ML041600563

Response to Request for Additional Information Regarding License Amendment Request 03-02

ML041600563
Person / Time
Site:	Seabrook
Issue date:	05/24/2004
From:	Warner M Florida Power & Light Energy Seabrook
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NYN-04046
Download: ML041600563 (30)

Text

FPL Energy Seabrook Station FPL Energy P.O. Box 300 Seabrook, NH 03874 Seabrook Station (603) 773-7000 MAY 2 4 2004 Docket No. 50-443 NYN-04046 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D.C. 20555-0001

Reference:

1. FPLE Seabrook Letter NYN-04039, Seabrook Station Response to Request for Additional Information Regarding License Amendment Request 03-02," dated May 5, 2004.

Seabrook Station Response to Request for Additional Information Regarding License Amendment Request 03-02 Enclosed is the FPL Energy Seabrook, LLC (FPLE Seabrook) response to requests for additional information associated with License Amendment Request (LAR) 03-02 received on January 28, 2004, March 1, 2004 and March 23, 2004.

The initial responses to the RAIs were submitted on May 5, 2004 (Ref. 1). At that time, the responses to several RAIs were still being developed.

The enclosed response specifically addresses RAI 2 from the request for additional information dated March 1, 2004 and RAI 5,

6dW, 6d2, 6d4, 6e4, 6g and 6i from the request for additional information dated March 23, 2004 and completes the RAI responses.

For continuity, this response also includes the information previously provided to each of the three requests. No changes have been made to previous responses.

Should you have any questions concerning this response, please contact Mr. James M. Peschel, Regulatory Programs Manager, at (603) 773-7194.

Very truly yours, FPL Energy Seabrook, LLC Mark f Warner Site Vice President te' an FPL Group company

U. S. Nuclear Regulatory Commission NYN-04046 /Page 2 cc:

H. J. Miller, NRC Region I Administrator V. Nerses, NRC Project Manager, Project Directorate 1-2 G. T. Dentel, NRC Senior Resident Inspector Mr. Bruce Cheney, Director New Hampshire Office of Emergency Management State Office Park South 107 Pleasant Street Concord, NH 03301 OATH AND AFFIRMATION I, Mark E. Warner, Site Vice President of FPL Energy Seabrook, LLC, hereby affirm that the information and statements contained within this response to the Request for Additional Information to License Amendment Request 03-02 are based on facts and circumstances which are true and accurate to the best of my knowledge and belief.

Sworn and Subscribed before me this a9 tSday of

/) Liz/

.2004 Mark E. Warner Site Vice President

/4a24(

Notary Public

Enclosure to NYN-04046 Responses to NRC Request for Additional Information Seabrook Station Alternative Source Term

Seabrook AST RAI Responses-First Set dated 1/28/04 RAI I The fission products which result from significant core damage can be limited from escaping the containment environment through chemical means.

In particular, elemental iodine formed during the accident can be held in the liquid phase in the sump if the sump is at a pH of 7.0 or greater.

Sump pH is dependent on the post-LOCA production of acids (Hydrolodic Acid, Hydrochloric Acid, and Nitric Acid) and the post-LOCA production and/or addition of bases (Cesium Hydroxide and Sodium Pentaborate).

1. The staff requests the licensee to provide additional details related to the assumptions used in calculating sump pH. Specifically, the staff requests the licensee to list the sources of post-LOCA acid generation affecting the sump pH and the mechanism of acid formation for each source (e.g., formation of HCL from the decomposition of cable insulation).

In addition, the staff requests a time dependent list of the calculated post-LOCA acid generation and sump pH values similar to the following example table.

l Time (hours) l HCL moles l HI moles l HNO3 l pH 2

102

.3 4

7.12 l

FPLE Response to RAI 1:

Following a LOCA, borated water is added to the containment sump from the Refueling Water Storage Tank (RWST) and the Reactor Coolant System. In order to maintain the required pH of the Containment Spray and the containment sump, sodium hydroxide from the Spray Additive Tank (SAT) is mixed with the borated water in the mixing chamber of the Refueling Water Storage Tank. The borated water from the Refueling Water Storage Tank and the sodium hydroxide from the Spray Additive Tank are added to the containment sump within one hour after a LOCA. The descriptions of the Containment Spray and Safety Injection systems are provided in the Seabrook Station UFSAR Section 6.2.2. Achievement of a sump pH greater than 7 occurs early in the time interval where processes leading to the formation of elemental iodine by radiolysis occur. Following this initial chemical addition/homogenation, the sump pH would only be impacted from the nitric acid produced from irradiation of water, the hydrochloric acid produced by irradiation of electrical cable insulation, and sump water temperature changes. Note that the analysis conservatively does not consider the addition of base forming components such as cesium hydroxide, which would increase pH.

The minimum containment sump pH was developed based on using the combination of boric acid and sodium hydroxide concentrations and tank volumes that would produce a minimum pH value. In addition to the boric acid, nitric and hydrochloric acids produced from irradiation of water, and electrical cable insulation were included.

The nitric and hydrochloric acids were developed using the guidelines provided in NUREG/CR-5950, "Iodine Evolution and pH

Control," using 30-day integrated doses in the sump and containment atmosphere at the stretch power uprate conditions.

The containment sump pH at 30-days was determined to be greater than 8.0 at a nominal temperature of 860F. This is the lowest pH value expected to be found inside containment during the 30-day evaluation period after sump mixing and well above the minimum required value of 7 to control the re-evolution of iodine. The use of 860F in the analysis is conservative because, in the highly buffered system of boric acid and caustic, the pH increases with increasing temperature.

Since all of the boric acid and sodium hydroxide is added to the containment sump within one hour, the moles of nitric (479.2 g-mol) and hydrochloric (1.6E3 g-mol) acids formed are based on a 30-day integrated dose and the 30-day pH value is well above 7, a time dependent sump pH with component inventories has not been provided. The pH would actually be higher, but less than the maximum pH limit of 10.5, during the early time period because smaller quantities of nitric and hydrochloric acids would be produced.

The 30-day containment sump pH was determined with and without the additional acids produced from the irradiation of water and electrical cable insulation.

The additional acid reduces the final pH by less than 0.1 pH units. Hydriodic acid was not considered in this evaluation since only small amounts are released from the core and are much less than the hydrochloric and nitric acid present.

RAI 2

The staff requests the licensee to demonstrate how the sump pH will be maintained above 7 for the period of 30 days.

FPLE Response to RAI 2:

As discussed in FPLE's response to RAI 1 above, the containment sump pH at 30 days is well above the required value of 7 to control the re-evolution of iodine. The containment sump 30 day pH value is the minimum value for pH after depletion of the RWST and SAT in approximately 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

RAI 3

If the sump pH was determined through the use of a computer code, the staff requests the licensee to identify the code and to detail the input and output of the code.

If the sump pH was determined through another analysis, the staff requests the licensee to provide details of the calculations including the assumptions.

FPLE Response to RAI 3:

A computer program was not used to develop the containment sump pH. As stated above, the pH value was determined at 30-days which is a lower value than expected in the containment sump at approximately I hour when recirculation begins. A manual calculation was performed to determine the pH value at 30-days using the inputs described in Table 1 below.

Table 1 Input Description Value Refueling Water Storage Tank Volume Released 428,000 gallons (Maximum Value @ 980F)

Refueling Water Storage Tank Boron Concentration 2900 ppm (Maximum Value)

Spray Additive Tank Volume Released (Minimum 8,520gallons Value @ 50'F)

Spray Additive Tank Sodium Hydroxide 19 wt.%

Concentration (Minimum Value)

Reactor Coolant System Mass 492,200 Ibm Reactor Coolant System Boron Concentration 4,000 ppm (Maximum Value)

Accumulator Volume (Maximum Value @ 100IF) - 4 6,596 gallons/accumulator Accumulators Accumulator Boron Concentration (Maximum Value) 2900 ppm Mass of Electrical Cable Insulation' 50,000 Ibm Post-LOCA 30-Day Integrated Radiation Dose in 3.3E07 rads Containment Sump water at SPU Conditions (Beta &

Gamma)

Post-LOCA 30-Day Integrated Radiation Dose in 7.1E07 rads Containment Air at SPU Conditions (Beta & Gamma)

Notes:

1. Moles of acid formed due to the decomposition of cable insulation are presented in the response to the first set of RAIs -RAI 1.

Seabrook AST RAI Responses-Second Set dated March 1, 2004

RAI 1

Provide the 1998 through 2002 hourly meteorological data used in ARCON96 calculations and joint wind speeds, wind direction and atmospheric stability distributions (jfd) used in the PAVAN calculations. If these data have already been provided on the docket, please cite an appropriate reference. Please specify if the jfd data are formatted as discussed on Page 14 of Enclosure 2 to NYN-03061 (Enclosure 2) provided by letter dated October 6, 2003, or if they are the reformatted data input to the PAVAN calculations.

FPLE Response to RAI 1:

The 1998 through 2002 meteorological data used in the ARCON96 calculations and joint wind speeds, wind directions and atmospheric stability distributions (jfd) used in the PAVAN calculationswere submitted with the RAI response dated May 5, 2004.. The jfd data used to determine the offsite X/Q values were reformatted for the PAVAN code input in the PAVAN input file (seabrookinput.dat). The process to reformat the data from the raw jfd files is discussed on page 14 of Enclosure 2 to NYN-03061.

The lower meteorological measurement height is 10.05 m and the upper is 60.66 m. The windspeed values are in miles per hour.

RAI 2

Page 15 of Enclosure 2 states that the exit velocity from the MSSVs is greater that the 95 percentile wind speed for the first 2 I/2 hours of the events during which a release is postulated to occur. What are the estimated exit velocities, flow rates, and pressures as a function of time for the MSSVs and ASDVs during this interval? What is the basis for the estimates? The 95 percentile wind speeds are estimated to be 16.72 and 16.81 miles per hour. How were these values estimated?

FPLE Response to RAI 2:

The X/Q reduction factor of five was applied to Main Steam Safety Valve (MSSV) and Atmospheric Steam Dump Valve (ASDV) releases for the first 2.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> of three events:

(1) Main Steam Line Break (MSLB), (2) Rod Cluster Control Assembly (RCCA)

Ejection, and (3) Steam Generator Tube Rupture (SGTR).

Rather than calculating exit velocities, flow rates and pressures versus time for the MSSVs and ASDVs, bounding minimum exit velocities during the first 2.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> of these events were determined. The minimum bounding exit velocities were then compared and shown to be above the 95th percentile wind speed for the release location.

In determining the bounding minimum exit velocity for the MSSVs, the following assumptions apply:

1. The smallest value flow rate is used for MSSVs since this will provide the lowest exit velocity and largest radiological effect in the vicinity of the control room.

This value includes setpoint tolerances and valve accumulation.

2. When the lowest MSSV setpoint is reached, the valve is assumed to fully open.

As such, the exit velocity is based on the flow rate when full open.

3. Flow velocity versus time for the MSSVs is not critical considering that the MSSVs will reclose at a pressure near the setpoint.

Based on the above, the bounding minimum exit velocity from an MSSV was determined to be 2642.6 fl/sec. This value is significantly greater than the 95 percentile wind speed of 122.7 ft/sec (the 95 percentile wind speed for the MSSVs of 16.72 mph increased by a factor of five to apply the X/Q reduction factor).

Please note that although the x/Q reduction factor was applied to the MSSV releases for the first 2.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> of the three events, the reduction factor could be applied to the entire eight hours of each event.

For the ASDVs a slightly different approach was utilized.

In order to demonstrate that the XIQ reduction factor could be applied to the first 2.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> of the MSLB and RCCA Ejection events, the ASDV exit velocity was calculated at 2.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> as follows:

1. First, a reactor coolant temperature was determined at 2.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> following initiation of the event using the maximum allowable cooldown rate. Note that conservatively no credit was taken for time at hot shutdown to allow systems to stabilize.

2. Next, a saturated liquid temperature for the Steam Generator was calculated based on the reactor coolant temperature determined above and subsequently the Steam Generator pressure was determined. Since the relationship of ASDV mass flow rate to inlet pressure is linear, an ASDV mass flow rate was determined from the inlet pressure.

3. Finally, the mass flow rate was converted to velocity and verified to be greater than the 95 percentile wind speed.

Based on the above, the bounding minimum exit velocity from an ASDV for the MSLB and RCCA Ejection events was determined to be 124.8 ftl/sec. This value is greater than the 95 percentile wind speed of 123.4 ft/sec (the 95 percentile wind speed for the ASDVs of 16.81 mph increased by a factor of five to apply the X/Q reduction factor).

Application of the X/Q reduction factor for the SGTR event requires further clarification because there are releases from intact and faulted steam generators and there are two ASDV failure cases.

For simplicity, the SGTR analysis in the License Amendment Request assumed that the X/Q reduction factor could be applied to ASDV releases for the first 2.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> of the SGTR event for both the intact and faulted steam generators. However, the X/Q reduction factor could be applied to all ASDV and MSSV releases from the faulted steam generator (the significant contributor to offsite dose) for the entire eight hours of the event, since the releases will be at higher steam generator pressures and thus higher exit velocities than the 124.8 ft/sec case described above. For the ASDV releases from the intact steam generators, the X/Q reduction factor should only be applied for the first 30 minutes of the event, since a rapid Reactor Coolant System cool-down is required beginning after 30 minutes. The application of the X/Q dilution factor for the first 2.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> of the SGTR event for ASDV and MSSV releases from the intact and faulted steam generators as presented in the LAR versus eight hours for the faulted steam generator releases and 30 minutes for intact steam generator releases is slightly conservative as demonstrated below.

Allowable Unfiltered EAR Dose Control Case CR (rem LPZ Dose Room Dose Inleakage TEDE)

(rem TEDE)

(rem (CfM)TEDE)

SGTR - Pre-accident Iodine Spike (Case 1 is limiting)

X/Q reduction factor applied for 2.5 300 3.78 1.85 2.07 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> for faulted and intact steam generator releases (LAR)

SGTR -

Pre-accident Iodine Spike (Case 1 is limiting)

X/Q reduction factor applied for 8 300 3.78 1.85 2.02 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for faulted steam generator releases and 30 minutes for intact steam generator releases Acceptance Criteria

_ 25

< 25

< 5 SGTR -

Concurrent Iodine Spike (Case 1 is limiting)

X/Q reduction factor applied for 2.5 300 2.21 1.09 1.38 hours4.398148e-4 days <br />0.0106 hours <br />6.283069e-5 weeks <br />1.4459e-5 months <br /> for faulted and intact steam generator releases (LAR)

SGTR -

Concurrent Iodine Spike (Case I is limiting)

X/Q reduction factor applied for 8 300 2.21 1.09 1.34 hours3.935185e-4 days <br />0.00944 hours <br />5.621693e-5 weeks <br />1.2937e-5 months <br /> for faulted steam generator releases and 30 minutes for intact steam generator releases Acceptance Criteria

< 2.5

< 2.5

< 5

To determine the 9 5 th percentile wind speed at the MSSV and ASDV release heights, the meteorological data used for the ARCON96 runs was evaluated. Hourly entries with bad data were neglected in the evaluation. Wind speed multipliers for an MSSV release and for an ASDV release were selected based upon the stability class for each hour of data. The wind speed multiplier and the 10 m wind speed were multiplied together to obtain the wind speed at the height of the release for each hour of data.

The wind speed multiplier is selected based upon the stability class, and is taken from ARCON96 case runs. The results of these case runs, present the wind speed correction factors for the MSSV and ASDV release location.

The valid hourlyZ release height wind speeds were then utilized to determine the 95th percentile value. The 95' percentile wind speed for the MSSV release height is 16.72 miles per hour.

The 9 5 th percentile wind speed for the ASDV release height is 16.81 miles per hour. That is, 95% of all of the hourly wind speeds at the MSSV release height are less than 16.72 miles per hour, and 95% of all of the hourly wind speeds at the ASDV release height are less than 16.81 miles per hour.

RAI 3

With regard to the ARCON96 relative concentration (X/Q) estimates provided in Table 1.8.1-2 of Enclosure 2, were all of the calculations based upon the same assumptions, other than 1) the differences noted in Table 1.8.1-1, 2) reduction by a factor of two when either the east or west intake was the postulated receptor location, or 3) when the diesel building intakes were assumed receptors and the X/Q value is a weighted average of the two intakes. In the third case, are the input values given in Table 1.8.1-1 for the more limiting intake and inputs for the less limiting intake not provided? Are the inflow rates assumed to be equal? Were any releases assumed to be diffuse or to have a vent flow? If so, what were the inputs? Are values provided in Table 1.8.1-2 applicable to loss of offsite power and single failure scenarios?

FPLE Response to RAI 3:

All of the calculations performed on the X/Q estimates are based on the same assumptions other than the items listed in this RAI question.

The only additional differences not listed in Table 1.8.1-1 are: (1) the cases with releases from the plant vent and containment building which credit building wake, and (2) the cases with the releases from the RWST with the diesel building intakes as the receptors which credit building wake in the ARCON96 case runs. The building area used for the plant vent and containment building release cases in the ARCON96 files is 1,506 m2. The building area used for the RWST to diesel building intake cases is 337 m2.

For the third case as stated in the RAI, where a weighted average of the two diesel building intakes is used, the values given in Table 1.8.1-1 are for the more limiting intake, and the less limiting intake values are not provided. The inputs for the case from the RWST to the non-limiting diesel intake are the same as the case to the limiting intake except that the distance is

52.4 m and the direction is 1370. The inflow rates for the diesel building intakes are assumed to be equal as previous analyses have shown.

Conservatively, no releases were assumed to be diffuse and no releases were assumed to have a vent flow.

The values provided in Table 1.8.1-2 are applicable to loss of offsite power and single failure scenarios.

Seabrook AST RAI Responses-Third Set dated March 23, 2004

RAI 1

Regarding the proposed technical specification change in the definition of "dose equivalent I-131," Seabrook uses the thyroid dose as the basis of the proposed change. This definition finds use in the Reactor Coolant System (RCS) and secondary specific activity technical specifications. The purpose of those technical specifications is to control the actual specific activities to levels less than those which would exceed the initial assumptions made in the radiological consequences analyses. Previously, those analyses determined whole body and thyroid does, consistent with the dose guidelines in 10CFR100.11.

However, with the proposed implementation of the Alternative Source Term (AST), the total effective dose equivalent (TEDE) criteria supercede the whole body and thyroid dose. The staff has not required licensees to revise this definition. Since you have proposed a change, please provide a justification for the use if thyroid dose conversion factors when the effective factors provided in Federal Guidance Report (FGR) 11 Table 2.1 would be more appropriate.

FPLE Response to RAI 1:

In the Seabrook dose calculations, the dose conversion factors referenced in the Technical Specification definition of dose equivalent I-131 (D.E. 1-131) are used to adjust the initial primary coolant iodine activities.

The primary coolant iodine activities adjusted to 1.0 ILCi/gm D.E. 1-131 based on the thyroid dose conversion factors are:

Primary Coolant Iodine Activities Based on FGR-11 Thyroid Dose Conversion Factors Adjusted Isotope pCi/gm 1-131 0.7727 1-132 0.2813 I-133 1.2363 1-134 0.1793 1-135 0.6800 Use of FGR-l 1 Table 2.1 effective dose conversion factors results in the following iodine activities (after adjusting the initial primary coolant iodine activities to 1.0 VCi/gm D.E. I-131):

Primary Coolant Iodine Activities Based on FGR-11 Effective Dose Conversion Factors Adjusted Isotope gLCi/gm I-131 0.7562 1-132 0.2753 I-133 1.2099 I-134 0.1754 1-135 0.6655 As can be seen by examining the two tables, using the thyroid dose conversion factors in the definition of D.E. 1-131 results in higher iodine concentrations in the primary coolant. Thus, using the thyroid dose conversion factors produces a more conservative determination of the primary coolant iodine activity for use in the dose calculations.

In order for Technical Specifications to be consistent with the conservative analytical basis, the definition of dose equivalent I-13 1 (D.E. I-13 1) was established based on the thyroid dose conversion factors. This approach is consistent with other previously approved Alternative Source Term submittals. For example, Page 22 of the Safety Evaluation for Shearon Harris Nuclear Power Plant Unit 1 Amendment No. 107 to Facility Operating License No. NPF-63 dated October 12, 2001 (ADAMS Accession No. ML012830516) states:

"Revise TS 1.11 definition of dose equivalent iodine-131 to read in part, The thyroid dose conversion factors used for this calculation shall be those listed in the International Commission on Radiological Protection (ICRP), "Limits for Intakes of Radionuclides by Workers," ICRP Publication 30, Volume 3, No. 1-4, 1979 (or equivalently, Federal Guidance Report No. 11, "Limiting Values of Radionuclides Intake and Air Concentration and dose Conversion Factors for Inhalation, Submersion, and Ingestion," EPA 520/1-88-020, September 1988)."

RAI 2

For the gaseous and waste system failure events, Seabrook proposes to use the current licensing basis criterion of a "small fraction of the guidelines." The staff did not address these two events in Regulation Guide 1.183 since these events are not likely to result in core damage. Therefore, no AST specific dose criteria were provided. Nonetheless, the staff notes that the Standard Review Plan Sections 15.7.1 and 15.7.2 and 15.7.4 impose acceptance criteria from Branch Technical Position 11-5. These in turn derived from 10CFR Part 20 rather than Part 100. The staff's original Safety Evaluation Report (SER) does not appear to address the radiological consequences of these events. Please briefly describe the basis of the Seabrook Offsite Dose Calculation Manual (ODCM) controls that limit the contents of these tanks.

Please explain any significant differences between these basis and the acceptance criteria you are proposing in the License Amendment Request (LAR).

based on a source term of noble gas inventory from five carbon delay beds and of a activity inventory of 1% failed fuel. The original analysis also assumes 100% of the noble gas would be released to the environment over two hours. The Alternative Source Term analysis also assumes the maximum possible source term is in the tank and is available for release. The contents of the Radioactive Gaseous Waste System are controlled by plant operations. There are no specific restrictions for discharge to the Radioactive Gaseous Waste System.

The Seabrook Station ODCM ensures that the dose, at any time, at and beyond the site boundary from effluent will be within the annual dose limits of 10 CFR 20 to unrestricted areas. These releases do not include accident conditions; therefore, these limits apply to normal plant operation and its effluents and not for accident conditions.

RAI 3

With regard to control room emergency ventilation actuation, Seabrook has assumed a 30 second delay in actuation for all analyzed accidents. In paragraph 1.6.3, Seabrook states that this actuation is based on high radiation being detected in the remote air supply piping. On page 20 of 94, it is stated that for the Loss of Coolant Accident (LOCA), a contaimnent high pressure signal actuates isolation and that the 30 seconds are provided for diesel generator start time and damper actuation and positioning time. Please explain how the assumed 30 second delay is conservative for all accidents, considering the response considerations identified by FPL, but also how the time for the input activity to ramp up to the alarm set point level and the impact of differences in accident specific radionuclide effluent mixes on monitor response are considered.

FPLE Response to RAI 3:

The current licensing basis for control room habitability is described in Seabrook Station UFSAR Section 6.4. and the emergency mode of operation is described in UFSAR section 6.4.3.2.

The control room ventilation system will be isolated if high radiation is sensed at either intake, a safety injection signal (S signal) occurs, or the system is manually isolated. Once a control room ventilation isolation signal is generated, the following occurs:

• The normal operating intake fans stop and their discharge dampers close

• The emergency intake and cleanup filter fans start and their dampers open.

The response time for the control room to switch over to recirculation/filtration is 5 seconds (after receipt of a high radiation signal, or S signal). The radiation monitors are located directly in the entrances of the control room remote air intakes. The distance from the location of the radiation monitors to the ventilation isolation dampers is quite long. Based on the air intake flow rate and based on the dimensions of the intake, the minimum transit time from a radiation monitor to an isolation damper is 8.3 seconds. This transit time is longer than the response time for the isolation of the control room. Therefore, the control room will be isolated in 5 seconds, and the isolation system will prevent any non-filtered air from entering the Control Room.

The response times for the Engineered Safety Features Actuation System (ESFAS) are provided in Technical Requirements Manual, TR 2. The response times are summarized as follows:

Event CBS Emergency Fan/Filter Actuation Time Control Room Intake Hi Radiation Containment Pressure - Hi-I (S-signal)

Pressurizer Pressure - Low (S-signal)

< 5 seconds

< 5 seconds

< 5 seconds For the Alternative Source Term analysis, a bounding and conservative isolation time of 30 seconds was used in order to provide design margin.

As per Section 12.3 and Table 12.3-16 of the Seabrook Station UFSAR, the air intake radiation monitors are GM-type detectors that are located directly in the duct air stream.

These detectors do not depend on air pumps to provide an air sample for analysis, nor are they shielded. The setpoint for these monitors is 2 times background (2 x 0.5 = 1 mR/hr) or 100 cpm; thus, for any dose significant event, it can be assumed that these monitors will initiate a control room isolation signal. These detectors have a low sensitivity and will respond to low levels of radionuclide effluent mixes.

Thus, there is no delay or ramp up time for the exposure of the radiation monitors to the release.

The assumption that a high radiation signal will be generated due to the low setpoint of the radiation monitors can be confirmed by examining the average whole body dose rates due to noble gas at the entrance to the most limiting control room air intake for the first 30 seconds of the Seabrook Station events.

The RADTRAD-NAI dose calculation model assumes instantaneous transport of releases from the release point to the receptor point; thus, there is no delay or ramp up time for the exposure of the radiation monitors to the release:

Whole Body Dose Rate at Entrance to Most Limiting Air Intake Due to Noble Event Gas (average over first 30 seconds) mremlhr MSLB 1.27 SGTR 264.83 Locked Rotor 260.93 Letdown Line Break 95.15 Radioactive Gaseous 18471.16 Waste System Failure 18471.16

RAI 4a:

Regarding the control room unfiltered inleakage assumptions:

4.a For those events in which the 20 cfin door leakage is not assigned to a particular infiltration point, is the value included in the inleakage values shown in Table 1.6.3-1?

FPLE Response to RAI 4a:

Yes, the values listed under "Unfiltered Inleakage (Total)" in Table 1.6.3-1 are the total assumed unfiltered inleakage including the 20 cfm door leakage.

RAI 4b:

Regarding the control room unfiltered inleakage assumptions:

3.4.b In its Generic Letter 2003-01 response, Seabrook reported preliminary results for the Seabrook inleakage testing. Please confirm that the final test results are bounded by the minimum inleakage assumption shown in Table 1.6.3-1.

FPLE Response to RAI 4b:

The final tracer gas test results are as follows:

Mode Tested Unfiltered In-leakage Train A 8 + 11 SCFM Train B 14 + 22 SCFM Based on the above, the final test results are bounded by the minimum in-leakage assumptions shown in Table 1.6.3-1 of Enclosure 2 of NYN-03061.

RAI 5

Regarding Table 1.8.2-1, the staff is of the opinion that only the 0-2 hour exclusion area boundary (EAB) X/Q value has applicability to the radiological consequence calculations that determine the worst two hour EAB dose. If the values for time periods beyond two hours were used in the analysis of the worst two hour dose, please explain how the values were used and why this approach should be considered acceptable.

delayed period of more severe release, the LPZ X/Q's are not specifically manipulated to apply the 2-hr dispersion factor to that period of the most severe release even though doing so might produce a more conservative value for the LPZ dose.

It is assumed that the conservatisms inherent in the probabilistic determination of the LPZ's time dependent X/Q's adequately address the potential for such a "delayed release" scenario. The same observations can be made concerning the X/Q's used in the control room dose assessment.

The EAB is subjected to the same time-dependent release as was assumed for the LPZ dose assessment. Both the EAB and the LPZ dose assessments use dispersion factors based upon a statistical treatment of the site-specific meteorological data applied to the site specific EAB and LPZ geometries.

Therefore it was deemed appropriate to apply the same method to establish the time dependent X/Q's for the EAB dose as was specified for the LPZ dose.

Although it might be even more conservative to ensure that the 0-2 hour X/Q is applied during the most severe 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> release period, that practice was not required for the LPZ and would introduce an apparently inconsistent treatment of the scenario as analyzed for the EAB and LPZ.

Justification for this practice is also provided by precedent in previous Safety Evaluation Reports for AST submittals such as Table 1 of the Safety Evaluation for Duane Arnold Energy Center Amendment No. 240 to Facility Operating License No. DPR-49 dated July 31, 2001(ADAMS Accession No. ML011660142).

For most postulated Seabrook Station events, the most severe period of the release occurs within the first two hours of the event and therefore the X/Q values assigned for the duration of the event are insignificant to the calculation of the 0-2 hour EAB dose.

In addition, Summary of Results Table 3-1 of Enclosure 2 to NYN-03061demonstrates that the EAB doses for the Seabrook Station are not generally severe. Even if the most severe (0-2 hr) X/Q for the EAB was used throughout the event, it would not challenge the acceptance criteria.

Four postulated events for Seabrook Station included EAB dose contributions that were not more severe during the first two hours of the event; however, each of these events included multiple EAB dose contributions that were not maximized during the same two-hour period.

For example, the two-hour maximum EAB dose contribution of the various LOCA dose components (i.e., ECCS leakage, containment leakage, containment purge, RWST backleakage and direct shine) all occur over different two-hour time periods.

It was not believed to be reasonable to assume that the worst-case two-hour EAB atmospheric dispersion factor would statistically exist at four different times during the first day of the LOCA event.

Also, summing dose components occurring at four different two-hour time periods (e.g., 0 - 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, 0.5 - 2.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />, 1.3 - 3.3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />, etc.) is not considered reasonable.

Based on feedback from the NRC Staff during a telephone call on April 20, 2004, the EAB worst two-hour doses for the four postulated events (that included dose components that were not most severe during the first two hours of the event) were re-evaluated with the conservatism of applying the limiting two-hour EAB atmospheric dispersion factor over a time period of up to a full day. The worst two-hour dose contributions were then summed regardless of when they occurred during the event. The table below provides the results of the new analyses and demonstrates that all four events are still well within the EAB dose acceptance criteria.

Worst 2-hour EAB Worst 2-hour EAB Dose Acceptance Dose (a) applying the worst case two-Criteria hour X/Q (b)

LOCA 4.40 rem TEDE 4.63 rem TEDE 25 rem TEDE RCCA Ejection 6.3 rem Secondary Side 1.52 rem TEDE 1.85 rem TEDE TEDE Release MSLB Concurrent 0.20 rem TEDE 0.29 rem TEDE 2.5 rem Iodine Spike TEDE Locked Rotor 0.56 rem TEDE 0.73 rem TEDE 2.5 rem Notes:

(a) Two hour EAB dose from Table 3-1 of Enclosure 2 to NYN-03061 (b) Two hour EAB dose applying the worst case two hour X/Q for the entire event and summing the worst two-hour dose contributions from different time periods (c) Note that utilizing the limiting two-hour EAB atmospheric dispersion factor for up to a full day and combining multiple EAB two-hour dose components for an event that do not occur at the same point in the event conservative.

RAI 6a Regarding the LOCA analysis:

3.6.a In paragraph 2.1.2.4, Seabrook states that they are assuming an aerosol deposition rate of 0.1 per hour, based on Industry Degraded Core Rulemaking Program (IDCOR)

Technical Report 11.3. RG 1.183 Regulatory Position 3.4 identifies NUREG/CR-6189 is an acceptable approach. Since this parameter is somewhat dependant on plant parameters, the staffs prior approval of 0.1 per hour for another licensee may not be relevant to Seabrook. Please provide a Seabrook specific justification for your proposed deviation of this guidance.

FPLE Response to RAI 6a:

The IDCOR value and methodology used are applicable to Seabrook Station, and are reasonable and conservative when compared to NUREG/CR-6189. Table 34 of NUREG/CR-6189 presents decontamination coefficients for design basis accident aerosol deposition.

These decontamination coefficients are presented as a function of thermal power, time range and release phase. Table 36 of NUREG/CR-6189 presents correlations to model these decontamination coefficients as a function of thermal power, time range and release phase NUREG/CR-6604 (RADTRAD: A Simplified Model for RADionuclide Transport and Removal And Dose Estimation) Table 2.2.2.1-1 presents correlations for determining the same natural deposition aerosol decontamination coefficients as a function of power, time and

release phase (same as Table 36 of NUREG/CR-6189, but sums the gap and early in-vessel release phases). Thermal power is the only parameter that is varied in this table. The following is an excerpt from page 6 of the Safety Evaluation for Indian Point Nuclear Generating Unit No. 2 Amendment No. 211 to DPR-26 dated July 27, 2000 (ADAMS Accession No. ML003727500):

"Lower bound (10 percentile) natural processes decontamination coefficients for radiological design-basis accidents were identified in Table 34 of NUREG/CR-6189, "A Simplified Model of Aerosol Removal by Natural Processes in Reactor Containments," July 1996. The natural processes aerosol removal model in the staff's confirmatory analysis code RADTRAD is based on NUREG/CR-6189. Based on Table 34, the staff finds the sedimentation removal coefficient of 0.1 hr l to be reasonable."

This conclusion applies to Seabrook Station since the analyzed thermal power level (3659 MWt) is greater than the analyzed Indian Point Unit 2 thermal power level (3216.5 MWt) and the values of decontamination coefficients in Table 2.2.2.1-1 of NUREG-6604 increase with thermal power for these two values.

Several other Safety Evaluation Reports also support the use of 0.1 hr'. The following is an excerpt from page 5 of the Safety Evaluation for Kewaunee Nuclear Power Plant Amendment No. 166 to DPR-43 dated March 17, 2003 (ADAMS Accession No. ML030210062):

"The fission products in the containment atmosphere following the postulated LOCA [are]

mitigated by natural deposition processes and by the containment spray system (CSS). The licensee assumed a radioactive aerosol removal rate of 0.1 per hour in the containment atmosphere. This removal credit is taken after the CSS operation is terminated. The NRC staff finds 0.1 per hour aerosol removal rate to be reasonable (within the 85 percent of the uncertainty distribution) based on [the] study published in NUREG/CR-6189, "A Simplified Model of Aerosol Removal by Natural Processes in Reactor containments," and [it] is therefore acceptable."

The Seabrook Station analyzed thermal power (3659 MWt) is also greater than the Kewaunee Nuclear Power Plant analyzed thermal power (1851.3 MWt). As mentioned previously, the values of the decontamination coefficients in Table 2.2.2.1-1 of NUREG-6604 increase with thermal power for these two values.

RAI 6b:

Regarding paragraphs 2.1.2.11,.12,.15, please confirm the staff's understanding that paragraph 2.1.2.11 and 2.1.2.12 apply to 40 percent La leakage and that the drawdown does not change the 60 percent bypass assumption.

FPLE Response to RAI 6b:

The staffs understanding that paragraph 2.1.2.11 and 2.1.2.12 of Enclosure 2 to NYN-03061 apply to 40% La leakage and that the drawdown does not change the 60 percent bypass assumption is correct.

RAI 6c Regarding paragraph 2.1.2.15, what is the basis of the 40-60 split in containment leakage.

FPLE Response to RAI 6c:

The 60/40 split is based on the Seabrook Station current Licensing Basis and is specified in UFSAR Section 15.B.2.IID as 0.60 La for the Containment Enclosure Emergency Exhaust Filter Bypass Fraction (conservative analysis). The maximum allowable leakage (La) from the containment structure following an accident is 0.15 percent of the mass of its atmosphere per day. This would occur at a maximum pressure. The direct leakage to the environs of radioactive contaminants from the containment is within the guidelines of 10 CFR 100.

Although a containment enclosure emergency cleanup system has been provided to minimize leakage to the environs, a significant number of lines penetrate the containment and terminate in areas not treated by this cleanup system. Therefore, all leakage attributed to penetrations and isolation valves, requiring Type B and Type C Test per 10 CFR 50, Appendix J, is conservatively assumed to bypass the cleanup system, and is limited to 60% of the maximum allowable leakage (La). The remaining 40% of La is assumed to enter the containment enclosure, and to be treated by the containment enclosure emergency cleanup filtration system, prior to release. Seabrook Station tests to the design basis values in accordance with Appendix J.

RAI 6d1 and 6d2:

Regarding paragraph 2.12.19 through 2.1.2.22, the staff cannot find FPL's treatment of emergency core cooling system (ECCS) leakage acceptable without additional supporting justification for the following deviations from guidance:

• Regulatory Position 5.3 states that "with the exception of iodine, all radioactive materials in the recirculating liquid are assumed to be retained in liquid phase."

Seabrook has stated the "with the exception of the non-particulate iodines, all radioactive materials in the recirculating liquid are assumed to be retained in liquid phase."

• Regulatory Position 5.4 and 5.5 provide that the flashing fraction is to be based on the fraction of the total iodine in the liquid. Seabrook proposes that 100% of the non-particulate iodine becomes airborne, but none of the particulate iodine becomes airborne.

• Regulatory Position 5.6 states that the radioiodine available for release is assumed to be 97 percent elemental and 3 percent organic.

Seabrook states that the temperature and pH history of the sump and refueling water storage tank (RWST) are considered in determining the chemical form of iodine.

The staff structured these regulatory positions to be deterministic and conservative in order to compensate for lack of research into iodine speciation beyond the containment and the

The ECCS portion of the Seabrook Station LOCA dose was considered in two parts; the dose due leakage from ECCS components outside of containment and the dose due to leakage back into the RWST.

Leakage from ECCS Components Outside Containment For the Seabrook analysis, it was assumed that a minimum of 1% of the total iodine in the ECCS leakage is elemental (even though the sump pH would preclude elemental iodine). All of the volatile (elemental and organic) iodine was assumed to be released. This resulted in 1.15% of the total iodine in the ECCS leakage being released with a conservative assumption of the volatile forms of 87% elemental and 13% organic. The CEVA filter efficiency is lower for organic iodine; therefore the assumption of 13% organic form results in conservatively higher dose than using the 3% organic form stipulated in Regulatory Position 5.6.

The non-RWST ECCS leakage methodology employed for the Seabrook AST calculations is based on current AST submittals and Safety Evaluation Reports. A review of prior submittals and SERs reveals that there are a variety of interpretations employed for Regulatory Position 5 of Appendix A to Regulatory Guide 1.183, but all rely on pH control reducing the amount of iodine available for release.

For example:

H.B. Robinson AST Submittal:

H.B. Robinson Steam Electric Plant, Unit No. 2 Docket No. 50-261/License No. DPR-23 Request for Technical Specifications Change Regarding Full Implementation of Alternative Radiological Source Term (ADAMS Accession No. ML021420151).

Page 27 of the H.B. Robinson submittal states:

"Regulatory Position 5.3 - With the exception of the non-particulate iodines, all radioactive materials in the recirculating liquid are assumed to be retained in the liquid phase."

Another example:

SER for the Shearon Harris AST Submittal Safety Evaluation by the Office of Nuclear Reactor Regulation Related to Amendment No. 107 to Facility Operating License No. NPF-63 Carolina Power & Light Company Shearon Harris Nuclear Power Plant, Unit 1 Docket No. 50-400 dated October 12, 2001 (ADAMS Accession No. ML012830516).

Page 30 of the Shearon Harris SER states:

"The licensee also assumed 2 percent of all forms of iodine contained in the leakage is released into the auxiliary building atmosphere, consistent with the current design

basis leakage rate assumed in the HNP UFSAR. This assumption is based on the containment sump water pH and the initial sump water temperature."

The Shearon Harris FSAR states that the sump temperature is 230&F.

This temperature will result in a flashing fraction of 2%. Thus, due to sump pH control, Shearon Harris assumed that the amount of iodine released is equal to the flashing fraction.

The approached used in the H.B. Robinson submittal was used for determining the Seabrook ECCS component leakage dose. The minimum sump pH at Seabrook is 8.7. At this pH level there will be essentially no elemental iodine. Since the ECCS leakage during recirculation consists of water drawn directly from the containment sump, it can also be concluded that there will be no elemental iodine present in the leaking ECCS fluid. For the Seabrook analysis, it was conservatively assumed that a minimum of 1% of the total iodine in the ECCS leakage is elemental (even though the sump pH would preclude elemental iodine). All of the volatile (elemental and organic) iodine was assumed to be released. This resulted in 1.15% of the total iodine in the ECCS leakage being released (87% elemental and 13% organic). The dose due to this release is 8.8E-3 rem for the control room.

The initial flashing fraction for Seabrook is 4.7%, dropping to 0% flashing at 3.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />. If the additional conservatism in the Shearon Harris method is followed, (iodine release equal to the flashing fraction with all volatile iodine being released when there is no flashing, with a released iodine composition of 97% elemental and 3% organic), the control room dose is 9.5E-3 rem. This results is no change in the reported control room dose (4.73 rem total dose).

Although the total iodine release is conservatively higher using the Harris method (0.047 versus 0.0115), the method used for Seabrook has a higher fraction of organic iodine. The Containment Enclosure Emergency Air Cleanup System filter efficiency is lower for organic iodine; therefore, the doses produced by the two methods are approximately equal.

Another example:

SER for the Kewaunee AST Submittal Safety Evaluation by the Office of Nuclear Reactor Regulation Related to Amendment No. 166 to Facility Operating License No. DPR-43 Nuclear Management Company, LLC Kewaunee Nuclear Power Plant Docket No. 50-305 dated March 17, 2003 (ADAMS Accession No. ML030210062).

Page 6 of the Kewaunee SER states:

"The iodine partition factor of I percent is a departure from the guidance provided in RG 1.183 which states that an iodine partition factor of 10 percent should be assumed unless a smaller amount can be justified based on the actual sump water pH history and area ventilation rate. The Kewaunee Nuclear Power Plant (KNPP) USAR Section 6.2.5, "Effects of Leakage from Residual Heat Removal System," identifies that the temperature of the containment recirculation water is below 212 TF when ECCS recirculation begins and that any leakage liquid is therefore assumed to be subcooled and to remain liquid retaining iodine contained in any leakage water in the liquid.

The licensee stated that 1 percent iodine partition factor and 50 percent iodine plateout assumptions were contained in the original KNPP USAR Section 14.3.7 (original page 14.3-68) which were approved by the Atomic Energy Commission. Therefore, the licensee considers that these assumptions are parts of the Kewaunee original license and that they are still current licensing bases. Furthermore, the NRC staff believes that this assumption is unrelated to the use of AST. Therefore, the NRC staff finds that 1 percent iodine partition factor and 50 percent iodine plateout assumptions are acceptable."

While the Kewaunee SER points to prior licensing bases, it recognizes that containment sump temperature and pH may result in iodine partition factors that are lower than 10 for Engineered Safety Features (ESF) leakage. It should also be noted that (conservatively) no plateout in the auxiliary building was credited for Seabrook Station ESF leakage.

Based on feedback from the NRC staff during the telephone call on April 20, 2004, an additional ECCS leakage analysis was performed to determine the effect of a conservatively assumed 10% iodine release on the Seabrook LOCA dose results. The following changes were made to the submitted ECCS leakage case:

10% of the total iodine in the leaked ECCS fluid was assumed to be released.

The makeup of the released iodine was assumed to be 97% elemental and 3% organic.

The conservative assumption that 10% of the total iodine in the ECCS leaked fluid is released results in a Control Room dose of 4.74 rem (versus 4.73 rem in the submitted analysis).

Leakage to RWST The RWST pH at the time of recirculation is 7.1. Since the sump water leaking into the RWST is of a higher pH, it can be concluded that elemental iodine generation will also be precluded in the RWST.

However, for the purposes of the Seabrook analysis it was conservatively assumed that 1% of the total iodine in the RWST was elemental. Assuming 1% elemental iodine in the RWST with a high pH is conservative relative to the methodology documented in the Shearon Harris AST submittal. Consistent with the Shearon Harris AST submittal, it was also assumed that for the RWST backleakage determination, all of the sump iodine is of non-volatile form (i.e., particulate).

Iodine speciation is not expected to change as a result of ECCS backleakage into the RWST.

The RWST water pH is above 7, which would preclude re-generation of elemental iodine via radiolytic conversion. The high pressure ECCS flowpath to the RWST consists of backleakage across check valves in the water-filled RWST outlet piping. The leakage enters the RWST underneath the water level in the RWST. Therefore, airborne particulate release is not considered credible. For the Design Basis LOCA with ECCS leakage and emergency recirculation functioning, no additional water would be added to the RWST based on current Seabrook Station operating procedures and licensing basis. This precludes any addition of water to the RWST from sources other than the ECCS backleakage that might have the potential to impact the pH. The dose in the RWST is also lower than the dose in containment.

Most of the data on radiolytic conversion is based on containment dose rates. Despite the fact that the sump and RWST pH conditions are expected to preclude any radiolytic conversion,

the calculation conservatively assumes that 1% of the iodine is available for release in the elemental form.

RAI 6d3 A discussion of the impact of all possible post accident liquid inputs to the RWST with other sources of water.

FPLE Response to RAI 6d3:

For the Design Basis LOCA with ECCS leakage and emergency recirculation functioning, no additional water would be added to the RWST based on the Seabrook Station current Licensing Basis and operating procedures.

RAI 6d4:

A discussion on how the iodine speciation might change as the ECCS leakage is sprayed out of a leak, or streams across a floor into a building sump.

FPLE Response to RAI 6d4:

See FPLE response to RAIs 6d1 and 6d2.

RAI 6el, 6e2, and 6e3:

On page 20 of 94, the basis for the air flow rate is provided. Please address the following:

RAI 6el The air flow is based on the average daily temperature swing of 18.2 degrees.

This temperature swing appears low for a summer day.

Please explain how this value was determined and why it should be considered adequately conservative.

RAI 6e2 Was evaporation of the RWST water considered as a contributor to the air flow rate?

RAI 6e3 Since the iodine partition is the ratio of the vapor pressure of the iodine in the liquid and gas phases in the RWST, please discuss the impact of tank pressure changes associated with the diurnal temperature swings.

FPLE Response to RAI 6el, 6e2. and 6e3 Response As presented on page 14 of Enclosure 2 to NYN-03061, the average daily temperature swing used in the RWST releases is determined from the 2001 ASHRAE fundamentals handbook.

The value of 18.2 TF for the dry bulb temperature swing from Portsmouth NH was used as the

As demonstrated by the discussion above, the complexity and uncertainties in the analysis of the dose contribution from the RWST have been addressed by imposing some very conservative simplifying assumptions on aspects of the analysis that are difficult to quantitatively assess. As discussed in other responses to questions concerning the iodine species in the sump and RWST water, this very conservative treatment of uncertain conditions is also inherent in the assumptions and modeling used in those aspects of the analysis. The imposition of these multiple combined conservatisms provide confidence that the overall assessment of the dose from this aspect of the event is conservative.

RAI 6e4 As noted above in question 6d, the staff questions the iodine fraction value.

FPLE ResDonse to RAI 6e4:

See FPLE response to RAIs 6dl and 6d2.

RAI 6f On page 21 of 94, a mixing rate of two turnovers per hour is assumed. Regulatory position 3.3 provides this as a default assumption, if adequate flow exists between these two regions.

Please briefly describe the basis for assuming that this flow will exist between the sprayed and unsprayed regions.

FPLE Response to RAI 6f Per Seabrook Station UFSAR Section 1 5.6.5.4.a, "Radiological Consequences -

Spray Removal Analysis", The mixing rate is 2 hrf' from the unsprayed region. The UFSAR states the following:

"The effectiveness of the Containment Spray System has been evaluated using a two region spray model which assumes that 85.4 percent of the containment volume is directly wetted by the spray solution. Mixing between the sprayed and unsprayed region is assumed to occur at a rate of 13,000 cfin, corresponding to a mixing rate of is 2 hr-' from the unsprayed region."

The two air changes an hour are attributable strictly to natural convection.

RAI 62

On page 21 of 94, the maximum decontamination factor (DF) for elemental and particulate iodines are discussed. Please explain how the initial maximum airborne iodine concentration in the containment was determined for this determining DF.

The second RADTRAD-NAI case determined the time required to reach a DF of 200 based on the peak iodine concentration from the first RADTRAD-NAI case.

The second RADTRAD-NAI case included:

Containment sprays actuated at 0.018 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br />

• No surface deposition

• No decay Based on this method, a DF of 200 is reached at 3.09 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br />.

Thus, the time to reach a DF of 200 used for the Seabrook AST submittal is 10 minutes shorter than allowed by a method that does not include sprays in the determination of the peak containment atmosphere elemental iodine concentration and follows the SRP definitions.

Therefore, the method used in the Seabrook AST submittal conservatively determines a shorter length of time during which the sprays are effective than the time determined using an alternate method that establishes the peak containment atmosphere elemental iodine concentration without sprays and based on definitions in the SRP.

RAI 6h:

Table 2.1-1 identifies the containment enclosure drawdown time for the LOCA as 4.5 minutes (270 seconds). Table 2.6-identifies the draw down time for the rod control cluster assembly (RCCA) ejection accident as 360 seconds. Appendix A of the Seabrook Updated Final Safety Analysis Report (UFSAR) states that filtration credit is not assumed for the first eight minutes. Please explain the difference in these values. What is the value of the acceptance criteria for surveillance testing of this system safety function?

FPLE Response to RAI 6h:

As described in Seabrook Station UFSAR Section 6.2.3.5, the Containment Enclosure Emergency Air Cleanup System is automatically initiated on a 'T' (containment isolation Phase A) signal. The required response time for the Containment Enclosure Emergency Air Cleanup System is provided in Technical Specification 3/4.6.5. The required response time is summarized as follows:

Event CBS Emergency Fan/Filter Actuation (Drawdown) Time Containment Isolation Phase 'A' (T-signal)

< 4 minutes For the Alternative Source Term analysis, a bounding and conservative drawdown time of 6 minutes ( 360 seconds) was used in order to provide design margin for the RCCA ejection accident.

The value of the acceptance criteria for surveillance testing of this system is to evacuate to -

0.25 in H20 in less than four minutes. This acceptance criteria is only for drawdown and does not include the signal time.

The value of 8 minutes in Appendix B of the UFSAR is in error and will be corrected through the Seabrook Station Corrective Action Program.

RAI 6i Section 2.1.2.13 addressed Regulatory Position 4.3 and states that the containment enclosure emergency air cleaning system is capable of maintaining a negative pressure with respect to high wind speeds. UFSAR sections 6.5.1.1 and 6.5.1.3 are cited. UFSAR Section 6.5.1.3 states, "The calculated wind speed that would initiate building exfiltration is 17 miles per hour. At this or at a higher wind speed, any exfiltration will be adequately dispersed." Please explain the basis of this conclusion. What is meant by adequately dispersed? What is the 95 percentile wind speed at Seabrook? What impact does this windspeed have on the time to reach 0.25 inch water gage (WG).

FPLE Response to RAI 6i:

The issue of "adequate dispersion" is not relevant or appropriate with regard to the Seabrook Station Alternative Source Term radiological dose analyses and associated X/Q atmospheric dispersion factors. Point-to-point X/Q atmospheric dispersion factors were determined with ARCON96 from the closest point on containment to the Control Room intake/inleakage points. For both the LOCA and RCCA Ejection events, these limiting point-to-point X/Q atmospheric dispersion factors were applied to all containment leakage prior to adequate drawdown by the Containment Enclosure Emergency Air Cleanup System and applied to the containment enclosure ventilation area bypass flow for the entire event. After the containment enclosure is adequately drawn down by the Containment Enclosure Emergency Air Cleanup System, the limiting point-to-point X/Q atmospheric dispersion factors from the unit plant vent to the Control Room intake/inleakage points are used for the containment leakage that is processed and filtered by the Containment Enclosure Emergency Air Cleanup System (i.e.,

the leakage that does not bypass the containment enclosure ventilation area). No diffuse source X/Q atmospheric dispersion factors are utilized for the Seabrook Station Alternative Source Term analyses.

The statement in Section 6.5.1.3 of the Seabrook Station UFSAR. will be revised concurrent with other Seabrook Station UFSAR sections impacted by the implementation of Alternative Source Term methods.

RAI 6i:

The Seabrook UFSAR provides an analysis of the consequences of post accident venting as a backup to the redundant hydrogen recombiners.

This analysis was not address in the submittal. Is it Seabrook's intent to remove this analysis from the licensing basis? If not, why was this component of LOCA not address in the license amendment request?

FPLE Response RAI 6j:

The Combustible Gas Control System meets the redundancy and power source requirements for engineered safety features. No single failure will incapacitate the system. The Alternative Source Term Analysis credited the hydrogen recombiners for hydrogen control and considered post accident venting of hydrogen as a backup to the redundant hydrogen recombiners, and as such, beyond the design basis accident.

Seabrook Station is currently pursuing a License Amendment Request to eliminate hydrogen control from the Technical Specifications and the post accident venting will be addressed in this effort. The proposed Technical Specification changes support implementation of the revisions to 10CFR 50.44, "Standards for Combustible Gas Control Systems in Light-Water-Cooled Power Reactors,"

that became effective on October 16, 2003.

RAI 7

Regarding the main steam line break analysis, Table 2.3-1 lists an RCS mass of 539,037 Ibm.

Table 2.3-4 lists an RCS mass of 505,000 Ibm. Table 2.6-1 lists a minimum RCS mass of 434,000 and a maximum mass of 539,037 Ibm. While the staff understands minimum and maximum values may be used to maximize doses, it is not clear why the RCS mass assumed in establishing the iodine appearance rate was assumed to be 505,000 Ibm. Please explain the basis for this assumption.

FPLE Response to RAI 7 This RAI response applies to RAIs 7, 8a, and 10 for Main Steam Line Break, Steam Generator Tube Rupture, and Letdown Line Break, respectively.

The dose calculations for the AST analyses used the maximum RCS mass for calculations where the entire activity of the RCS was released or the release contribution from the initial RCS activity was being computed.

The minimum RCS mass was used for cases where fuel failure activity was released into the RCS and mixing with the inventory in the pressurizer could not be assured.

For the concurrent Iodine spiking cases, an RCS mass of 505,000 lbm was originally used to establish the RCS initial equilibrium iodine concentrations documented in the Seabrook Station UFSAR which are subsequently used in this analysis to determine the appearance rate.

To be consistent with that basis, the appearance rate calculation (source term) uses the same RCS mass as was assumed to establish the initial concentrations. To be consistent with the source term, the RADTRAD model uses the same RCS mass as was used to generate the source term.

The 1.0 jiCi/gm dose equivalent 1-131 equilibrium iodine activities were established by scaling the relative equilibrium iodine isotopic concentrations in the coolant presented in Table 11.1.1 of the Seabrook Station UFSAR to achieve the Technical Specification limit of 1.0 ptCi/gm dose equivalent I-131. The value of 505,000 lbm obtained from UFSAR Table 11.1-3 was determined to be an appropriate RCS mass value for the concurrent spiking calculations.

RAI 8a:

With regard to steam generator tube rupture analysis:

Regarding the steam generator tube rupture analysis, Table 2.4-1 lists an RCS mass of 539,037 Ibm. Table 2.4-4 lists an RCS mass of 505,000 Ibm. Table 2.6-1 lists a minimum RCS mass of 434,000 and a maximum mass of 539,037 Ibm. While the staff understands minimum and maximum values may be used to maximize doses, it is not clear why the RCS mass assumed in establishing the iodine appearance rate was assumed to be 505,000 Ibm.

Please explain the basis for this assumption.

FPLE Response to RAI 8a:

See FPLE Response to RAI 7 above.

RAI 8b In paragraph 2.4.2.12, please clarify the phrase, "...without flashing for all steam generators..." as used in the first bullet. The use of "all" appears to be in conflict with the second bullet.

FPLE Response to RAI 8b:

In Section 2.4.2.12, the correct phrasing for the first bullet is "... without flashing for all intact steam generators...".

RAI 8c The Table 1.8.1-3 entry for steam generator tube rupture (SGTR) uses language different from that for the main steam line break (MSLB), locked rotor, or RCCA ejection events. It appears that this difference in language provides for the factor of five plume rise reduction to be applied to noble gas releases for the entire eight hour release duration rather than 2.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />. If this is Seabrook's intent, please provide a justification for this assumption.

FPLE Response to RAI 8c:

The phrasing used to describe plume rise credit in Table 1.8.1-3 is confusing. When credit for plume rise is taken, it is only applied for the first 2.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> of the event. The descriptions used in Table 1.8.1-3 could be interpreted to imply that the Table 1.8.1-2 X/Qs include the factor of 5 reduction for plume rise; however, the X/Qs listed in Table 1.8.1-2 do not include a factor of 5 reduction for plume rise. For the steam generator tube rupture and main steam line break events, the factor of 5 reduction for plume rise was not applied to the noble gas release. In addition, for the main steam line break event, all of the noble gas was assumed to exit from the steam line break location.

RAI 9a With regard to the RCCA ejection analysis:

Please respond to Questions 6a through c and 6h in the context of the RCCA ejection event.

FPLE Response to RAI 9a:

See FPLE Response to RAIs 6a, 6b, 6c and 6h.

RAI 9b Please confirm the staffs understanding that the 0.375 percent full centerline melt is referenced to the entire core and not only that fraction of the core that exceeds departure from nucleate boiling (DNB)

FPLE Response to RAI 9b:

The NRC staff's understanding is correct. The 0.375 percent fuel centerline melt was applied to the entire core.

RAI 10

Regarding the letdown line break analysis, Table 2.7-4 lists an RCS mass of 505,000 Ibm.

Table 2.7-1 lists a minimum RCS mass of 434,000 lbm and a maximum mass of 539,037 lbm.

While the staff understands minimum and maximum values may be used to maximize doses, it is not clear why the RCS mass assumed in establishing the iodine appearance rate was assumed to be 505,000 Ibm. Please explain the basis for this assumption.

FPLE Response to RAT 10:

See FPLE Response to RAI 7 above.

RAI 11

Table 2.9-1 refers to non-existent Tables 2.10-2 and 2.10-3.

Please confirm the staffs understanding that Table 2.9-2 is the intended reference.

FPLE Response to RAI 11:

The NRC staff's understanding is correct. Table 2.9-2 is the intended reference for the RLWS release inventory in Table 2.9-1.

Additionally, Table 2.8-2 (versus Table 2.9-2) is the intended reference for the RGWS release inventory in Table 2.8-1.