 ft3 SBVS Flow Rate: 11,000 CFM per train Primary System Activity: 1.0 piCi/g DEI- 131, 100/E Secondary System Activity: 0.1 jiCi/g DEI- 131 13

TABLE 1-2 (C. ont.)

GENERAL RADIOLOGICAL ANI kLYSIS ASSUMPTIONS Filtration Efficiency (Aerosol/Elemental/Organic Iodine)

CVAS 99%

SBVS 99%*

Control Room 99%

• Effective filter efficiency of 90% due to assumed induced flow in idle train Primary Coolant Volume: 7115 ft3 Secondary Coolant Inventory: 153,700 Ibm @ HFP 241,450 Ibm @ HZP Breathing Rate:

0-8 hr 3.47E-04 m3 /sec 8 - 24 hr 1.75E-04 m3 /sec

>24 hr 2.32E-04 m3 /sec Control Room Parameters Volume: 220,000 ft3 Recirculation Flow Rate: 3800 CFM Filter Efficiency: 99% (elemental/organic/particulate)

Pressurization Flow: 225 CFM (maximum) 50 CFM (minimum after 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />) 0 CFM (minimum, 0-2 hours)

Unfiltered In-leakage: See Section 1.3 Breathing Rate: 3.47E-04 m3 /sec Control Room Occupancy Factors:

0-24 hour 1.00 24-96 hours 0.60 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> - 30 days 0.40 14

TABLE 1-3 X/Q ATMOSPHERIC DISPERSION FACTORS (s/mr3)

Duration Distance (m) 5% (s/mr3 )

2 HR (0-2 hr EAB) 914 4.31 E-04 8 HR (0-8 hrs LPZ) 3200 6.58E-05 16 HR (8-24 hrs) 3200 4.45E-05 3 DAYS (24-96 hrs) 3200 1.91 E-05 26 DAYS (96-720 hrs) 3200 5.88E-06 15

TABLE 1-3 (Cont.)

MAIN CONTOL ROOM X/Q ATMOSPHERIC DISPERSION FACTORS (s/m 3 )

5% PROBABILITY - LEVEL X/Q VALUES (s/m 3 )

East MSSV West MSSV East ADV Time Period East CR Air Intake West CR Air Intake East CR Air Intake West CR Air Intake East CR Air Intake West CR Air Intake Oto 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 4.36E-02 1.37E-03 1.52E-03 7.40E-03 1.06E-01 1.23E-03 2 to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 3.08E-02 9.34E-04 9.44E-04 5.44E-03 7.45E-02 8.3 1E-04 8 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 1.33E-02 4.48E-04 4.OOE-04 2.46E-03 3.30E-02 4.00E-04 I to4 days 9.01 E-03 2.99E-04 2.81E-04 1.92E-03 2.31E-02 2.63E-04 4to30 days 6.57E-03 2.10E-04 2.07E-04 1.50E-03 1.62E-02 1.85E-04 West ADV East MSL West MSL Time Period East CR Air Intake West CR Air Intake East CR Air Intake West CR Air Intake East CR Air Intake West CR Air Intake Oto 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 1.36E-03 7.50E-03 5.09E-02 1.44E-03 1.54E-03 9.28E-03 2 to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 8.29E-04 5.62E-03 3.26E-02 9.78E-04 9.62E-04 6.84E-03 8 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 3.55E-04 2.57E-03 1.39E-02 4.68E-04 4.1 1E-04 3.1 1E-03 Ito 4 days 2.48E-04 2.04E-03 8.8 1E-03 3.05E-04 2.89E-04 2.37E-03 4 to 30 days 1.85E-04 1.57E-03 6.87E-03 2.18E-04 2.15E-04 1.85E-03 Plant Stack FHB Truck Bay FHB Personnel Door Time Period East CR Air Intake West CR Air Intake East CR Air Intake West CR Air Intake East CR Air Intake West CR Air Intake Oto 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 2.77E-03 2.06E-03 7.50E-04 7.63E-04 9.75E-04 1.05E-03 2 to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 1.78E-03 1.56E-03 6.15E-04 6.32E-04 7.74E-04 8.72E-04 8 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 7.22E-04 7.16E-04 2.62E-04 2.95E-04 3.33E-04 4.02E-04 I to 4 days 5.27E-04 5.49E-04 1.82E-04 2.27E-04 2.22E-04 3.08E-04 4to 30 days 4.05E-04 4.32E-04 1.25E-04 1.70E-04 1.55E-04 2.29E-04 16

TABLE 1-3 (Cont.)

MAIN CONTOL ROOM X/Q ATMOSPHERIC DISPERSION FACTORS (s/M3 )

5% PROBABILITY - LEVEL X/Q VALUES (s/m 3 )

Containment Hatch Containment Purge Time Period East CR Air Intake West CR Air Intake East CR Air Intake West CR Air Intake 0 to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 1.22E-03 1.93E-03 I.55E-02 1.68E-03 2 to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 8.54E-04 1.60E-03 1.0 IE-02 1.20E-03 8 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 3.64E-04 7.42E-04 4.18E-03 5.75E-04 I to 4 days 2.43E-04 5.61 E-04 2.72-03 3.90E-04 4 to 30 days 1.86E-04 4.24E-04 2.13E-03 2.67E-04 17

TABLE 1-4 RCS INITIAL RADIOISOTOPIC DISTRIBUTION (Based on 1% Fuel Failure)

Radioisotope Maximum RCS Activity (Ci/cm 3)

Ba-140 7.70E-03 Br-84 2.1 OE-02 Ce-144 8.501E-04 Cs-134 5.20E-01 Cs-136 9.80E-02 Cs-137 7.90E-O1 1-131 4.80E+OO 1-132 9.50E-Ol 1-133 5.60E+OO 1-134 4.80E-O1 1-135 2.70E+OO Kr-85 5.50E+OO Kr-85M 1.1IOE+OO Kr-87 8.70E-O1 Kr-88 2.40E+OO La-140 3.30E-03 Mo-99 5.1OE-Ol Pr-143 11.OOE-03 Rb-88 2.40E+OO Ru-103 3.70E-04 Ru-106 1.70E-04 Sr-89 5.90E-03 Sr-90 3.70E-04 Sr-91 6.40E-03 Te-129 1.1 OE-02 Te-132 3.60E-O I Xe-131M 5.40E+OO Xe-133 3.30E+02 Xe-135 7.40E+OO Xe-135M 6.80E-01 Xe-138 5.70E-01 Y-90 1.30E-04 Y-91 8.80E-04 Y-91M 3.80E-03 Zr-95 9.80E-04 Co-60 4.06E-04 Co-58 1.76E-02 Mn-54 1.89E-04 Cr-51 5.37E-03 Fe-59 7.26E-05 H-3 3.5 18

TABLE 1-5 DOSE CONVERSION FACTORS - EFFECTIVE CLOUDSHINE FROM NUREG/CR-6604, TABLE 1.4.3.3-2 Isotope Sv-slBn-m 3 Isotope Sv-s/Bq-m 3 Isotope Sv-s/Bq-rm3 Co-58 4.760E- 14 Ru-103 2.251E-14 Cs-136 1.060E-13 Co-60 1.260E- 13 Ru-105 3.810E-14 Cs-137 2.725E-14 Kr-85 1.190E-16 Ru-106 1.040E-14 Ba-139 2.170E-15 Kr-85m 7.480E-15 Rh-105 3.720E-15 Ba-140 8.580E-15 Kr-87 4.120E-14 Sb-127 3.330E-14 La-140 1.170E-13 Kr-88 1.020E-13 Sb-129 7.140E-14 La-141 2.390E-15 Rb-86 4.810E-15 Te-127 2.420E-16 La-142 1.440E-13 Sr-89 7.730E-17 Te-127m 1.470E-16 Ce-141 3.430E-15 Sr-90 7.530E-18 Te-129 2.750E-15 Ce-143 1.290E-14 Sr-91 4.924E- 14 Te-129m 3.337E-15 Ce-144 2.773E-15 Sr-92 6.790E-14 Te-131m 7.463E-14 Pr-143 2.100E-17 Y-90 1.900E-16 Te-132 1.030E-14 Nd-147 6.190E-15 Y-91 2.600E- 16 1-131 1.820E-14 Np-239 7.690E-15 Y-92 1.300E-14 1-132 1.120E-13 Pu-238 4.880E-18 Y-93 4.800E-15 1-133 2.940E-14 Pu-239 4.240E-18 Zr-95 3.600E-14 1-134 1.300E-13 Pu-240 4.750E-18 Zr-97 4.432E-14 1-135 8.294E- 14 Pu-241 7.250E-20 Nb-95 3.740E-14 Xe-133 1.560E-15 Am-241 8.180E-16 Mo-99 7.280E-1 5 Xe-135 1.190E-14 Cm-242 5.690E- 18 Tc-99m 5.890E- 15 Cs-134 7.570E- 14 Cm-244 4.910E-18 19

TABLE 1-5A DOSE CONVERSION FACTORS - EFFECTIVE INHALED CHRONIC FROM NUREG/CR-6604, TABLE 1.4.3.3-2 Isotope Sv-stBn-m 3 Isotone Sv-s/Bq-m3 Isotope Sv-s/Bq-m3 Co-58 2.940E-09 Ru-103 2.421 E-09 Cs-136 1.980E-09 Co-60 5.91 OE-08 Ru-105 1.230E-10 Cs-137 8.630E-09 Kr-85 O.OOOE-00 Ru-106 1.290E-07 Ba-139 4.640E-11 Kr-85m O.OOOE-00 Rh-105 2.580E-10 Ba-140 1.010E-09 Kr-87 O.OOOE-00 Sb-127 1.630E-09 La-140 1.31 OE-09 Kr-88 O.OOOE-O0 Sb-129 1.740E-10 La-1 41 1.570E-10 Rb-86 1.790E-09 Te-127 8.600E-1 1 La-142 6.840E-1 1 Sr-89 1.120E-08 Te-127m 5.81 OE-09 Ce-141 2.420E-09 Sr-90 3.51 OE-07 Te-129 2.090E-1 1 Ce-143 9.160E-10 Sr-91 4.547E-10 Te-129m 6.484E-09 Ce-144 1.0

.OE-07 Sr-92 2.180E-10 Te-131m 1.758E-09 Pr-143 2.190E-09 Y-90 2.280E-09 Te-132 2.550E-09 Nd-147 1.850E-09 Y-91 1.320E-08 1-131 8.890E-09 Np-239 6.780E- 10 Y-92 2.11OE-10 1-132 1.030E-1 0 Pu-238 7.790E-05 Y-93 5.820E- 10 1-133 1.580E-09 Pu-239 8.330E-05 Zr-95 6.390E-09 1-134 3.550E-1 I Pu-240 8.330E-05 Zr-97 1.171E-09 1-135 3.320E-10 Pu-241 1.340E-06 Nb-95 1.570E-09 Xe-133 O.OOOE-OO Am-241 1.200E-04 Mo-99 1.070E-09 Xe-135 O.OOOE-OO Cm-242 4.670E-06 Tc-99m 8.800E- 12 Cs-134 1.250E-08 Cm-244 6.700E-05 20

TABLE 1-6 ISOTOPIC GAMMA AND BETA DISINTEGRATION ENERGIES Isotope Half-life (s) Er (MeV/dis) Rp (MeV/dis) 1-131 6.9466E+05 0.381 0.194 I-132 8.2800E+04 2.283 0.496 1-133 7.4880E+04 0.608 0.410 1-134 3.1560E+03 2.624 0.623 1-135 2.3796E+04 1.580 0.367 Kr-83m 6.5880E+03 0.002 0.037 Kr-85m 1.6127E+04 0.159 0.258 Kr-85 3.3830E+08 0.002 0.251 Kr-87 4.5780E+03 0.793 1.324 Kr-88 1.0224E+04 1.957 0.364 Xe-131m 1.0282E+06 0.020 0.143 Xe-133m 1.8904E+05 0.042 0.190 Xe-133 4.5317E+05 0.045 0.135 Xe-135m 9.1740E+02 0.432 0.098 Xe-135 3.2724E+04 0.247 0.316 Xe-138 1.0199E+03 1.128 0.672 21

2.0 CONCLUSION

S A summary of the calculated dose consequences of each analyzed scenario is presented in Table 2-1. All accident radiological consequences meet the acceptance criteria for the MCR, EAB, and LPZ.

TABLE 2-1

SUMMARY

OF RESULTS Acceptance Event Scenario Dose Consequences Criterion EAB LPZ MCR EAB&LPZ/IMCR LBLOCA 5.421 2.612 4.405 25/5 SBLOCA 1.9597 1.0831 3.928 25/5 IC-MSLB 0.5998 0.1925 4.888 25/5 SGTR- PIS 0.494 0.143 4.93 25/5 SGTR-GIS 0.0564 0.196 Bounded by 2.5/5 PIS case OC-MSLB/FWLB - PIS 0.2219 0.0774 2.2029 25/5 OC-MSLB/FWLB - GIS 0.2262 0.1240 3.6158 2.5/5 CEA Ejection 1.028 0.653 3.19 6.3/5 FHA 0.55 0.085 0.19 6.3/5 Notes: Detailed discussions for each individual event are presented in Sections 4 through 11.

The detailed analyses for each event demonstrate that radiological consequences meet the TEDE dose acceptance limits for off-site dose. The radiological consequences for MCR dose for all events are < 5 Rem TEDE.

All Results are presented in units of rem TEDE.

For cases that have secondary side steaming release paths, the off-site EAB worst-case 2-hour TEDE dose due to the noble gas and iodine/alkali metals do not necessarily correspond to the same accident time interval. The worst case 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> dose due to noble gas has been added to the worst case 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> dose for iodines even if the times do not coincide.

22

3.0 REFERENCES

1. USAEC TID-14844, "Calculation of Distance Factors for Power and Test Reactor Sites," March 23, 1962.

2. USNRC NUREG-1465, "Accident Source Terms for Light-Water Nuclear Power Plants,"

February 1995.

3. Regulatory Guide 1.183, "Alternative Radiological Source Terms for Evaluating Design Basis Accident at Nuclear Power Reactors," July 2000.

4. W3F1-2003-0074, "License Amendment Request NPF-38-249, Extended Power Uprate, Waterford Steam Electric Station, Unit 3, Docket No. 50-3 82, License No. NPF-3 8," November 13, 2003.

5. USNRC Generic Letter 2003-0 1,"Control Room Habitability," June 12, 2003.

6. ICRP, Publication 30, "Limits for Intakes of Radionuclides by Workers," 1979.

7. NUREG/CR-6604, "RADTRAD, A Simplified Model for Radionuclide Transport and Removal and Dose Estimation," April 1998.

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

9. NUREG-0800, Section 6.5.2, Rev. 2, "Containment Spray as a Fission Product Cleanup System,"

December 1988.

10. NUREG/CR-633 1,Rev. 1,"Atmospheric Relative Concentrations in Building Wakes," May 1977.

11. NUREG/CR-2858, "An Atmospheric Dispersion Program for Evaluating Design Basis Accidental Releases of Radioactive Materials from Nuclear Power Stations," November 1982.

12. 10CFR50.67, "Accident Source Term," December 23, 1999.

13. ANSI/ANS 58.8, "Time Response Design Criteria for Safety Related Operator Actions," 1984.

14. Oak Ridge National Laboratory Code, "ORIGEN 2 Isotopic Generation and Depletion Code -

Matrix Exponential Method," September 1983.

15. NUREG/CR-5009, "Assessment of the Use of Extended Burnup Fuel in Light Water Reactors,"

February 1988.

16. Letter from N. Kalyanam (USNRC) to J. Venable (Entergy), "Waterford Steam Electric Station Unit 3 - Issuance of Amendment Re: Letdown Line Break Dose Consequences Revision (TAC No. MB323 1)," January 8, 2003.

17. W3F 1-2004-0017, "Supplement to Amendment Request NPF-38-249, Extended Power Uprate, Waterford Steam Electric Station, Unit 3, Docket No. 50-3 82, License No. NPF-3 8" March 4, 2004.

23

4.0 POST-LOSS OF COOLANT ACCIDENT PH EVALUATION

SUMMARY

The NUREG-1465 accident isotopic release specification allows deposition of iodine in the containment sump. The iodine is assumed to remain in solution as long as the pool pH is maintained at a pH of 7 or greater. Per RG 1.183 (Appendices A & H), a containment sump water pH history as a function of time following a severe accident, using AST. The containment sump pH calculation was performed using the STARpH 1.04 code. The analysis conservatively maximizes the acid contributions from Hydroiodic Acid from core halogens and Nitric Acid from radiolysis of water and Hydrochloric Acid from radiolysis of chloride bearing cables inside containment. Table 4-1 below contains a summary of the key design inputs used to generate the containment sump water pH history.

The STARpH code considers several key elements in its calculations for containment sump pH.

They are:

1. The Radiolysis of Water Model in the STARpH code calculates the nitric acid [HNO 3 ]

generated in the containment sump water pool generated by radiolysis.

2. The Radiolysis of Cable Model in the STARpH code calculates the concentration of hydrochloric acid [HCI] in the containment sump wvater pool as a result of radiolysis of electrical cable insulation.

3. For the purpose of this analysis, the generation of organic acids from the radiolysis of organic materials dissolved from containment surface coatings was neglected.

4. Cesium hydroxide that is produced as a result of fission products released from cesium that is not in the form of Csl is typically assumed to exit the core in the form of cesium hydroxide.

For the purposes of containment sump pH analysis, this was conservatively ignored as a source in the sump pH calculation.

5. Table 4-1 illustrates the mass of hypalon jackets in containment. The hypalon material used in cable insulation is the only material of interest due to its chloride content, which is released and forms hydrochloric acid as part of the radiolysis process.

The results of the containment sump water pH using the STARpH predict that the containment sump pH will be 7.0 or greater for the 30-day duration of the analysis, thus preventing the re-evolution of elemental iodine dissolved in the containment sump water.

24

TABLE 4-1

SUMMARY

OF KEY DESIGN INPUTS FOR CONTAINMENT SUMP PH CALCULATION Core Power Level: 3716 MWt Mass of TSP in containment: 20,521 Ibm 30-day sump average temp: 77.70 C Mass of hypalon jackets: 17,649.7 Ibm RWSP Boron concentration: 3,000 ppm SIT Boron concentration: 3,000 ppm BAMT Boron concentration: 6,187 ppm RCS Boron concentration: 2,500 ppm RWSP liquid volume: 576,859.1 gallons @ 501F SIT liquid volume: 48,100 gallons @ 90'F BAMT liquid volume: 22,920 gallons @ 490 F RCS liquid mass: 498,000 Ibm Initial Sump pH: 7.1 Containment Free Volume: 2.568x IO6 ft3 Waterford 3 Core Fission Products:

Iodine Group 32.4 Kg Cesium Group 414 Kg Tellurium Group 71.4 Kg Strontium Group 127 Kg Barium Group 186 Kg Ruthenium Group 1023 Kg Cerium Group 1390 Kg Lathenides 1493 Kg 25

5.0 LARGE BREAK LOSS OF COOLANT ACCIDENT (LBLOCA)

The design basis LBLOCA is postulated as a break in the reactor coolant pressure boundary piping. An abrupt failure of the main reactor coolant piping is assumed to occur and it is assumed that the emergency core cooling system fails to prevent the core from experiencing significant degradation. This is considered a Limiting Fault event. Activity from the core is released to containment and subsequently to the environment by means of containment leakage or leakage from the emergency core cooling system. Release of core radioactive inventory to the containment is postulated in accordance with RG 1.183 guidance on activity release and timing for the gap fraction release and early-in vessel release phases.

Other than adoption of the RG 1.183 methodology, the LBLOCA dose analysis is relatively unchanged compared to the analysis presented in EPU Licensing Amendment Request, Reference 4.

5.1. Input Parameters and Assumptions The input parameters and assumptions are listed in Table 5-1. Certain assumptions are discussed in additional detail below.

5.1.1. Source Term Table 1-1 documents the core inventory assumed for the LBLOCA radiological dose calculations.

Twvo separate ORIGEN calculations were conducted for the Waterford 3 EPU project to provide core inventories. One calculation was performed to determine the gap fission product activities in peak power rods. A second calculation was performed to determine the core-wide fission product inventory. There was generally good agreement between these two calculations, with their slightly separate biases. A LOCA source term (Table 1-1) was constructed using the more conservative (larger) value of core inventory from the two sources. Several isotopes are modeled in RADTRAD for which inventories were not calculated in the ORIGEN calculations. For those isotopes, the default Pressurized Water Reactor (PWR) core inventories (on a Ci/MWt basis) from NUREG/CR-6604 (Reference 7) were assumed.

The release fractions applied to the various species of fission products are consistent with Table 2 of RG 1.183 for PWR core inventory fraction releases for the gap release phase and early in-vessel phase of release. Timing of the release phases is from Table 4 of RG 1.183 for LOCA release phase timing. This information is documented in Table 5-2.

The reactor coolant initial activity is insignificant in comparison with the releases due to the postulated core damage for this event.

5.1.2. Iodine Chemical Form As listed in Table 5-2, iodine released to containment is assumed to be 95% aerosol/particulate, 4.85% elemental, and 0.15% organic. This is consistent with Section 3.5 of RG 1.183.

The radioiodine postulated to be available for release to the environment through ESF leakage is assumed to be 97% elemental and 3% organic. This is consistent with Section 5.6 of RG 1.183 Appendix A.

26

5.1.3. Release Pathway Activity from the RCS and the failed core is released into the containment. Releases are postulated from the containment to the environment by three containment air leakage pathways (RAB/CVAS, Shield Building, and Direct Bypass) and by leakage from Engineered Safety Feature (ESF) systems (safety injection and containment spray) which take suction, upon recirculation, from the safety injection sump. The fraction of the release associated with each of the three containment air leakage pathways is specified in Table 1-2.

The containment is modeled as a sprayed and an unsprayed region, where the sprayed region is subject to fission product removal due to the action of the containment sprays (80% of the containment volume is assumed subject to containment spray). Consistent with RG 1.183, Appendix A, a mixing rate due to natural convection between the sprayed and unsprayed regions of containment can be assumed to equal two turnovers of the unsprayed region per hour; this assumption has been adopted for the LBLOCA dose calculation. This is considered a conservative assumption since at least one containment fan cooler is assumed available, providing forced circulation mixing within the containment.

The containment is assumed to leak at the design rate of 0.50 w/o per day for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, and at half that rate (0.25 w/o per day) thereafter. This is consistent with RG 1.183, Appendix A.

Direct bypass releases are assumed to be released unfiltered directly to the environment.

Releases to the area of the Reactor Auxiliary Building (RAB) serviced by the Controlled Ventilation Areas System (CVAS) are assumed to be filtered and directly released to the environment; no credit is taken for holdup in the RAB. Shield Building holdup and dilution is modeled. A SBVS maximum flow rate of 11,000 CFM per train is modeled. It is assumed that when one train is operating, flow is induced in the second train, which is assumed to be unfiltered. The Shield Building pressure transient following a LBLOCA is documented in UFSAR Figures 6.2-47a and 6.2-47b. Conservatively, when the SBVS is in exhaust mode releasing to the environment, a total flow of 24,244 CFM is assumed with a 89.8% filter efficiency; this very conservatively assumes that even though each train is operating, it is also inducing the unfiltered flow. When the SBVS is in recirculation, only a nominal flow rate of 10,000 CFM is assumed and it is assumed that only one train is operating. Thus, the modeling of the SBVS is very conservative. A small effective exhaust flow of approximately 35 CFM is assumed for long-term operation of the SBVS (i.e., beyond about 43 hours4.976852e-4 days <br />0.0119 hours <br />7.109788e-5 weeks <br />1.63615e-5 months <br />); the remaining flow, based on the nominal 10,000 CFM flow rate, is assumed to be in recirculation. After 168 hours0.00194 days <br />0.0467 hours <br />2.777778e-4 weeks <br />6.3924e-5 months <br />, the SBVS is assumed to be exhausting to account for the postulated failure of the containment Maintenance Hatch seal.

The analysis considers a leak rate of 1.0 GPM from ECCS systems that are recirculated and may leak to locations serviced by the CVAS system in the RAB. While no credit is taken for holdup and dilution in the RAB, CVAS filtration is credited. A flashing fraction of 10% is assumed, consistent with RG 1.183. The release is assumed to begin at the postulated earliest time before Emergency Core Cooling System (ECCS) Recirculation of 23.4 minutes.

5.1.4. Removal Coefficients Containment spray removal coefficients consistent with NUREG-0800, Section 6.5.2 (Reference

9) are assumed. One train of Containment Spray is assumed to operate following a LOCA, with a minimum flow rate of 1750 GPM. These values are documented in Table 5-1. Removal of elemental iodine from the containment atmosphere is assumed to be terminated when the airborne 27

inventory drops to 0.5% of the total elemental iodine released to the containment (a PF of 200).

With RG 1.183 source term methodology, this is interpreted as being 0.5% of the total inventory of elemental iodine that is released to the containment atmosphere over the duration of the gap and in-vessel release phases. This occurs after 1.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. The removal coefficient for particulate/aerosol iodine is assume, consistent with NUREG-0800, to decrease by a factor of ten when the airborne inventory has dropped to 2% of the total particulate iodine released to the containment (a PF of 50). This also occurs after 1.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

Per RG 1.183 Appendix A Section 3.2, reduction of airborne activity by natural deposition within the containment may be credited for LOCA. The Powers 10% Aerosol Deposition is specified for natural deposition of aerosols/particulates. This model is described in NUREG/CR-6604. The lower bound of this deposition model (I 0th percentile) is specified. Use of this model is consistent with RG 1.183, Appendix A, Section 3. The guidance of NUREG-0800, Section 6.5.2 is applied for natural deposition of elemental iodine. Natural deposition removal coefficients are documented in Table 5-1.

5.1.5. Main Control Room Model The MCR ventilation model is described in Section 1.3. The LBLOCA dose model for secondary steaming assumes an unfiltered in-leakage of 200 CFM for the event duration. It is assumed that the preferred control room intake is selected at two hours into the event, at which time the operators also initiate the pressurized mode of control room operation. However, no credit is taken in this event scenario for the lower in-leakage during the pressurized mode of operation.

28

5.2. Results The radiological consequence results in Rem TEDE are listed below and compared with the acceptance criteria for LOCA provided by RG 1.183 and I OCFR50.67:

Acceptance LBLOCA Criteria EAB (worst t-wo hour dose) 5.421 25 Rem TEDE LPZ (worst 30 day duration) 2.612 25 Rem TEDE MCR 4.405 5 Rem TEDE Thus, the radiological consequences for LBLOCA are < 25 Rem TEDE for the EAB and LPZ doses and < 5 Rem TEDE for the MCR, based on a maximum control room unfiltered in-leakage of 200 CFM.

29

TABLE 5-1 ASSUMPTIONS USED FOR LBLOCA RADIOLOGICAL ANALYSIS Core Power Level: 3735 MWt Containment Leak Rate: 0.50 % volume/day (0-24 hours) 0.25 % volume/day (24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> - 30 days)

Natural Deposition:

Elemental 0.40/hr Organic 0 Particulate Powers 10% Aerosol Decontamination Factor Spray Fission Product Removal (LBLOCA):

Elemental 20/hr (maximum PF = 200)

Organic 0 Particulate 3.596/hr (until PF = 50) 0.3596/hr (once PF > 50)

Containment Mixing Rate Between Sprayed and 17,122 CFM Unsprayed Regions:

Maximum Spray Delay Time: 60 seconds Containment Leakage Pathway:

Controlled Ventilation Area System (CVAS)

Filtration (Reactor Auxiliary Building) 54%

Shield Building 40%

Unfiltered Direct Bypass 6%

Control Room Parameters See Table 1-2 Main Control Room X/Q Assumed:

Time Unfiltered In-leakage Pressurization Flow 0-2 hr 2.77E-03 2.77E-03 2-8 hr 1.78E-03 3.90E-04*

8-24 hr 7.22E-04 1.79E-04*

1-4 days 5.27E-04 1.37E-04*

4-30 days 4.05E-04 1.08E-04*

• factor of 4 reduction credited per SRP 6.4.

30

TABLE 5-2 SOURCE TERM ASSUMPTIONS: LBLOCA RADIOLOGICAL ANALYSIS Core Inventory Fraction Released into Containment:

Group Gap Release Phase Early In-Vessel Phase Noble Gas

• 0.05 0.95 Halogens 0.05 0.35 Alkali Metals 0.05 0.25 Tellurium Metals 0.00 0.05 Ba, Sr

• 0.00 0.02 Noble Metal 0.00 0.0025 Cerium group 0.00 0.0005 Lanthanides 0.00 0.0002 LOCA Release Phases:

Phase Onset Duration Gap Release 30 sec 0.5 hr Early In-Vessel 0.5 hr 1.3 hr Iodine Chemical Form (release to containment):

Aerosol/Particulate 95%

Elemental 4.85%

Organic 0.15%

Iodine Chemical Form (ESF system leakage):

Elemental 97%

Organic 3%

31

6.0 SMALL BREAK LOSS OF COOLANT ACCIDENT (SBLOCA)

This event is not part of the existing licensing basis documented in the Waterford 3 UFSAR. For small breaks, primary pressure control and decay heat removal are accomplished through steaming from the secondary system. The release dynamics and locations for a SBLOCA can differ from those of the traditional LBLOCA.

The SBLOCA has been analyzed for two different release pathways:

1) Reactor containment building release pathway, similar to that for LBLOCA.

2) Secondary steaming pathway, consisting of releases from the MSSVs or ADVs to the environment.

This is similar to the two different release pathways which are postulated for a PWR CEA Ejection event. For the reactor containment building release pathway, activity released to containment is assumed to be released to the environment due to containment leaking at its design rate. For the secondary steaming pathway, secondary steaming to remove decay heat and to cooldown the plant to shutdown cooling entry conditions is assumed; primary-to-secondary leakage provides a release path for activity to the secondary system, from which it is released to the environment via secondary steaming.

SBLOCA ECCS performance is discussed in Section 2.12 of the EPU Licensing Amendment Request (Reference 4). Although that analysis is currently undergoing revision, the general behavior which supports the logic of the SBLOCA dose analyses herein remains valid. Dynamics for a SBLOCA are very different than for a LBLOCA. As shown in the EPU Licensing Amendment Request, the top of the core remains covered for at least ten minutes for the break sizes considered. For Waterford 3, the smallest break size for which containment spray would not actuate is small enough that the core remains covered during the transient, thus there would be no core damage. Larger break sizes would result in lower RCS pressures, resulting in discharge of the safety injection tanks. The limiting break size will be one where the hot rod cladding heat-up transient is terminated by only the high pressure safety injection pumps. Because the heat-up transient only starts after core uncovery, at least ten minutes into the event, there is no challenge to fuel melt limits. Thus, the only mechanism for fuel damage is clad damage that results in release of the gap activity. A 5% gas gap activity for iodines, noble gas, and alkali metals will be assumed, consistent with RG 1.183. This is consistent with the assumptions used for other SBLOCA dose analyses, such as that for D.C. Cook. No credit will be taken for the effects of containment spray for fission product removal in the SBLOCA dose analyses.

6.1. Input Parameters and Assumptions The input parameters and assumptions are listed in Table 6-1. Certain assumptions are discussed in additional detail below.

6.1.1. Source Term The use of a NUREG-1 465 AST modeling results in a gap fraction of 5.0% being assumed, consistent with the guidance for LOCA of RG 1.183, Table 2. This is an appropriate assumption when 100% of the fuel rods are assumed to fail in a mode that releases the gas gap activity. The gas gap fraction of 5.0% is assumed for iodines, noble gases, and alkali metals (cesium and rubidium). A near instantaneous release duration of 30 seconds or less is assumed.

32

It is assumed that all of the activity release is directed to whichever of the two pathways is under consideration. For example, for the secondary steaming pathway involving primary-to-secondary leakage, the effects to reduce primary activity inventory of any activity release due to the break to containment are neglected.

The curie releases for the secondary steaming pathway are based on the source term of Table I-1.A. The core inventory assumed for the reactor building release pathway are based on the LOCA core inventory of Table 1-1.

6.1.2. Iodine Chemical Form For the secondary streaming pathway, the guidance of Appendix H of RG 1.183 for CEA Ejection is considered the most pertinent guidance. Thus, iodine releases from the SG to the environment are assumed to be 97% elemental and 3% organic. Note since the control room filtration system has the same filter efficiency for the various iodine forms, there is no impact associated with this assumption.

For the reactor containment building release pathway, the chemical form of the radioiodine release is assumed to be 95% particulate, 4.85% elemental,'and 0. 15% organic, consistent with RG 1.183.

6.1.3. Release Pathways Conservatively, all the iodine, alkali metal, and noble gas activity due to the postulated SBLOCA is assumed to be in the primary coolant when determining dose consequences due to primary-to-secondary SG tube leakage and subsequent secondary steaming. Releases are assumed to be terminated once shutdown cooling is initiated and the SGs are no longer providing decay heat removal capability, thus, no further releases would occur for the cooldown to cold shutdown conditions. This is consistent with Table 6 of RG 1.183 for the comparable CEA Ejection event.

A primary-to-secondary SG tube leak rate of 75 gallons per day (gpd) per SG is assumed for the analysis. This is the value which Waterford 3 is proposing to adopt as the Technical Specification limit on primary-to-secondary leak rate. The assumption of a constant 75 gpd per SG value is conservative. A liquid density corresponding to shutdown cooling entry conditions of 350 IF is assumed. This limit will be imposed for normal operating conditions, for which there would be a considerably larger difference between RCS pressure (nominally 2250 psia) and hot full power condition SG pressure (nominally around 800 psia) than is present under SBLOCA conditions.

Under SBLOCA conditions, where secondary steaming is providing decay heat removal and primary pressure control, the pressure differential would be considerably less and secondary pressure will even exceed primary pressure for certain conditions. It is noted that other plants have postulated a duration for SBLOCA of, for example, 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> until it is assumed that primary pressure drops below secondary pressure.

For purposes of evaluating the secondary steaming release for MCR dose, the worst case single failure would be the failure of a DC power bus, which would result in failure of one emergency diesel generator and failure of the control logic for one ADV. It is assumed that the ADV which responds to provide decay heat removal for the event is the ADV with the worst case X/Q, i.e., the East ADV which is assumed to contribute to unfiltered in-leakage at the East MCR Outside Air Intake. No local manual action to open the other ADV and/or to close this ADV to lower releases to the main control room is assumed.

33

The reactor building leakage pathway is similar to the LBLOCA release model. For the reactor containment building leakage pathway, all of the iodine, alkali metal, and noble gas activity is assumed to be released to containment. The design basis containment leak rate of 0.5 % by volume per day for 0-24 hours and 0.25% by volume per day for 1-30 days is assumed. This is the same assumption used for LBLOCA. Since containment pressurization for a SBLOCA would be expected to be less than for a LBLOCA, this is a more conservative assumption in this case.

6.1.4. Removal Coefficients For the secondary steaming path, iodine and alkali metal releases to the secondary side via SG tube leakage are assumed subject to a PF. Consistent with RG 1.183, Appendix E, Section 5.5, a PF of 100 is assumed for iodine and alkali metals. Conservatively, a PF of 10 is assumed for the first 30 minutes of the event to account for potential elevated releases due to the initial transient.

All noble gas release to the secondary side via SG tube leakage is assumed to be immediately released to the environment.

For the reactor containment building leakage pathway, the Powers 10% Aerosol Decontamination Factor model is assumed for natural deposition. This model is contained in the RADTRAD analysis code of NUREG/CR-6604. A natural deposition factor of 0.40/hr is assumed for elemental iodine.

No credit is taken for removal of fission products due to containment spray actuation.

6.1.5. Main Control Room Model The MCR ventilation model is described in Section 1.3. The SBLOCA dose model for secondary steaming assumes an unfiltered in-leakage of 100 CFM for 0-2 hours and credits the reduced unfiltered in-leakage of 65 CFM corresponding to the pressurized mode of operation of control room ventilation thereafter. This corresponds to the assumption that the control room is placed in the pressurized mode of operation at no longer than two hours into the event. It is also assumed that the preferred control room intake is selected at two hours into the event.

The reactor containment building leakage pathway model includes a more conservative assumption of a constant 100 CFM unfiltered in-leakage throughout the event. As for LBLOCA, the plant stack is the release location assumed for determining X/Q for control room intakes.

6.2. Results The radiological consequence results in Rem TEDE are listed below and compared with the acceptance limits for LOCA provided by RG 1.183 and 10CFR5O.67:

Secondary Steaming Reactor Building Acceptance Release Pathway Leakage Pathway Criteria EAB (worst two hour dose) 0.885 1.9597 25 Rem TEDE LPZ (duration) 0.329 1.0831 25 Rem TEDE MCR 3.928 0.7823 5 Rem TEDE 34

The worst case off-site dose occurs for the case of no ADV failure, due to the noble gas contribution being released from both SGs. The MCR dose for the secondary steaming release pathway assumes the failure of the ADV which would have the preferred (lower) value for the xlQ atmospheric dispersion factor.

Thus, the radiological consequences for SBLOCA are < 25 Rem TEDE for the EAB and LPZ doses and < 5 Rem TEDE for the MCR, based on a 75 gpd primary-to-secondary leak rate per SG and maximum control room unfiltered in-leakage of 100 CFM in recirculation mode and 65 CFM in pressurized mode.

35

TABLE 6-1 ASSUMPTIONS USED FOR SBLOCA RADIOLOGICAL ANALYSIS Core Power Level: 3735 MWt Core Inventory: See Table 1-1 Fission Product Gap Fractions:

Iodines 5%

Noble Gases 5%

Alkali metals (Cs & Rb-86) 5%

Fraction of Fuel Rods in Core Failing: 100%

Reactor Building Release Pathway Containment Leak Rate: 0.50 % volume/day (0-24 hours) 0.25 % volume/day (24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> - 30 days)

Natural Deposition:

Elemental 0.40/hr Organic 0 Particulate Powers 10% Aerosol Decontamination Factor Spray Fission Product Removal: Not Credited Iodine Chemical Form (Reactor Building Release Path):

Elemental 4.85%

Organic 0.15%

Particulate 95%

Secondary Steaming Pathway Primary-to-Secondary Leak Rate: 75 gpd per SG Iodine Chemical Form (Reactor Building Release Path):

Elemental 97%

Organic 3%

Particulate 0%

Steaming PF (Iodine and Alkali Metals):

0-30 minutes 10

> 30 minutes 100 Duration of Release: 7.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> Control Room Parameters See Table 1-2 36

TABLE 6-1 (Cont.)

ASSUMPTIONS USED FOR SBLOCA RADIOLOGICAL ANALYSIS Main Control Room x!Q Assumed:

Reactor Building Reactor Building Secondary Steaming Secondary Steaming Tm Unfiltered In-leakage Pressurization Flow Unfiltered In-leakage Pressurization Flow 0-2 hr 2.77E-03

• NA 1.06E-01 NA .

2-8 hr ] .78E-03 3.90E-04

• NA 2.08E-04 *

• factor of 4 reduction credited per SRP 6.4.

Steaming (Ibm) and Activity (DEI-13 1, Ci) Releases 0-2 hr Steaming 2-8 hr Steaming 627,512 858,838 0-15 min 15-30 min 1/2-1 hr 1-2 hr 2-4 hr 4-6 hr 6-7.5 hr 3.28 3.77 1.85 6.51 19.32 21.96 21.33 Alkali Metal Source Term Data, Ci Released:

Cs-134 16.016 Cs-136 4.211 Cs-137 8.529 Rb-86 0.029 37

7.0 INSIDE CONTAINMENT MAIN STEAM LINE BREAK (IC-MSLB)

For this Limiting Fault event, the affected SG will rapidly depressurize due to the assumed complete break of a Main Steam Line inside containment. The affected SG will release activity to the containment due to the MSLB and continue to release activity directly to containment due to the primary-to-secondary leakage of the affected SG Activity is transported to the intact SG via primary-to-secondary leakage, and is subsequently available for release via steaming as the intact SG is used to cool down the RCS to shutdown cooling entry conditions. The atmospheric dump valves are assumed for steaming.

As documented in USAR Chapter 15 and the EPU Licensing Amendment Request (Reference 4),

two cases of MSLB are analyzed:

1. Post-Trip Return-to-Power (RTP) - Reference 4, Section 2.13.1.3.1

2. Pre-Trip Power Excursion - Reference 4, Section 2.13.1.3.3 As described in the EPU Licensing Amendment Request, fuel failure limits for the IC-MSLB are:

1. 2% fuel failure via Departure from Nucleate Boiling Ratio (DNBR) for inside containment RTP MSLB with a LOOP

2. 10% fuel failure (via DNBR) for inside containment MSLB with a LOOP due to combined RTP and Pre-Trip power excursion results

3. 2% fuel failure (centerline melting violation) for inside containment RTP MSLB without LOOP The OC-MSLB does not result in fuel failure. The analysis for the OC-MSLB is presented in conjunction with that for the FWLB in Section 9.0 of this report.

The 2% fuel melt case (Case 3) results in a higher release to the RCS and a slightly higher secondary steaming release (due to the assumption that reactor coolant pumps remain running for 30 minutes) than the corresponding 10% DNBR fuel failure case associated with a LOOP (Case 2). Therefore, only the limiting case of 2% fuel melt with off-site power available (Case 3) is analyzed.

The IC-MSLB accident has been analyzed for two release pathways:

1. Reactor containment building release pathway, similar to that for LBLOCA

2. Secondary steaming pathway, consisting of releases from the MSSVs or ADVs to the environment.

Releases for these two pathways are added to determine the dose consequences for this event.

The dominant release path is the secondary steaming pathway. The secondary steaming pathway assumes that releases are occurring due to primary-to-secondary leakage from the RCS to the SGs. Activity is then released to the environment through the use of the ADVs to remove decay heat and to cool the plant to shutdown cooling entry conditions. Since the control room X/Q values are worse (higher) for ADV releases than for MSSV releases, any releases which would 38

occur through the MSSVs are instead assumed from the ADVs. Release to the environment is terminated for this pathway when shutdown cooling is initiated.

The containment building release pathway model is similar to the LBLOCA release model. Also, per RG 1.183, it is conservatively assumed that all available fission product activity is released to the containment over the first second of the event. Activity released to the containment atmosphere is assumed to be released to the environment due the containment leaking at its design rate.

Fission product removal via the operation of containment sprays is conservatively not credited for this calculation.

7.1. Input Parameters and Assumptions A summary of input parameters and assumptions is provided in Table 7-1. Certain assumptions are discussed in additional detail below.

7.1.1. Source Term While RG 1.183, Appendix E (MSLB) does not explicitly address the source term associated with fuel melt (linear heat generation rate SAFDL violation), this is discussed in Appendix H (PWR Rod Ejection) of RG 1.183. Since that is also a non-LOCA transient like the MSLB, this guidance is considered applicable to linear heat generation rate SAFDL violations for MSLB.

Appendix H specifies that 100% of the noble gases and 25% of the iodines contained in the fuel fraction that experiences melting are available for release from containment. Thus, for a 2% fuel melt, the release fractions would be:

lodines: 2%

• 25% = 0.005 Noble Gas: 100% * .02 = .02 Alkali metals: 100% * .02 = .02 Pre-accident RCS activity is assumed to be at the Technical Specification limit of 1.0 JLCi/gm for Dose Equivalent Iodine- 131 (DEI- 131) and 100/E for noble gases. Secondary activity is assumed to be at the Technical Specification limit of 0.1 tCi/gm for DEI-131.

7.1.2. Iodine Chemical Form For secondary steaming, the iodine is assumed to be 97% elemental and 3% organic in accordance with in RG 1.183 provisions for a CEA Ejection event, a non-LOCA event that also causes fuel melting. Also in accordance with RG 1.183, the iodine is assumed to be 95% aerosol, 4.85% elemental, and 0.15% organic for releases from the containment.

7.1.3. Release Pathways The duration of the release from the SGs for the secondary steaming pathway is assumed to be 7.5 hrs, the assumed time until initiation of shutdown cooling.

The primary-to-secondary leak rate for the intact SG is assumed to be 150 gpd. The primary-to-secondary leak rate for the faulted SG is assumed to be 540 gpd.

The release duration for the reactor containment building release pathway is assumed to be 30 days. The containment is assumed to leak at the design rate of 0.50 w/o per day for the first 24 39

hours, and at half that rate (0.25 w/o per day) thereafter. The fraction of the release associated with each of the three containment air leakage pathways (RAB/CVAS, Shield Building, and Direct Bypass) is specified in Table 1-2. Credit is taken for CVAS and SBVS filtration systems, but no credit is taken for holdup or dilution in the RAS or the Shield Building.

7.1.4. Removal Coefficients For the intact SQ an iodine PF of 1.0 for 0-30 minutes and of 100 beyond 30 minutes is assumed for activity transported to the secondary side. The activity mixes with the SG inventory and is released with the specified PF. RG 1.183, Appendix E (for PWR MSLBs), Position 5.5.4, calls for assuming an iodine PF of 100 when the tubes are not in dry-out (i.e., when top of U-tubes are not uncovered). Assuming a PF of 1.0 for the first 30 minutes accounts for any potential transient which might uncover the SG U-tubes during recovery from the plant scram. Noble gases released to the secondary side of the unaffected SG are assumed to be immediately released to the atmosphere.

For releases into containment from the affected SQ no iodine scrubbing is assumed. Removal of iodine by containment sprays and natural deposition are conservatively ignored.

7.1.5. Main Control Room Model The MCR ventilation model is described in Section 1.3. The IC-MSLB dose model for both the secondary steaming and reactor building release pathways assumes an unfiltered in-leakage of 100 CFM for the event duration. It is assumed that the preferred control room intake is selected at two hours into the event, at which time the operators also initiate the pressurized mode of control room operation. However, no credit is taken for the reduced in-leakage due to the pressurized mode of operation. The filtered intake in the pressurized mode is assumed to be 225 CFM.

The control room dispersion factors for containment leakage were also determined accounting for operator action to initiate the pressurization mode of operation and select the preferred intake at 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. Releases are assumed to occur from the plant stack.

7.2. Results The radiological consequence results in Rem TEDE are presented below and compared to the acceptance limits for the IC-MSLB event provided in RG 1.183, Table 6, and I OCFR50.67 (Reference 12).

The results for the two different pathways considered are:

EAB (worst 2 hr) LPZ (duration) MCR Secondary Steaming 0.5971 0.1778 4.625 Reactor Building 0.0027 0.0147 0.263 Total 0.5998 0.1925 4.888 40

Acceptance Criteria Dose (Rem TEDE)

(Rem TEDE)

EAB (worst two hour dose) 0.5998 25 LPZ (duration) 0.1925 25 MCR 4.888 5 Thus, the radiological consequences for IC-MSLB are < 25 Rem TEDE for the EAB and LPZ doses and < 5 Rem TEDE for the MCR, based on a 150 gpd primary-to-secondary leak rate on the unaffected SQ a 540 gpd primary-to-secondary leak rate on the affected SQ and a maximum control room unfiltered in-leakage of 100 CFM.

41

TABLE 7-1 ASSUMPTIONS USED FOR IC-MSLB RADIOLOGICAL ANALYSIS Core Power Level: 3735 MWt Core Inventory: See Table 1-1 Fission Product Gap Fractions:

Iodines 10%

Noble Gases 10%

Alkali metals (Cs & Rb-86) 12%

Fuel Damage: 2% fuel melt Containment Leakage Pathway Containment Leak Rate: 0.50 % volume/day (0-24 hours) 0.25 % volume/day (24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> - 30 days)

Natural Deposition: Not Credited Spray Fission Product Removal: Not Credited Iodine Chemical Form (Reactor Building Release Path):

Elemental 4.85%

Organic 0.15%

Particulate 95%

Partition Factor (affected SG) 1.0 Secondary Steaming Pathway Primary-to-Secondary Leak Rate: 150 gpd unaffected SQ, 540 gpd faulted SG Iodine Chemical Form (Secondary Steaming Pathway):

Elemental 97%

Organic 3%

Steaming PF (Iodine and Alkali Metals, Intact SG):

0-30 minutes I

> 30 minutes 100 Duration of Release: 7.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> Control Room Parameters See Table 1-2 42

TABLE 7-1 (Cont.)

ASSUMPTIONS USED FOR IC-MSLB RADIOLOGICAL ANALYSIS Main Control Room x/Q Assumed:

Time Reactor Building Reactor Building Secondary Steaming Secondary Steaming Unfiltered In-leakage Pressurization Flow Unfiltered In-leakage Pressurization Flow 0-2 hr 2.77E-03 2.77E-03 1.06E-01 1.06E-01 2-8 hr 1.78E-03 3.90E-04

• 7.45E-02 2.08E-04

• 8-24 hrs 7.22E-04 1.79E-04

• NA NA 1-4 days 5.27E-04 1.37E-04

• NA NA 4-30 days 4.05E-04 1.08E-04

• NA NA

• factor of 4 reduction credited per SRP 6.4.

Steaming (Ibm) and Activity (DEI-13 1, Ci) Releases 0-2 hr Steaming 2-8 hr Steaming 627,512 858,838 150 gpd Leakage 0-15 min 15-30 min 1/2-1 hr 1-2 hr 2-4 hr 4-6 hr 6-7.5 hr 18.95 18.09 0.95 3.67 11.25 13.01 12.70 540 gpd Leakage 0-15 min 15-30 min A2-I hr 1-2 hr 2-4 hr 4-6 hr 6-7.5 hr 53.35 60.66 3.30 13.04 40.26 46.69 45.62 Alkali Metal Source Term Data, Ci Released (150 gpd):

Cs-134 19.273 Cs-136 5.067 Cs-137 10.263 Rb-86 0.036 Alkali Metal Source Term Data, Ci Released (540 gpd):

Cs-134 19.408 Cs-136 5.103 Cs-137 10.335 Rb-86 0.036 43

8.0 STEAM GENERATOR TUBE RUPTURE (SGTR)

For the SGTR event, a complete circumferential rupture of a single SG tube is assumed to occur.

Due to the higher primary coolant pressure, radioactive reactor coolant is discharged to the secondary system. A portion of this activity is released via the main condenser and, after reactor trip and a postulated LOOP, via secondary steaming from the ADVs and MSSVs. No fuel failure is postulated to occur for this event. The initial iodine activity in the SG secondary inventory is also available for release as a result of secondary steaming during the event.

A LOOP is assumed for analyzing the radiological consequences of a SGTR. Thus, steaming releases to the environment occur through ADVs rather than being able to credit the non-safety steam bypass control system for directing steam releases to the main condenser.

As discussed in UFSAR Section 15.6.3.2.3, since greater radiological releases result if early operator actions are assumed, the first operator action is postulated to occur at 7 minutes after reactor trip. Subsequently, a time delay of two minutes between discrete operator actions is assumed. This timing is consistent with ANSI/ANS 58.8 (Reference 13). The operator actions assumed are consistent with Waterford 3 Emergency Operating Procedures (EOPs). The EOPs and the ANSI/ANS standard are used to develop a conservative sequence and timing for assumed operator actions.

These operator actions have been modeled in the SGTR analysis presented in Section 2.13.6.3 of the EPU Licensing Amendment Request (Reference 4).

Table 2.13.6.3.2-1 of the EPU Licensing Amendment Request provides assumptions used in the 3716 MWt EPU analysis of SGTR. Table 2.13.6.3.2-2 provides the sequence of events, including:

Time (sec.) Event 0.0 SG tube rupture occurs 442.7 Core protection calculator hot leg saturation trip condition reached 445.4 Trip breakers open 449 Loss of off-site power 450 SG ADVs open 485 Safety injection actuation signal

>875 Operator takes manual control of SG ADVs. Plant cooldown initiated by steaming using ADV of unaffected SG 1980 Operator isolates affected SG 23630 Operator opens ADV to the affected SG as needed to maintain level below 94% wide range.

28800 Shutdown cooling entry conditions achieved; release stopped.

As described in EPU Licensing Amendment Request, Section 2.13.6.3, the SGTR analysis assumes that a LOOP occurs 3 seconds after reactor trip. This is consistent with the assumption made on other plants and included in CESSAR FSAR Chapter 15. On the basis that no fuel 44

failure occurs for the event, the radiological analyses would be insensitive to this assumption for the Waterford 3 design.

As described in EPU Licensing Amendment Request, Section 2.13.6.3.2, during the SGTR, a total of 65,200 Ibm of primary coolant passes through the rupture into the affected SG Prior to reactor trip, both SGs are steaming normally to the condenser. Due to the high PFs and to the release geometry, releases from this source do not contribute to MCR dose. Following reactor trip, both SGs are steamed through the ADVs. The operator takes manual control of the ADVs at approximately 875 seconds and uses the ADV on the intact SG to cooldown the plant; at 1980 seconds the affected SG is isolated. 46,510 Ibm of steam are released from the affected SG between trip and 875 seconds. No release from the affected SG is seen between 875 and 1980 seconds. Although the EPU Licensing Amendment Request documents operator action to open the ADV to the affected SG to maintain level below 94%, other operator actions are available in the EOPs to maintain SG level and prevent SG overfill, thus this late release from the affected SG is not considered part of the licensing basis releases for the dose analyses (for conservatism, this term has been included in the input parameters for the limiting PIS case for evaluation of radiological consequences). This is also consistent with the guidance of RG 1.183, which calls for considering the releases from the affected SG until it is isolated, whereas releases for the unaffected SG should consider the time until cold shutdown.

The majority of the cooldown of the plant is performed by steaming the unaffected SG A total of 998,000 Ibm of steam are released through the unaffected generator's ADV during the plant cooldown. Radioactivity release through this intact SG is assumed due to primary-to-secondary SG tube leakage of 0.375 gpm per generator (540 gpd). Once shutdown cooling is initiated, there are no further releases postulated from the unaffected SG.

8.1. Input Parameters and Assumptions The input parameters and assumptions are listed in Table 8-1. Certain assumptions are discussed in additional detail below.

8.1.1. Source Term No fuel damage is postulated for the Waterford 3 SGTR event, as documented in Section 2.13.6.3.2 of the EPU Licensing Amendment Request. Two different cases of iodine spiking are considered:

1. A PIS case where a reactor transient is postulated to have occurred prior to the SGTR and has raised primary coolant iodine concentration to the maximum value (60 JtCi/gm DEI-131) allowed perTechnical Specifications.

2. An accident GIS case, where the primary system transient associated with the SGTR causes an iodine spike in the primary system. A spiking factor of 500 is assumed, which is conservative compared to the value of 335 specified per Appendix F to RG 1.183.

Initial reactor coolant activity distribution for SGTR is given by Table 1-4. This distribution is based on predicted activity distributions for power uprate conditions and is applicable for events, such as SGTR, for which no fuel failure will occur. This activity corresponds to the Technical Specification activity limit of I 00/E jCi/gm.

A maximum Technical Specification secondary activity of 0.1 tCi/gm DEI-131 is assumed.

45

Since no fuel failure is postulated, the small contribution to dose consequences from alkali metals have been ignored for this non-limiting event.

8.1.2. Iodine Chemical Form Per RG 1.183, iodine releases from the SGs to the environment are assumed to be 97% elemental and 3% organic. Since the same filter efficiency is specified for all iodine forms, there is no impact on the results of this assumption.

8.1.3. Release Pathway Consistent with RG 1.183, Table 6, the SGTR event is analyzed for a release duration until the affected SG is isolated (1980 seconds) and until cold shutdown is established for the unaffected SG. Once the affected SG is isolated, releases are via steaming from the unaffected SG Releases are terminated once shutdown cooling is initiated.

Releases from the transient analyses documented in Section 2.13.6.3.2 of the Waterford 3 EPU Licensing Amendment Request are used in the radiological analyses. These analyses were conducted using the CENTS transient analysis code.

8.1.4. Removal Coefficients Iodine releases to the affected SG are assumed to flash to vapor whenever the top of the SG U-tubes are uncovered and are available for release without mitigation. The flashing fraction is based on the difference between the primary side fluid enthalpy and the saturation enthalpy on the secondary side.

An iodine PF of 100 is assumed for activity transported to the secondary side prior to reactor trip.

The un-flashed portion of the tube rupture flow mixes with the SG inventory and is released with a PF of 100. RG 1.183, Appendix F,for PWR SGTR, endorses the Appendix E, Position 5.5.4 which calls for assuming an iodine PF of 100.

A PF of 100 is assumed for the 0.375 gpm primary-to-secondary leak rate assumed for the unaffected SQ Prior to reactor trip, any releases from the condenser could be assumed to have an iodine PF of 100 applied to those releases. Credit for additional iodine removal in the condenser is not assumed. The pre-trip releases from the condenser are assumed to contribute only to off-site dose as discussed in Section 8.1.5 below.

All noble gas release to the secondary side is assumed to be immediately released to the environment.

8.1.5. Main Control Room Model The MCR ventilation model is described in Section 1.3. The SGTR dose model for secondary steaming assumes an unfiltered in-leakage of 250 CFM, which is conservatively large compared to the minimum assumed in-leakage of 100 CFM being used for most events. It is assumed that the preferred control room intake is selected at two hours into the event.

46

Note that the limiting (PIS) case was adjusted to protect assumptions of a maximum control room filtered intake flow of 300 CFM and a minimum control room filtered intake flow of 0 CFM (0-2 hours) or 50 CFM (beyond 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />). These are mutually exclusive but bounding assumptions.

Due to geometry considerations, the pre-trip releases from the condenser are assumed to contribute only to the off-site dose consequences. Because the condenser release point and the worst case ADV release locations are in the opposite directions from the worst case control room air intake, and since the MCR envelope is isolated on any high radiation signal prior to the radiation entering the envelope, releases from the condenser are not assumed to contribute to the control room dose. Were wind speed and direction conditions such that releases from the condenser were to be directed to the MCR air intakes, the atmospheric dispersion factors for the ADVs would be greatly reduced. Also, the control room would be isolated on a high radiation signal prior to any of the release entering the control room envelope. Thus, any scenario involving releases from the condenser to the MCR would be less limiting than scenarios involving worst case atmospheric dispersion factors for releases from the ADVs to the control room.

Because the control room envelope would be isolated on a high radiation signal prior to any of the release reaching the envelope, the main control room ventilation system is assumed to be in either of its radiological emergency modes (pressurized or recirculation) once the ADVs open for the subsequent duration of the event.

Because the same 5 Rem TEDE acceptance criterion applies for both the GIS and PIS cases, only the result of the limiting PIS case is reported for MCR dose.

At 450 seconds, the ADVs (and MSSVs) first open following the plant scram. It is assumed that all releases from the main steam lines are through the ADVs, which have worse (higher) X/Q values than the MSSVs. The 5% probability level X/Q values for the ADVs to the two control room intakes are:

47

MCR Atmospheric Dispersion Factors, 7/Q (s/mr3 )

East ADV to East ADV to West ADV to West ADV to East MCR West MCR East MCR West MCR Time AirIntake Air Intake Air Intake Air Intake 450 s - 2 hrs 1.06E-01 1.23E-03 1.36E-03 7.50E-03 2-8 hrs 7.45E-02 8.3 1E-04 8.29E-04 5.62E-03 The unfiltered in-leakage to the control room will be conservatively assumed to be subject to the worst case X/Q values, i.e., those for East ADV releases to East MCR Air Intake. Since about 98% of the releases are from SG #1, that is assumed to be the East SG, and it is assumed to be the source for all of the activity for the unfiltered in-leakage. This X/Q for the East ADV to the East MCR intake is applied to the entire unfiltered in-leakage for the 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> period for which a release is postulated.

Note that per NUREG-0800, Section 6.4, operator action to switch the assumed location of the control room emergency air intake to the more favorable location may be assumed. Per NUREG-0800, a X/Q corresponding to the X/Q for the more favorable intake, divided by a factor of 4, may be assumed. Thus, a reduced X/Q may be applied to the pressurization flow after the operator is assumed to select the WVest MCR Air Intake at 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> into the event:

MCR Atmospheric Dispersion Factors, 7/Q (s/mr3 )

After Operator Action to Select Preferred Air Intake East ADV to WVest West ADV to West Time MCR Air Intake MCR Air Intake 2-8 hrs 2.0775E-04 5.62E-03 Since each SG contributes to the source of the pressurization flow, a scaled X/Q can be developed to account for the relative contribution from each of the modeled sources. Thus, for each of the three dose cases (PIS, GIS, and Noble Gas), a scaled effective X/Q can be defined as:

x/Qcff= ((RI x X/Ql) + (R2 x X/Q2))/(Ri + R2) where R; is the release fraction for each source/volume (i.e., SG, or SG2 ) and X/Qi is the corresponding atmospheric dispersion factor. The R, values are based directly on the curie releases from the CENTS analyses documented in Section 2.13.6.3.2 of the EPU Licensing Amendment Request.

The effective control room X/Qs for the pressurization flow for each case are computed below:

48

y/lQ (s/M3): 0-120 min 2-8 hr SG, to MCR 0.106 0.0003075 SG2 to MCR 0.0075 0.0075 Release Fractions:

SG,, PIS 0.630811 0.350225 SG,, GIS 0.036274 0.951845 SG,,NG 0.087749 0.908001 SG 2 , PIS 0.007802 0.009339 SG2 , GIS 0.000709 0.01173 SG2 , NG 0.001448 0.002801 Effective X/Q Values (Filtered In-leakage):

PIS 0.10480 0.00049431 GIS 0.10411 0.00039506 NG 0.10440 0.00032962 8.2. Results The radiological consequence results in Rem TEDE are listed below and compared with the acceptance criteria for SGTR provided by RG 1.183, Table 6 and I OCFR50.67:

TEDE Dose Acceptance Criteria PIS case:

EAB (worst two hour dose) 0.494 25 Rem TEDE LPZ (duration) 0.143 25 Rem TEDE Main Control Room 4.93 5 Rem TEDE GIS case:

EAB (worst two hour dose) 0.0564 2.5 Rem TEDE LPZ (duration) 0.196 2.5 Rem TEDE 49

Because the same 5 Rem TEDE acceptance criteria applies for both the GIS and PIS case, only the limiting PIS case is reported for MCR dose. Note, as discussed in Section 8.0 above, there is conservatism in the PIS dose in that approximately 35% of the releases are assumed to occur after 6.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> into the event, where it is conservatively assumed that the operator opens the ADVs to control SG level.

Thus, the radiological consequences for SGTR are < 25 Rem TEDE for the EAB and LPZ doses and < 5 Rem TEDE for the MCR for the PIS case. Radiological consequences are < 2.5 Rem TEDE for EAB and LPZ doses for the GIS case. These are based on a 540 gpd primary-to-secondary leak rate for the unaffected SG and maximum control room unfiltered in-leakage of 250 CFM.

Activity releases for the SGTR are as follows:

DEI-131 Release Noble Gas Release (Table 1-4 isotopic distribution) 2 hr EAB (Ci) 8 hr LPZ (Ci) 2 hr EAR (Ci) 8 hrLPZ (Ci)

PIS 80.232 125.633 6237.2 69925.5 GIS 6.534 176.682 6237.2 69925.5 50

TABLE 8-1 ASSUMPTIONS USED FOR SGTR RADIOLOGICAL ANALYSIS Core Power Level: 3735 MWt RCS Noble Gas Activity: See Table 1-4 Core Inventory: See Table 1-1 RCS Initial Activity: I 00/E fiCi/gm Pre-existing Iodine Spike (PIS): 60 gtCi/gm DEI-131 Accident Generated Iodine Spike (GIS): 1.0 [tCi/gm DEI-131 Iodine Spiking Factor: 500 Secondary Coolant Initial Activity: 0.1 pCi/gm DEI-13 1 Fraction of Fuel Rods in Core Failing: 0%

Iodine Chemical Form:

Elemental 97%

Organic 3%

Particulate 0%

Primary-to-Secondary Leak Rate (unaffected SG): 540 gpd Steaming PF: 100 Steam Releases:

Affected SQ time of reactor trip to isolation 46,510 Ibm Intact SQ time of reactor trip to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 439,400 Ibm Intact SQ 2-8 hours 558,700 Ibm Duration of Release: 8.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> Control Room Parameters See Table 1-2 Main Control Room X/Q assumed: See Section 8.1.5 Pressurization Flow: 300 CFM (maximum) 50-

- - _(m -I --- -after u 2hu 50 CFM (minimum after 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />) 0 CFM (minimum, 0-2 hours)

Unfiltered In-leakage 250 CFM 51

9.0 OUTSIDE CONTAINMENT MAIN STEAM LINE BREAK/FEEDWATER LINE BREAK (OC-MSLB/FWLB)

As documented in Sections 2.13.1.3 and 2.13.2.3 of the EPU license amendment request (Reference 4), the OC-MSLB and the FWLB will be documented/quantified together in terms of dose assessment. Also, it is stated that no fuel failure is predicted or allowed for the OC-MSLB as well as for the FWLB. These sequences are quantified together since, for the radiological analysis models, the plant response and accident progression characteristics for these two events are similar.

The faulted or affected SG share sufficient plant response characteristics that their releases are quantified as a single release for both scenarios. In both cases, the unaffected SG is used to cooldown and depressurize the primary system to cold shutdown conditions using the ADV to remove decay heat. However, for the FWLB, the unaffected SG undergoes significant depletion of its inventory prior to reactor trip. Therefore, it is conservative to use the FWLB releases from the unaffected SG for this analysis.

The OC-MSLB would experience closure of the MSIVs at about the same time as the reactor trip on low steam generator pressure. Prior to this point, the feedwater system, sensing the increased steam flow would increase feedwater delivery. Therefore, since the outside containment MSL break location is limited in size by the integral flow restrictors, minimal inventory depletion would be experienced. Inventory loss is minimal, even for the larger inside containment MSLB as demonstrated in Figures 2.13.1.3.1-10 and 2.13.1.3.1-23 of Reference 4.

Single failure of a DC bus resulting in the failure of controls for one ADV is not considered. If a SG is assumed to have an ADV that is inoperable due to the assumed loss of a DC bus, the activity would be contained in that SG before manual action is assumed to operate the ADV.

Under these conditions, delays in taking manual control of the ADV would postponeihe releases due to cooldown to later in the event, when atmospheric dispersion values have decreased and allowing more time for operator actions to select the preferred control room air intake.

The analysis considers the secondary steaming pathway from the intact SG. This is a minor contributor compared to the releases from the broken SG, contributing less than 1%to the overall doses. The operators are assumed to not commence cooldown for this event until 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> into the event when the level has recovered to above the top of the U-tubes; this is consistent with the assumptions on PF. At that point, a cooldowvn at 50'F/hr is assumed such that shutdown cooling is entered and releases from the intact SG are secured at 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> into the event. Note that the cooldown analysis neglects any cooldown that would occur due to the break itself. Also note that for the OC-MSLB, level would be recovered in significantly less time than the 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> assumed herein. That level recovery time is based on FWLB analyses, including the assumption that the break is connected to the bottom of the steam generator liquid inventory. In reality, a FWLB communicates to the SG through the feedring, which is located relatively high in the SG. Thus, a realistically modeled FWLB behaves in a similar manner to a MSLB and the time until the top of the U-tubes would be recovered would be significantly less than 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

9.1. Input Parameters and Assumptions A summary of input parameters and assumptions is provided in Table 9-1. Certain assumptions are discussed in additional detail below.

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9.1.1. Source Term No fuel damage is predicted for the OC-MSLB or FWLB event as documented in the EPU Licensing Amendment Request. Two cases of iodine spiking are considered:

1. A PIS case where a reactor transient is postulated to have occurred prior to the event and has raised primary coolant iodine concentration to the maximum value (60 [ICi/gm DEI-131) allowed per Technical Specifications.

2. A GIS case, where the primary system transient associated with the event causes an iodine spike in the primary system. A spiking factor of 500 is assumed.

A maximum RCS DEI-131 activity of 1.0 ItCi/gm is assumed. Initial reactor coolant isotopic activity distribution is given by Table 14. This distribution is based on predicted activity distributions for power uprate conditions and is applicable for events for which no fuel failure will occur. The activity of Table 1-4 is adjusted to correspond to the Technical Specification activity limit of 100/E ItCi/gm.

A maximum Technical Specification secondary activity of 0.1 ILCi/gm DEI-131 is assumed.

Since no fuel failure is postulated, the small contribution to dose consequences from alkali metals has been ignored.

9.1.2. Iodine Chemical Form Per RG 1.183, iodine releases from the SGs to the environment are assumed to be 97% elemental and 3% organic. Since the same filter efficiency is specified for all iodine forms, there is no impact on the results of this assumption.

9.1.3. Release Pathways Consistent with RG 1.183, Table 6, the OC-MSLB and FWLB events are analyzed for a release duration until the affected SG is isolated and until cold shutdown is established for the unaffected SG (36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />). As discussed in Section 2.6.4.4 of Reference 4, Waterford-3 can achieve cold shutdown conditions under natural circulation conditions within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> of shutdown.

The release pathway for the intact SG is due to secondary steaming with radiological releases resulting from primary-to-secondary leakage from the RCS into the SGs. Activity is assumed to be released to the environment through the use of the ADVs to remove decay heat and to cool the plant to cold shutdown. Once cold shutdown is reached, the release to the environment is terminated. The doses due to releases from the affected and unaffected (intact) SGs are added to obtain the total dose.

The release pathway for the affected SG is due to primary-to-secondary leakage from the RCS to the secondary side of the SGs. Due to the postulated secondary system piping failures, activity transferred to the secondary side is assumed to be released immediately to the environment.

Activity is assumed released until cold shutdown is achieved.

The primary-to-secondary leak rate for the intact SG is assumed to be 150 gpd. The primary-to-secondary leak rate for the faulted SG is assumed to be 540 gpd.

53

9.1.4. Removal Coefficients A PF of 1.0 was assumed for the duration of the event for the affected SG For the intact SQ a PF of 1.0 was assumed for the first 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to account for the inventory loss due to the initial transient and the time to recover the SG U-tubes. A PF of 100 for the non-faulted SG is subsequently assumed.

9.1.5. Main Control Room Model.

The MCR ventilation model is described in Section 1.3. The OC-MSLB/FWLB dose model for secondary steaming assumes an unfiltered in-leakage of 100 CFM for the event duration. It is conservatively assumed that the pressurized mode is initiated at the start of the event and that the preferred control room intake is selected at two hours into the event. The filtered intake in the pressurized mode is assumed to be 225 CFM. The out-leakage from the control room envelope was assumed to be equal to the unfiltered in-leakage rate (100 CFM) during the first two hours and equal to the unfiltered (100 CFM) plus the minimum pressurization flow (50 CFM) for the accident duration.

9.2. Results The radiological consequence results in Rem TEDE are listed below and compared to the acceptance criteria provided by RG 1.183, Table 6 and IOCFR50.67:

TEDE Dose Acceptance Criteria PIS case:

EAB (worst two hour dose) 0.2219 25 Rem TEDE LPZ (duration) 0.0774 25 Rem TEDE Main Control Room 2.2029 5 Rem TEDE GIS case:

EAB (worst two hour dose) 0.2262 2.5 Rem TEDE LPZ (duration) 0.1240 2.5 Rem TEDE Main Control Room 3.6158 5 Rem TEDE Thus, the radiological consequences for OC-MSLB/FWLB are < 25 Rem TEDE for the EAB and LPZ doses and < 5 Rem TEDE for the MCR for the PIS case. Radiological consequences are <

2.5 Rem TEDE for EAB and LPZ doses and < 5 Rem TEDE for the GIS case. These are based on a 150 gpd primary-to-secondary leak rate for the unaffected SG and maximum control room unfiltered in-leakage of 100 CFM.

54

TABLE 9-1 ASSUMPTIONS USED FOR OC-MSLB/FWLB RADIOLOGICAL ANALYSIS Core Power Level: 3735 MWt Core Inventory: See Table 1-1 Secondary Steaming Pathvay Primary-to-Secondary Leak Rate: 150 gpd unaffected SQ 540 gpd faulted SG Iodine Chemical Form (Reactor Building Release Path):

Elemental 97%

Organic 3%

Steaming PF (Iodine and Alkali Metals):

4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> 1

> 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> 100 Duration of Release: 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> Control Room Parameters See Table 1-2 Affected SG is Assumed to be the East SG.. MCR X/Q Assumed:

Unfiltered In- Filtered In- Filtered In- Filtered In- Filtered In-Time leakage, East leakage, East leakage, East leakage, WVest leakage, WVest ADV to East ADV to East ADV to WVest ADV to East ADV to WVest MCR Air Intake MCR Air Intake MCR Intake MCR Air Intake MCR Air Intake 0-2 hr 1.06E-01 1.06E-01 NA 1.36E-03 NA 2-8 hr 7.45E-02 NA 2.08E-04

• NA 1.411E-03

• 8-24 hrs 3.30E-02 NA 1.OOE-04

• NA 6.42E-04

• 1-4 days 2.3 1E-02 NA 6.58E-05

• NA 5.1 OE-04

• 4-30 days 1.62E-02 NA 4.63E-05

• NA 3.93E-04 *

• factor of 4 reduction credited per SRP 6.4.

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TABLE 9-1 (Cont.)

ASSUMPTIONS USED FOR OC-MSLB/FWLB RADIOLOGICAL ANALYSIS Steaming (Ibm) and Activity (DEI-13 1, Ci) Releases 0-2 hr Steaming 2-8 hr Steaming 588,365 1,333,286 150 gpd Leakage 0-15 min 15-30 min A2-1 hr 1-2 hr 2-4 hr 4-6 hr 6-8 hr PIS 16.00 6.72 1.14 2.08 4.36 2.66 2.67 GIS 15.72 6.19 0.40 1.04 4.09 4.27 5.93 540 gpd Leakage 0-15 min 15-30 min /2-1 hr 1-2 hr 2-4 hr 4-6 hr 6-8 hr PIS 16.80 8.34 3.78 7.39 15.55 9.47 9.53 GIS 15.80 6.43 1.14 3.72 14.57 15.21 21.15 56

10.0 CONTROL ELEMENT ASSEMBLY (CEA) EJECTION For the CEA Ejection dose analysis per RG 1.183, two different fission product release paths are to be considered:

1. Fission product releases via SG steaming (releases via ADVs or MSSVs). Per RG 1.183 Table 6, this pathway is analyzed until cold shutdown is established.

2. Fission product releases via normal containment leakage. Per RG 1.183 Table 6, this pathway is analyzed for a 30-day release duration.

Each of these release pathways must independently meet the RG 1.183 AST dose limits defined in Section 10.2.

The secondary steaming pathway assumes that a CEA has been ejected concurrent with a LOOP and that releases are occurring due to primary-to-secondary leakage from the RCS to the SG.

Activity is then released to the environment through the use of the ADVs to remove decay heat and to cool the plant to shutdown cooling entry conditions. Since the control room X/Q values are worst for ADV releases, any releases which would occur through the MSSVs are instead assumed released from the ADV locations. Once shutdown cooling is initiated, the release to the environment is terminated for this pathway.

The fission product release via normal containment leakage assumes that the ejected CEA has caused a small break in the RCS coolant boundary. This results in the release to containment of the activity, due to the transient, in the RCS water inventory. A maximum limit of 15% fuel failure due to DNBR was established in and reported in the EPU License Amendment Request, Reference 4. No fuel failure due to incipient fuel melt is assumed or allowed, consistent with the 3716 MWt EPU License Amendment Request. The containment building release pathway model is similar to and derived from the LBLOCA release model. Fission product removal via the operation of containment sprays is conservatively not credited for this calculation. The LBLOCA ESF release pathway is considered small and is ignored for purposes of CEA Ejection sequences per RG 1.183. Also, per RG 1.183, it is conservatively assumed that all available fission product activity is released to the containment within a very short time for this case. This analysis assumes a 0.5% per day by volume containment leakage rate for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, then 0.25% by volume per day aftervard, consistent with the guidelines in RG 1.183.

10.1. Input Parameters and Assumptions The input parameters and assumptions for the CEA Ejection analysis are listed in Table 10-1.

Certain inputs and assumptions are discussed in additional detail below.

10.1.1. Source Term As a result of the CEA Ejection accident, a 15% fuel failure was reported in the EPU report. For the purpose of the CEA Ejection analysis, a conservative range of fuel failures was analyzed (up to 15%). The assumed fuel failure mechanism is up to 15% of the fuel rods in the core experiencing DNB. The non-LOCA gap fractions specified in Table 3 (Footnote 11) of RG 1.183 are selected for use in the CEA Ejection analysis to provide the most conservative set of results.

These gap fractions are 10% for iodines and noble gases and 12% for alkali metals.

57

No fuel failure due to incipient fuel melt is assumed, consistent with the 3716 MWt EPU License Amendment Request.

Per Appendix H of RG 1.183, independent calculations have been performed and show that the containment sump pH will be at or above 7.0, therefore no source term due to iodine re-evolution has been considered.

10.1.2. Iodine Chemical Form For the reactor containment leakage pathway, the chemical form of the radioiodine released is assumed to be 95% particulate, 4.85%. elemental iodine, and 0.15% organic iodine. This is consistent with the guidelines provided in Appendix H of RG 1.183.

For the secondary steaming release pathway, the iodine releases from the SG to the environment are assumed to be 97% elemental iodine and 3% organic iodine. This is consistent with the guidelines provided in Appendix H of RG 1.183.

10.1.3. Release Pathways Conservatively, all the iodine, alkali metal and noble gas activity due to the postulated CEA Ejection accident is assumed to be in the primary coolant when determining the dose consequences due to primary-to-secondary SG leakage and subsequent secondary steaming.

Releases are assumed to be terminated once shutdown cooling is initiated and the SGs are no longer providing decay heat removal (no further releases would occur due to cooldown to cold shutdown conditions). This is consistent with the guidelines provided in Table 6 of RG 1.183.

The two separate release path results are not to be summed together when calculating the TEDE doses for comparison to the RG 1.183 AST dose limits.

For the purposes of the CEA Ejection analysis, a primary-to-secondary SG leakage of 150 gpd is assumed. This value is consistent with the Technical Specification change to which Entergy committed in Reference 11 and is conservative with respect to a planned Technical Specification change request to further reduce this limit to 75 gpd.

Operator actions to select the more favorable of the control room air intakes in terms of X/Q are credited to occur after 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

10.1.4. Removal Coefficients For the secondary steaming path, iodine and alkali metal releases to the secondary side via primary-to-secondary side SG leakage are assumed to be subject to a PF. Consistent with RG 1.183, Section 5.5, a PF of 100 is assumed for iodines and alkali metals. For the sake of conservatism, a PF of 10 is assumed for the first 30 minutes of the event to account for potential elevated releases due to the initial transient.

Per RG 1.183, all noble gases released to the secondary side via primary-to-secondary side SG leakage are assumed to be immediately released to the environment.

For the Reactor Containment Leakage release pathway, the Powers 10% Aerosol Decontamination Factor is assumed for natural deposition. This model is contained in the RADTRAD analysis code documented in NUREG/CR-6604. A natural deposition factor of 0.4/hour is assumed for elemental iodine and 0.0 for organic iodine.

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No credit is taken for the removal of fission products due to containment spray initiation for conservatism.

10.1.5. Main Control Room Model The MCR ventilation model is described in Section 1.3. The CEA Ejection dose model for secondary steaming release pathway conservatively assumes a constant unfiltered in-leakage of 150 CFM for the entire duration of the secondary steaming release (7.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />). The CEA Ejection analysis assumes that the preferred control room intake is selected at two hours into the event.

The reactor containment building leakage pathway model also includes a conservative assumption of a constant 150 CFM unfiltered in-leakage throughout the event. Similar to what was done for the LBLOCA, the plant stack is the release location assumed for determining the X/Q for control room intake.

10.2. Results Acceptance limits for the PWR CEA Ejection transient per RG 1.183, Table 6 are:

EAB and LPZ: 6.3 Rem TEDE MCR: 5 Rem TEDE The radiological consequences in terms of Rem TEDE are listed below as a function of 15% fuel failure for an assumed primary- to-secondary SG leakage of 150 gpd and a control room unfiltered in-leakage of 150 CFM. Note, the results presented below conclude that the off-site EAB and LPZ TEDE doses are dominated by the reactor containment leakage pathway, whereas the MCR TEDE doses are controlled by the fission product release pathway (noble gases plus iodines and alkali metals) from secondary side SG steaming:

Secondary Steaming Reactor Building Acceptance 15%FuelFailure Release Pathway Leakage Pathway Criteria EAB (worst two hour dose) 0.552 1.028 25 Rem TEDE LPZ (duration) 0.213 0.653 25 Rem TEDE MCR 3.19 0.680 5 Rem TEDE Thus, the radiological consequences for the CEA Ejection are < 6.3 Rem TEDE for the EAB and LPZ doses and <5 Rem TEDE for the MCR, based on 15% fuel failure, a 150 gpd primary-to-secondary leak rate per SG and maximum control room unfiltered in-leakage of 150 CFM in recirculation mode.

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TABLE 10-1 ASSUMPTIONS USED FOR CEA EJECTION RADIOLOGICAL ANALYSIS Core Power Level: 3735 MWt Core Inventory: See Table 1-1 Fission Product Gap Fractions:

Iodines 10%

Noble Gases 10%

Alkali metals (Cs & Rb-86) 12%.

Fraction of Fuel Rods in Core Failing (maximum): 15%

Reactor Building Release Pathway Containment Leak Rate: 0.50 % volume/day (0-24 hours) 0.25 % volume/day (24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> - 30 days)

Natural Deposition:

Elemental 0.40/hr Organic 0 Particulate Powers 10% Aerosol Decontamination Factor Spray Fission Product Removal: Not Credited Iodine Chemical Form (Reactor Building Release Path):

Elemental 4.85%

Organic 0.15%

Particulate 95%

Secondary Steaming Pathlway Primary-to-Secondary Leak Rate: 150 gpd per SG Iodine Chemical Form (Reactor Building Release Path):

Elemental 97%

Organic 3%

Particulate 0%

Steaming PF (Iodine and Alkali Metals):

0-30 minutes 10

> 30 minutes 100 Duration of Release: 7.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> Control Room Parameters: See Table 1-2 60

TABLE 10-1 (Cont.)

ASSUMPTIONS USED FOR CEA EJECTION RADIOLOGICAL ANALYSIS Main Control Room X/Q Assumed:

Reactor Building Reactor Building Secondary Steaming Secondary Steaming Time Unfiltered In-leakane Pressurization Flow Unfiltered In-leakage Pressurization Flow 0-2 hr 2.77E-03 NA 1.06E-01 NA 2-8 hr 1.78E-03 3.90E-04

• 7.45E-02. 2.08E-04 *

• factor of 4 reduction credited per SRP 6.4.

Steaming (Ibm) and Activity (DEl-I 31, Ci) Releases 0-2 hr Steaming 2-8 hr Steaming 609,744 858,838 0-15 min 15-30 min 1/2-I hr 1-2 hr 2-4 hr 4-6 hr 6-7.5 hr 2.70 3.54 1.73 6.03 17.73 23.16 19.56 Alkali Metal Source Term Data, Ci Released:

Cs-134 18.506 Cs-136 4.866 Cs-137 9.855 Rb-86 0.035 61

11.0 FUEL HANDLING ACCIDENT (FHA)

The dropping of a fuel assembly is assumed to occur breaching the cladding and releasing the volatile fission products in the gas gap of the fuel pins. In addition to the area radiation monitor located in the spent fuel cask area, portable radiation monitors capable of emitting audible alarms are located in this area during fuel handling operations. Doors in the fuel handling building are closed to maintain controlled leakage characteristics in the spent fuel pool region during refueling operations involving irradiated fuel. Should a fuel assembly be dropped in the fuel transfer canal or in the spent fuel pool and released radioactivity be above a prescribed level, the airborne radiation monitors sound an alarm alerting personnel to the problem.

The MCR will be isolated upon detection of radioactive contamination at the control room air intakes. As discussed in Section I of this report, a study was performed to verify the assumptions that the control room is isolated on a high radiation signal before any radiation enters the control room envelope. Specifically, it was confirmed that the control room boundary will be isolated prior to contamination in the form of fission products entering the ductwork and reaching the control room envelope.

The analysis presented in the 3716 MWt EPU License Amendment Request, Reference 4, demonstrated the acceptability of the FHA analysis using a TID source term for an unfiltered in-leakage of 100 CFM.

11.1. Input Parameters and Assumptions 11.1.1. Source Term This calculation uses 15 isotopes modeled by the RADTRAD code as documented in NUREG/CR-6604. The isotope quantities are unchanged from what is currently listed in the

\Vaterford 3 UFSAR and are listed in Table 11-1. Consistent with the assumption of an instantaneous release, no decay is considered. The RG 1.183 assumptions for values of core inventory release fractions and gap iodine species are applied as modified by NUREG/CR-5009 (Reference 15). Pool PFs from RG1.183 are applied as well. Since particulate radionuclides (i.e.,

cesium and rubidium) are completely retained by the water pool, these nuclides are not modeled in this calculation. Consistent with the current licensing basis 1-132 was conservatively modeled using the longer half-life of its precursor, Te-132. The source terms are assumed to have decayed for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> in accordance with Waterford 3 Technical Specification 3/4.9.3. The entire FHA source term is assumed to be instantaneously released to the release volume.

11.1.2. Iodine Chemical Form The chemical form of radioiodines released from fuel pins in this calculation is assumed to be 95% cesium iodide, 4.85 % elemental iodine and 0.15 % organic iodide. This is consistent with the guidance of RG 1.183,Appendix B, Section 1.3.

11.1.3. Release Pathway Four possible release locations were evaluated for the FHA: the plant stack, the containment equipment hatch, the personnel airlock doors and the fuel handling building truck bay. The limiting release location for the FHA in the fuel handling building or the reactor building is the plant stack. Therefore, in this model, the resultant release of radioactivity is conservatively 62

assumed to be exhausted immediately (no holdup time) from the fuel handling building via the plant stack.

11.1.4. Removal Coefficients No Fuel Handling Building filtration or recirculation is credited in the analysis. This includes the fuel handling building filtration system. The Waterford 3 pools maintain at least 23 feet of water coverage above damaged fuel in accordance with Technical Specifications. This calculation applies the RG 1.1 83 pool overall iodine PF value of 200 for drops over the fuel racks.

11.1.5. Main Control Room Model The MCR ventilation model is described in Section 1.3. The FHA dose model assumes a constant unfiltered leakage of 200 CFM. This is conservatively large compared to the minimum assumed in-leakage of 100 CFM being used for most events. The filtered intake in the pressurized mode is assumed to be 225 CFM. Since an instantaneous release is assumed, the filtered intake is conservatively initiated at the start of the event. No credit for operator selection of the more favorable of the two emergency air intakes is assumed.

11.2. Results The radiological consequence results in Rem TEDE are listed below and compared with the acceptance limits for FHA provided by RG 1.183 and I OCFR50.67:

FHA Acceptance Criteria EAB (worst two hour dose) 0.55 6.3 Rem TEDE LPZ (duration) 0.085 6.3 Rem TEDE MCR 0.19 5 Rem TEDE Thus, the radiological consequences for the FHA are < 6.3 Rem TEDE for the EAB and LPZ doses and < 5 Rem TEDE for the MCR, based on a maximum control room unfiltered in-leakage of 200 CFM throughout the event.

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TABLE 11-1 ASSUMPTIONS USED FOR FHA RADIOLOGICAL ANALYSIS Core Power Level: 3735 MWt Core Inventory: (Ci/MWt)

Kr-85 4.520E+02 Kr-85m 1.378E-01 Kr-87 5.849E-14 Kr-88 2.071 E-04 1-131 9.867E+01 I-132 3.985E+01 I-133 9.764E+00 1-134 9.488E-23 1-135 4.991E-02 Xe-131m 1.551E+02 Xe-133 1.667E+04 Xe-133m 3.713E+02 Xe-135 2.223E+02 Xe-135m 1.625E+00 Fission Product Gap Fractions:

I-131 12%

Kr-85 14%

Other Noble Gases 5%

Other Halogens 5%

Alkali Metals 12%

.Fuel Rods Failing (maximum): 60 rods Iodine Chemical Form*:

Elemental 4.85%

Organic 0.15%

Particulate 95%

t he releases from the pool are conservatively modeled as 99.85% elemental and 0.15% organic iodine Control Room Parameters: See Table 1-2 Main Control Room X/Q Assumed:

Tie East Control Room Time Intake via Plant Stack 0-2 hr 2.77E-03 2-8 hr 1.78E-03 8-24 hrs 7.22E-04 1-4 days 5.2iE-04 4-30 days 4.05E-04 64

Attachment 3 W3FI -2004-0053 List of Regulatory Commitments W3FI-2004-0053 Page 1 of 1 List of Regulatory Commitments The following table identifies those actions committed to by Entergy in this document. Any other statements in this submittal are provided for information purposes and are not considered to be regulatory commitments.

TYPE (Check one) SCHEDULED ONE- CONTINUING COMPLETION COMMITMENT TIME COMPLIANCE DATE (If ACTION Required)

A second supplemental AST submittal will present X August 8, calculated dose results for the following events: 2004

• Reactor Coolant Pump Seized Rotor / Sheared Shaft

• Inadvertent Atmospheric Dump Valve Opening

• Excess Main Steam Flow with Loss of Offsite Power

• Letdown Line Break