Text

Final Safety Analysis Report (FSAR) - Chapter 15.5 Fuel Handling Accident (FHA) Dose Analysis
	ML11269A064
Person / Time
Site:	Watts Bar
Issue date:	09/23/2011
From:	Stinson D; Tennessee Valley Authority
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	Download: ML11269A064 (127)
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Tennessee Valley Authority, Post Office Box 2000, Spring City, Tennessee 37381-2000 September 23, 2011 10 CFR 50.34(b) 10 CFR 50.67 10 CFR 100 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555-0001 Watts Bar Nuclear Plant, Unit 2 NRC Docket No. 50-391

Subject:

Watts Bar Nuclear Plant (WBN) Unit 2 - Final Safety Analysis Report (FSAR) - Chapter 15.5 Fuel Handling Accident (FHA) Dose Analysis

References:

1. TVA letter to NRC dated June 27, 2011, "Watts Bar Nuclear Plant (WBN) Unit 2 - Final Safety Analysis Report (FSAR) - Response to Request for Additional Information (RAI) Regarding Accident Dose Analysis Basis"

2. TVA letter to NRC dated August 5, 2011, "Watts Bar Nuclear Plant (WBN) Unit 2 - Final Safety Analysis Report (FSAR) - Chapter 15.5 Design Basis Dose Analysis"

3. TVA letter to NRC dated September 15, 2011, "Watts Bar Nuclear Plant (WBN) Unit 2 - Final Safety Analysis Report (FSAR) - Chapter 15.5 Design Basis Dose Analysis" This letter provides revised FSAR Design Basis Accident (DBA) dose analysis results for the FHA. provides a discussion of changes to the FHA analysis currently described in the FSAR and a revised FSAR Section 15.5.6 and associated tables updated to reflect the new results. The changes in the dose results for this DBA were the result of the following three items:

U.S. Nuclear Regulatory Commission Page 2 September 23, 2011

1. Changes in assumptions on the closure time for Auxiliary Building and Main Control Room Dampers in the normal ventilation system.

2. Alternate source term being used as the basis for the dose calculations for the FHA in the Auxiliary Building and for the FHA in containment when the equipment hatch is open.

3. The incorporation of meteorology data for the 20 year period of 1991 to 2010 as opposed to the 1976 to 1993 data used for licensing Unit 1. provides a complete draft of FSAR Section 15.5 red-line showing the recent revisions to the analyses as provided in this letter and References 2 and 3. Enclosure 3 provides a clean copy of the draft FSAR section.

The submittal of the results of this analysis was a TVA commitment to NRC described in Reference 1. This letter closes Commitment 2 of Reference 2 to provide the FHA results and the commitment in Reference 3 to provide the FHA and the proposed FSAR Section 15.5. There are no new regulatory commitments in this letter.

I declare under penalty of perjury that the foregoing is true and correct. Executed on the 2 3 rd day of September, 2011.

Respectfully, David Stinson Watts Bar Unit 2 Vice President

Enclosures:

1. WBN Unit 2 Revised FSAR Section 15.5 Fuel Handling Accident Dose Analysis Results

2. WBN Unit 2- Revised FSAR Section 15.5. - Red-Lined

3. WBN Unit 2- Revised FSAR Section 15.5

U.S. Nuclear Regulatory Commission Page 3 September 23, 2011 cc (Enclosures):

U. S. Nuclear Regulatory Commission Region II Marquis One Tower 245 Peachtree Center Ave., NE Suite 1200 Atlanta, Georgia 30303-1257 NRC Resident Inspector Unit 2 Watts Bar Nuclear Plant 1260 Nuclear Plant Road Spring City, Tennessee 37381

Enclosure I WBN Unit 2 Revised FSAR Section 15.5 Dose Analysis The Fuel Handling Accident (FHA) was revised to account for a change in the meteorology data used and due to changes to Main Control Room and Auxiliary HVAC damper closure times.

The meteorology was updated to use the 20 year period of 1991 to 2010. The analysis for a dropped fuel assembly inside containment when the containment air locks and equipment hatch are closed continues to use the methodology of Regulatory Guide (RG) 1.25. The purge system is in operation and credit is taken for the HEPA and charcoal beds prior to purge system isolation. The purge system will automatically isolate on high radiation from radiation monitors located in the purge system exhaust. No credit is taken for this isolation in the analysis and all activity is released to the environment within two hours as specified in RG-1.25.

Alternate source term (AST) described in RG-1. 183 was selectively used to evaluate the FHA due to an event in the spent fuel pool located in the Auxiliary Building or in the containment when the equipment hatch or both doors in a personnel air lock are open. As part of this selective implementation of AST, the following changes are assumed in the analysis:

" The total effective dose equivalent (TEDE) acceptance criterion of 10 CFR 50.67(b)(2) replaces the previous whole body and thyroid dose guidelines of 10 CFR 100.11.

The gap activity is revised to be consistent with that required by RG-1.183.

The decontamination factors were changed to be consistent with those required by RG-1.183.

New onsite (control room) and offsite atmospheric dispersion factors (X/Q) are used.

The time to isolate the control room is increased from 20.6 seconds to 40 seconds.

No Auxiliary Building isolation is assumed.

No filtration of the release from the Containment or the spent fuel pool to the environment by the containment purge filters or the Auxiliary Building Gas Treatment System (ABGTS) is assumed.

The evaluation for the FHA at the spent fuel pool is a bounding analysis for a dropped assembly in containment when the containment is open. The release point for the containment purge system is the Unit 2 shield building stack. The X/Qs are lower for this release point than the normal Auxiliary Building exhaust. In addition, any release from the shield building stack would go through the purge system HEPA and charcoal filter assemblies prior to release. Currently, when the purge lines isolate on high radiation, the Auxiliary Building also isolates, and ABGTS is actuated. The release point for ABGTS is the shield building stacks, and the releases are filtered through HEPA and charcoal assemblies. Thus, the AST analysis for the FHA in the Auxiliary Building that considers no filtration and no Auxiliary Building isolation is conservative and acceptable as the basis for the containment open evaluation.

The following pages of Enclosure 1 provide the revised FSAR Section 15.5.6 on the FHA. The section has been divided into three subsections. The first provides the assumptions for the RG-1.25 closed containment analysis. The second subsection, 15.5.6.2, provides the new assumptions for the AST. This is new information. Section 3 provides a summary of the results. There were only minor editorial changes in the first five paragraphs. The AST results summary starts at paragraph six. Table 15.5-20 was modified to reflect the closed containment analysis. A new Table 15.5-20.a was added for the AST analysis. The results are shown in Table 15.5-23, and the changes are marked.

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Enclosure I WBN Unit 2 Revised FSAR Section 15.5 Dose Analysis 15.5.6 Environmental Consequences of a Postulated Fuel Handling Accident The analysis of the fuel handling accident considers three cases. The first case is for a Fuel Handling Accident inside containment with the containment closed and the Reactor Building Purge System operating. This analysis is discussed in Section 15.5.6.1 and is based on Regulatory Guide 1.25[111 and NUREG 5009[24]. The second case is for an accident in the spent fuel pool area located in the Auxiliary Building. This case is discussed in Section 15.5.6.2 and is evaluated using the Alternate Source Term based on Regulatory Guide 1.183[18], "Alternate Source Terms." The third case considered is an open containment case for an accident inside containment where there is open communication between the containment and the Auxiliary Building. This evaluation is discussed in Section 15.5.6.2 and is based on Regulatory Guide 1.183.

15.5.6.1 Fuel Handling Accident Based on Regulatory Guide 1.25 The parameters used for this analysis are listed in Table 15.5-20.

The bases for the Regulatory Guide 1.25 evaluation are:

(1) In the Regulatory Guide 1.25 analysis, the accident occurs 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months after plant shutdown. Radioactive decay of the fission product inventory during the interval between shutdown and placement of the first spent fuel assembly into the spent fuel pit is taken into account.

(2) In the Regulatory Guide 1.25 analysis damage is assumed for all rods in one assembly.

(3) The assembly damaged is the highest powered assembly in the core region to be discharged. The values for individual fission product inventories in the damaged assembly are calculated assuming full-power operation at the end of core life immediately preceding shutdown. Nuclear core characteristics used in the analysis are given in Table 15.5-21. A radial peaking factor of 1.65 is used.

(4) For the Regulatory Guide 1.25 analysis all of the gap activity in the damaged rods is released to the spent fuel pool and consists of 10% of the total noble gases and radioactive iodine inventory in the rods at the time of the accident with the following gap percentage exceptions which are based on NUREG/CR 5009 [24] as appropriate: 14% of the Kr-85, 5% of the Xe-133, 2% of the Xe-135, and 12% of the 1-131.

(5) Noble gases released in the containment are released through the Shield Building vent to the environment.

(6) In the Regulatory Guide 1.25 analysis the iodine gap inventory is composed of inorganic species (99.75%) and organic species (0.25%).

(7) A filter efficiency of-90% for inorganic iodine and 30% for organic iodine for the purge air exhaust filters is used since no relative humidity control is provided.

(8) No credit is taken for natural decay after the activity has been released to the atmosphere.

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Enclosure 1 WBN Unit 2 Revised FSAR Section 15.5 Dose Analysis (9) The short-term (i.e., 0-2 hour) atmospheric dilution factors at the exclusion area boundary and low population zone given in Table 15A-2 are used. The thyroid dose utilizes ICRP-30 [25] iodine dose conversion factors. Doses are based on the dose models presented in Appendix 15A.

15.5.6.2 Fuel Handling Accident Based on Regulatory Guide 1.183 The analysis of a postulated fuel handling accident in the Auxiliary Building refueling Area is based on Regulatory Guide 1.183. i.e., Alternate Source Terms (AST). The parameters used for this analysis are listed in Table 15.5-20.a.

The bases for evaluation are:

(1) In the Regulatory Guide 1.183 analysis, the accident occurs 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months after plant shutdown. Radioactive decay of the fission product inventory during the interval between shutdown and placement of the first spent fuel assembly into the spent fuel pit is taken into account.

(2) In the Regulatory Guide 1.183 analysis, damage was assumed for all rods in one assembly.

(3) The assembly damaged is the highest powered assembly in the core region to be discharged. The values for individual fission product inventories in the damaged assembly are calculated assuming full-power operation at the end of core life immediately preceding shutdown. Nuclear core characteristics used in the analysis are given in Table 15.5-21. A radial peaking factor of 1.65 is used.

(4) The Regulatory Guide 1.183 analysis assumes all of the gap activity in the damaged rods is released to the spent fuel pool and consists of 8% 1-131, 10% Kr-85, and 5% of other noble gases and other halogens.

(5) Noble gases released to the Auxiliary Building spent fuel pool are released through the Auxiliary Building vent to the environment.

(6) In the Regulatory Guide 1.183 analysis, the iodine gap inventory is composed of inorganic species (99.85%) and organic species (0.15%).

(7) In the Regulatory Guide 1.183 analysis, the overall inorganic and organic iodine spent fuel pool decontamination factor is 200.

(8) In the Regulatory Guide 1.183 analysis, all iodine escaping from the Auxiliary Building spent fuel pool is exhausted unfiltered through the Auxiliary Building vent.

(9) No credit is taken for the ABGTS or Containment Purge System Filters in the analysis.

(10) No credit is taken for natural decay either due to holdup in the Auxiliary Building or after the activity has been released to the atmosphere.

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Enclosure 1 WBN Unit 2 Revised FSAR Section 15.5 Dose Analysis (11) The short-term (i.e., 0-2 hour) atmospheric dilution factors at the exclusion area boundary and low population zone given in Table 15A-2 are used. The thyroid dose utilizes ICRP-30 [25] iodine dose conversion factors. Doses are based on the dose models presented in Appendix 15A.

15.5.6.3 Fuel Handling Accident Results The radiation dose results of the Regulatory Guide 1.25 with the containment closed fuel handling accident (FHA) are given in Table 15.5-23. For a FHA inside containment, no allowance has been made for possible holdup or mixing in the primary containment or isolation of the primary containment as a result of a high radiation signal from the monitors in the ventilation systems for the case where containment penetrations are closed to the Auxiliary Building. However, the containment purge filters are credited. Dose equations in TID-14844

[23] were used to determine the dose. Dose conversion factors in ICRP-30 [25] were used to determine thyroid doses in place of those found in TID-14844.

The ventilation function of the reactor building purge ventilating system (RBPVS) is not a safety-related function. However, the filtration units and associated exhaust ductwork do provide a safety-related filtration path following a fuel-handling accident prior to automatic closure of the associated isolation valves. The RBPVS contains air cleanup units with prefilters, HEPA filters, and 2-inch-thick charcoal adsorbers. This system is similar to the auxiliary building gas treatment system except that the latter is equipped with 4-inch-thick charcoal adsorbers.

Anytime fuel handling operations are being carried on inside the primary containment, either the containment is isolated or the reactor building purge filtration system is operational. The assumptions listed above are, therefore, applicable to a fuel handling accident inside primary containment.

The thyroid, gamma, and beta doses for FHAs for the closed containment are given in Table 15.5-23 for the exclusion area boundary and low population zone. These doses are less than 25% of the 10 CFR 100.11 limits of 300 rem to the thyroid, and 25 rem gamma to the whole body. These doses are calculated using the computer code FENCDOSE [16].

The whole body, beta, and thyroid doses to control room personnel from the radiation sources discussed above are presented in Table 15.5-23. The doses are calculated by the COROD computer code [17]. Parameters for the control room analysis are found in Table 15.5-14. The dose to whole body is below the 10 CFR 50 Appendix A, GDC 19 limit of 5 rem for control room personnel and the thyroid dose is below the limit of 30 rem.

The radiation dose results of the Regulatory Guide 1.183 fuel handling accident (FHA) are given in Table 15.5-23. Alternate source term (AST) described in RG 1.183 was selectively used to evaluate the FHA due to an event in the spent fuel pool located in the Auxiliary Building or in the containment when the equipment hatch or both doors in a personnel air lock are open. As part of this selective implementation of AST, the following assumptions are used in the analysis:

The total effective dose equivalent (TEDE) acceptance criterion of 10 CFR 50.67(b)(2) replaces the previous whole body and thyroid dose guidelines of 10 CFR 100.11.

The gap activity is revised to be consistent with that required by RG 1.183.

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Enclosure I WBN Unit 2 Revised FSAR Section 15.5 Dose Analysis

The decontamination factors were changed to be consistent with those required by RG.

1.183.

" No Auxiliary Building isolation is assumed.

" No filtration of the release from the Containment or the spent fuel pool to the environment by the Containment Purge filters or the ABGTS is assumed.

The evaluation for the FHA at the spent fuel pool is a bounding analysis for a dropped assembly in containment when the containment is open. The release point for the containment purge system is the Unit 2 shield building stack. The X/Qs are lower for this release point than for the normal auxiliary building exhaust. In addition, any release from the shield building stack would go through the purge system HEPA and Charcoal filter assemblies prior to release. Currently, when the purge lines isolate on high radiation, the auxiliary building also isolates and ABGTS is actuated. The release point for ABGTS is the shield building stacks and the releases are filtered through HEPA and Charcoal assemblies. Thus AST analysis for the FHA in the Auxiliary Building that considers no filtration is conservative and acceptable as the basis for the containment open evaluation.

The thyroid, gamma, and beta doses for FHAs in the Auxiliary and the open containment are given in Table 15.5-23 for the exclusion area boundary and low population zone. These doses are less than 25% of the 10 CFR 100.11 limits of 300 rem to the thyroid, and 25 rem gamma to the whole body and less than the 10 CFR 50.67 limit of 25 rem TEDE. These doses are calculated using the computer code FENCDOSE [16].

The whole body, beta, and thyroid doses to control room personnel from the radiation sources discussed above are presented in Table 15.5-23. The doses are calculated by the COROD computer code [17]. Parameters for the control room analysis are found in Table 15.5-14. The dose to whole body is below the 10 CFR 50 Appendix A, GDC 19 limit of 5 rem for control room personnel, and the thyroid dose is below the limit of 30 rem and the 1 OCFR 50.67 limit of 5 rem TEDE.

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Enclosure 1 WBN Unit 2 Revised FSAR Section 15.5 Dose Analysis Table 15.5-20 Parameters Used In Fuel Handling Accident Analysis Regulatory Guide 1.25 Analysis Time between plant shutdown and accident 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months Damage to fuel assembly All rods ruptured Fuel assembly activity Highest powered fuel assembly in core region discharged Activity release to spent fuel pool Gap activity in ruptured rods(l)

Radial peaking factor 1.65 Form of iodine activity released elemental iodine 99.75%

methyl iodine 0.25%

Filter efficiencies RBPVS (2) elemental iodine 90%

methyl iodine 30%

Amount of mixing of activity in Auxiliary Building None Meteorology See Table 15.5-14 and Table 15A-2 (1) 10% of the total radioactive iodine except for 12% of 1-131 and 10% of total noble gases, except for 14% for Kr-85, 5% for Xe-1 33 and 2% for Xe-1 35 in the damaged rods at the time of the accident.

(2) Reactor Building Purge Ventilation System E1-6

Enclosure I WBN Unit 2 Revised FSAR Section 15.5 Dose Analysis Table 15.5-20.a Parameters Used In Fuel Handling Accident Analysis Regulatory Guide 1.183 Analysis Time between plant shutdown and accident 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months Damage to fuel assembly All rods ruptured Fuel assembly activity Highest powered fuel assembly in core region discharged Activity release to spent fuel pool Gap activity in ruptured rods(l)

Radial peaking factor 1.65 Form of iodine activity released to spent fuel pool elemental iodine 99.85%(AST) methyl iodine 0.1 5%(AST)

Decontamination factor in spent fuel pool AST Overall=200 Filter efficiencies No credit taken Amount of mixing of activity in Auxiliary Building None Meteorology See Table 15.5-14 and Tablel5A-2 (1) 8% 1-131, 10% Kr-85, and 5% other gasses and other halogens.

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Enclosure I WBN Unit 2 Revised FSAR Section 15.5 Dose Analysis Table 15.5-23 Doses From A Fuel Handling Accident (FHA) (rem)

Doses from Fuel Handling Accident Regulatory Guide 1.183 Analyses FHA in Auxiliary Building (rem) or In Containment - Containment Open (rem) 2 HR EAB 30 DAY LPZ CONTROL ROOM Gamma 3.99"4E 0 4.29E-01 9.27-E -21.20E-01 4.935E!-945.86E-01 Beta 1.!77E+90 1.19E+00 2.734,E 0-3.33E-01 4 0f8E+00Q4.68E+00 Thyroid - ICRP-30 I .577EO+0 5.51 E+01 .662E 1.54E+01 .54.0E*+001.32E+01 TEDE 2.38E+00 6.66E-01 1.02E-00 Doses from Fuel Handling Accident Regulatory Guide 1.25 Analyses FHA in Reactor Building, Containment Closed (rem) 2 HR EAB 30 DAY LPZ CONTROL ROOM Gamma 4.102F-§-1-4.31 E-01 9.5229QFO21 .21 E01 2.477-E- 1.2.72E-01 Beta I1-8:2E*OO1 .24E+OO :2 46E-O1-3.48E-O1 2.207E+002.25E+00 Thyroid - ICRP-30 39.42E+0O4. 15E+O1 9.1 fE'-F=OO-1 .16E+O1 5.209E+006.81 E+00 E1-8

Enclosure 2 WBN Unit 2 - Revised FSAR Section 15.5 Red-Lined E2-1

WATTS BAR 15.5 ENVIRONMENTAL CONSEQUENCES OF ACCIDENTS 15.5.1 Environmental Consequences of a Postulated Loss of AC Power to the Plant Auxiliaries The postulated accidents involving release of steam from the secondary system will not result in a release of radioactivity unless there is leakage from the reactor coolant system (RCS) to the secondary system in the steam generator. A conservative analysis of the potential offsite doses resulting from this accident is presented with steam generator leakage as a parameter. This analysis incorporates assumptions of a Technical Specification limit of 0.1 pCi/gm 1-131 dose equivalent, and a realistic source term. Parameters used in both the realistic and conservative analyses are listed in Table 15.5-1.

The realistic assumptions that determine the equilibrium concentrations of isotopes in the secondary system are as follows:

(1) Primary coolant activity is associated with 0.125% defective fuel and is given in Table 11.1-7.

(2) The iodine partition factor in the steam generators is:

amount of iodine/unit mass steam , 0.01 amount of iodine/unit mass liquid

(3) No noble gas is dissolved or contained in the steam generator water, i.e., all noble gas leaked to the secondary system is continuously released with steam from the steam generators through the condenser off gas system.

(4) The 0-2 and 2-8 hour atmospheric dilution factors given in Appendix 15A and Table 15.5-14; the 0-8 hour breathing rate of 3.47 x 10- 4 m3/sec are applicable. Doses are based on the dose models in Appendix 15A.

(5) Primary and Secondary side source terms are based on ANSI/ANS-18.1-1984.

Assumptions used for the conservative analysis are the same as the realistic assumptions except the Secondary side source terms at the Technical Specification limit of 0.1 pCi/gm 1-131 dose equivalent are assumed.

The steam releases to the atmosphere for the loss of AC power are in Table 15.5-1.

The gamma, beta, and thyroid doses for the loss of AC power to the plant auxiliaries at the exclusion area boundary and low population zone are in Table 15.5-2 for the realistic and conservative analyses. These doses are calculated by the FENCDOSE ENVIRONMENTAL CONSEQUENCES OF ACCIDENTS 15.5-1

WATTS BAR computer code[1 6]. The doses for this accident are less than 25 rem whole body, 300 rem beta and 300 rem thyroid. This is well within the limits as defined in 10 CFR 100.

The whole body, beta, and thyroid doses to control room personnel from the radiation sources discussed above are presented in Table 15.5-2. The doses are calculated by the COROD computer code [171. Parameters for the control room analysis are found in Table 15.5-14. The dose to whole body is below the GDC 19 limit of 5 rem for control room personnel, and thyroid dose is below the limit of 30 rem.

Dose equations in TID-14844 [23] were used to determine the dose. Dose conversion factors in ICRP-30 [25] were used to determine thyroid doses in place of those found in TID-14844.

15.5.2 Environmental Consequences of a Postulated Waste Gas Decay Tank Rupture Two analyses of the postulated waste gas decay tank rupture are performed:

(1) a realistic analysis, and (2) an analysis based on Regulatory Guide 1.24 (Reference 2). The parameters used for each of these analyses are listed in Table 15.5-3.

The assumptions for the Regulatory Guide analysis are:

(1) The reactor has been operating at full power with 1% defective fuel for the RG 1.24 analysis.

(2) The maximum content of the decay tank assumed to fail is used for the purpose of computing the noble gas inventory in the tank. Radiological decay is taken into account in the computation only for the minimum time period required to transfer the gases from the reactor coolant system to the decay tank. For the Regulatory Guide 1.24 analysis, noble gas and iodine inventories of the tank are given in Table 15.5-4. For the realistic analysis, source terms are based on ANSI/ANS-18.1-1984 methodology[1 4 ].

(3) The tank rupture is assumed to occur immediately upon completion of the waste gas transfer, releasing the entire contents of the tank through the Auxiliary Building vent to the outside atmosphere. The assumption of the release of the noble gas inventory from only a single tank is based on the fact that all gas decay tanks will be isolated from each other whenever they are in use.

(4) The short-term (i.e., 0-2 hour) dilution factor at the exclusion area boundary given in Appendix 15A is used to evaluate the doses from the released activity. Doses are based on the dose models presented in Appendix 15A.

The gamma, beta, and thyroid doses for the gas decay tank rupture at the exclusion area boundary and low population zone are given in Table 15.5-5 for both the realistic and Regulatory Guide 1.24 analyses.

15.5 ENVIRONMENTAL CONSEQUENCES OF ACCIDENTS

WATTS BAR (5) The whole body, beta, and thyroid doses to control room personnel from the radiation sources discussed above are presented in Table 15.5-5. The doses are calculated by the COROD computer code [17]. Parameters for the control room analysis are found in Table 15.5-14. The dose to whole body is below the GDC 19 limit of 5 rem for control room personnel, and thyroid dose is below the limit of 30 rem.

Dose equations in TID-14844 [23] were used to determine the dose. Dose conversion factors in ICRP-30 [25J were used to determine thyroid doses in place of those found in TID-14844.

15.5.3 Environmental Consequences of a Postulated Loss of Coolant Accident The results of the analysis presented in this section demonstrate that the amounts of radioactivity released to the environment in the event of a loss-of-coolant accident do not result in doses which exceed the reference values specified in a 10 CFR 100.

The analysis is based on Regulatory Guide 1.4[3]. The parameters used for this analysis are listed in Table 15.5-6. In addition, an evaluation of the dose to control room operators and an evaluation of the offsite doses resulting from recirculation loop leakage are presented.

Fission Product Release to the Containment Following a postulated double-ended rupture of a reactor coolant pipe with subsequent blowdown, the emergency core cooling system keeps cladding temperatures well below melting, and limits zirconium-water reactions to an insignificant level, assuring that the core remains intact and in place. As a result of the increase in cladding temperature and rapid depressurization of the core, however, some cladding failure may occur in the hottest regions of the core. Thus, a fraction of the fission products accumulated in the pellet-cladding gap may be released to the reactor coolant system and thereby to the primary containment.

In this analysis, based on Regulatory Guide 1.4[31, a total of 100% of the noble gas core inventory and 25% of the core iodine inventory is assumed to be immediately available for leakage from the primary containment. Of the halogen activity available for release, it is further assumed that 91% is in elemental form, 4% in methyl form, and 5% in particulate form. The core inventory of iodines and noble gases is listed in Table 15.1-5.

Primary Containment Model The quantity of activity released from the containment was calculated with a single volume model of the containment.

If it is assumed that there are no sources of activity following the initial instantaneous release of fission products to the containment, the equation which describes the time dependent activity or quantity of material in a component is:

ENVIRONMENTAL CONSEQUENCES OF ACCIDENTS 15.5-3

WATTS BAR dt = -AijAij(t) Pij(t) where Aij is the activity or quantity of material i in component j. Pij is the rate at which activity or material i is added to component j, and Aij is the rate at which activity or material i is removed or lost from component j. Ifboth A and P are independent of time, then for one material and one component one obtains the solution:

e-At +PI-e -At(2 A = A0e-(1-e-) (2) where A0 is the initial activity. However, in general, P is time dependent and in some cases A is also time dependent.

The addition of material to the component, Pij(t), may come from two sources: (1) flow from another component in the system may add material to the component, (2) material may be produced within the component by radioactive decay. Thus, the addition rate for material i to component j can be expressed as:

P~ + P ~)

(t) 2 (3)

Pij(t) = pij)(t)+ P3M where:

n Pj M cijj-j(t)Aijj(t);cijj_j(t) is the transfer coefficient jj~j n of i from component jj to j, and P (t) ii- Aiij(t); Tii- i is the rate of production ii of i from ii in componentj. Note that Yii-i is not normally a function of time or component.

Similarly, the loss from a component can be due to: (1) loss within the component (such as radioactive decay), (2) flow out of the component to other components, and (3) removal from the system. Thus, the loss rate from component j for material i can be expressed as:

Aij(t) = X +A(2) (3)4) where Ai is the removal rate inside the component due to radioactive decay (neither time nor component dependent),

15.5-4 ENVIRONMENTAL CONSEQUENCES OF ACCIDENTS

WATTS BAR n

A(D)(t) = E f1j-jj(t);fij-jj(t) is the transfer coefficient of material i from component j tojj, jj j and A(Ii3)(t) is the removal from the system.

A computer program Source Transport Program (STP) has been developed to solve equation (1) for each isotope and for two halogen forms (i.e., elemental and or organic). From this, the isotopic concentration airborne in the containment as a function of time and the integrated isotopic leakage from the containment for a given time period can be obtained. Parameters used in the loss-of-coolant accident analysis are listed in Table 15.5-6.

Modeling of Removal Process For fission products other than iodine, the only removal processes considered are radioactive decay and leakage.

The fission product iodine is assumed to be present in the containment atmosphere in elemental, organic, and particulate form. It is assumed that 91% of the iodine available for leakage from the containment is in elemental (i.e., 12 vapor) form, 4% is assumed to be in the form of organic iodine compounds (e.g., methyl iodine), and 5% is assumed to be absorbed on airborne particulate matter. In this analysis it was conservatively assumed that the organic form of iodine is not subject to any removal processes other than radioactive decay and leakage from the containment. The elemental and particulate forms of iodine are assumed to behave identically.

The effectiveness of the ice condenser for elemental iodine removal is described in Section 6.5.4. For the calculation of doses, the ice condenser was treated as a time dependent removal process. The time dependent ice condenser iodine removal efficiencies for the Regulatory Guide 1.4 analysis are given in Table 15.5-7.

Ice Condenser The ice condenser is designed to limit the leakage of airborne activity from the containment in the event of a loss-of-coolant accident. This is accomplished by the removal of heat released to the containment during the accident to the extent necessary to initially maintain that structure below design pressure and then reduce the pressure to near atmospheric. The addition of an alkaline solution such as sodium tetraborate enhances the iodine removal qualities of the melting ice to a point where credit can be assumed in the radiological analyses.

The operation of the containment deck fans (air return fans) is delayed for approximately 10 minutes following a Phase B isolation signal resulting from the loss-of-coolant accident.

This delay in fan operation yields an initial inlet steam-air mixture into the ice condenser of greater than 90% steam by volume which results in more efficient iodine removal by the ice condenser.

ENVIRONMENTAL CONSEQUENCES OF A CCIDENTS 15.5-5

WATTS BAR As a result of experimental and analytical efforts, the ice condenser system has been proven to be an effective passive system for removing iodine from the containment atmosphere following a loss-of-coolant accident. (Reference 4)

With respect to iodine removal by the ice condenser, the following assumptions were made:

(1) The ice condenser is only effective in removing airborne elemental and particulate iodine from the containment atmosphere.

(2) The ice condenser is modeled as a time dependent removal process.

(3) The ice condenser is no longer effective in removing iodine after all of the ice has been melted using the most conservative assumptions.

Primary Containment Leak Rate The primary containment leak rate used in the Regulatory Guide 1.4 analysis for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is the design basis leak rate guaranteed in the technical specifications regarding containment leakage and it is 50% of this value for the remainder of the 30 day period. Thus, for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following the accident, the leak rate was assumed to be 0.25% per day and the leak rate was assumed to be 0.125% per day for the remainder of the 30 day period.

The leakage from the primary containment can be grouped into two categories: (1) leakage into the annulus volume and (2) through line leakage to rooms in the Auxiliary Building (see Figure 15.5-1). The environmental effects of the core release source events have been analyzed on the basis that 25% of the total primary containment leakage goes to the Auxiliary Building.

The leakage paths to the Auxiliary Building are tested as part of the normal Appendix J testing of all containment penetrations. An upper bound to leakage to the Auxiliary Building was estimated to be 25% of the total containment leakage. Selecting an upper bound is conservative because an increasing leakage fraction to the Auxiliary Building results in an increasing calculated offsite dose. This upper bound was also selected on the basis that it is large enough to be verified by testing. The periodic Appendix J testing will assure that leakage to the Auxiliary Building remains below 25%. The remaining 75% of the leakage goes to the annulus.

Bypass Leakage Paths There are no bypass paths for primary containment leakage to go directly to the atmosphere without being filtered. For further details see the discussion on Type E leakage paths in Section 6.2.4.3.1.

Auxiliary Building Release Path The Auxiliary Building allows holdup and is normally ventilated by the auxiliary building ventilation system. However, upon an ABI signal following a loss-of-coolant accident, the normal ventilation systems to all areas of the Auxiliary Building are shutdown and 15.5-6 ENVIRONMENTAL CONSEQUENCES OF ACCIDENTS

WATTS BAR isolated. Upon Auxiliary Building isolation, the Auxiliary Building gas treatment system (ABGTS) is activated to provide ventilation of the area and filtration of the exhaust to the atmosphere. This system is described in Section 6.2.3.2.3.

Fission products which leak from the primary containment to areas of the Auxiliary Building are diluted in the room atmosphere and travel via ducts and other rooms to the fuel handling area or the waste packaging area where the suctions for the Auxiliary Building gas treatment system are located. The mean holdup time for airborne activity in the Auxiliary Building areas other than the fuel handling area is greater than one hour with the Auxiliary Building isolated and both trains of the ABGTS operating. It has been conservatively assumed in the estimation of activity release that activity leaking to the Auxiliary Building is directly released to the environment for the-first four minutes and then through the ABGTS filter system, with a conservatively assumed mean hold-up time of 0.3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months in the Auxiliary Building before being exhausted. In the Regulatory Guide 1.4 analysis the ABGTS filter system is assumed to have a removal efficiency of 99% for elemental, organic, and particulate iodines. Minor leakage into the ABGTS and EGTS ductwork allows some unfiltered Auxiliary Building air to be released to the environment. This leakage, quantified by testing, is modeled in the LOCA analysis as indicated in Table 15.5-6 and does not significantly impact doses.

The Auxiliary Building internal pressure is maintained at less than atmospheric during normal operation (see Section 9.4.2 and 9.4.3), thereby preventing release to the environment without filtration following a LOCA. The annulus pressure is maintained more negative than the Auxiliary Building internal pressure during normal operation and after a DBA. Therefore, any leakage between the two volumes following a LOCA is into the annulus.

Shield Building Releases The presence of the annulus between the primary containment and the Shield Building reduces the probability of direct leakage from the vessel to the atmosphere and allows holdup, dilution, sizing, and plate-out of fission products in the Shield Building. The major factor in the effectiveness of the secondary containment is its inherent capability to collect the containment leakage for filtration of the radioactive iodine prior to release to the environment. This effect is greatly enhanced by the recirculation feature of the air handling systems, which forces repeated filtration passes for the major fraction of the primary containment leakage before release to the environment. Seventy-five percent of the primary containment leakage is assumed to go to the annulus volume.

The initial pressure in the annulus is less than atmospheric. However, the dose analysis conservatively assumes the Annulus is at atmospheric pressure at event initiation. After blowdown, the annulus pressure will increase rapidly due to expansion of the containment vessel as a result of primary containment atmosphere temperature and pressure increases. The annulus pressure will continue to rise due to heating of the annulus atmosphere by conduction through the containment vessel. After a delay, the EGTS operates to maintain the annulus pressure below atmospheric pressure.

The EGTS is essentially an annulus recirculation system with pressure activated valves which allow part of the system flow to be exhausted to atmosphere to maintain ENVIRONMENTAL CONSEQUENCES OF ACCIDENTS 15.5-7

WATTS BAR a "negative" annulus pressure. The system includes absolute and impregnated charcoal filters for removal of halogens. The EGTS combined with ABGTS ensures that all primary containment leakage is filtered before release to the atmosphere.

The EGTS suction in the annulus is located at the top of the containment dome, while nearly all penetrations are located near the bottom of the containment (see Section 6.2), thereby minimizing the probability of leakage directly from the primary containment into the EGTS.

Transfer of activity from the annulus volume to the EGTS suction is assumed to be a statistical process similar mathematically to the decay process, (i.e., the rate of removal from the annulus is proportional to the activity in the annulus). This corresponds an assumption that the activity is homogeneously distributed throughout the mixing volume. Because of the low EGTS flow rate (compared to the annulus volume), the thermal convection due to heating of the containment vessel, and the relative locations of the EGTS suctions (at the top of the dome) and the EGTS recirculation exhausts (at the base of the annulus), a high degree of mixing can be expected. It is conservatively assumed that only 50% of the annulus free volume is available for mixing of activity in the Regulatory Guide 1.4 analysis.

Tables 15.5-8 and 15.5-8A list the EGTS and recirculation flow rates as a function of time after the LOCA, which were used for calculation of activity releases for the Regulatory Guide 1.4 analysis. Table 15.5-8 flow rates are as a result of a postulated single failure loss of one train of EGTS concurrent with LOCA. Table 15.5-8A flow rates are as a result of an alternate single failure scenario resulting in one pressure control train in full exhaust to the shield building exhast stack while the other train remains functional. Both EGTS fans are in service until operator action is taken to place one fan in standby between one and two hours post accident. The flow path of fission products which are drawn into the air handling systems is shown schematically in Figure 15.5-1 where:

L0 Represents the flow of activity from primary containment to the annulus L1 Represents the flow of activity from primary containment to the Auxiliary Building L Represents the flow of activity from the annulus into the EGTS K Represents the ratio of EGTS recirculation flow to total EGTS flow rate nf Represents the appropriate filter efficiency Effectiveness of Double Containment Design The analysis has demonstrated clearly the benefits of the double containment concept.

As would be expected for a double barrier arrangement, the second barrier acts as an effective holdup tank, resulting in substantial reduction in the two-hour inhalation and whole body immersion doses. The expected offsite doses for the 30-day period at the 15.5-8 ENVIRONMENTAL CONSEQUENCES OFACCIDENTS

WATTS BAR low population zone are also substantially reduced, since the holdup process is effective for the duration of the accident.

The EGTS exhaust flow rate is dependent on the rate of air inleakage to the annulus.

In fact, after about 30 minutes following blowdown of the reactor vessel the EGTS exhaust flow is approximately equal to the air inleakage rate. Studies[5] made of leak rates from typical concrete buildings of this type have resulted in leak rates from 4% to 8% per day at a pressure differential of 14 inches of water. Although the pressure differential in this case will be much lower than this value, it has been assumed that a shield building inleakage flow of 250 cfm exists throughout the 30-day period for the single failure scenario which results in loss of one EGTS train concurrent with a LOCA.

The inleakage flow for the single failure scenario which results in one pressure control train in full exhaust to the shield building exhaust stack (while the other train remains functional) was conservatively assumed to be greater since the resulting annulus pressure is more negative than the original single failure scenario loss of one EGTS train. The long term inleakage flow rates of 832 cfm (until operator action to place one fan in standby) and 604 cfm thereafter are used in the dose analysis. This inleakage flow includes leakage past ventilation system primary containment isolation valves assuming that a single isolation valve fails in the open position.

In order to evaluate the effectiveness of the Shield Building, the following case was analyzed:

50% Mixing Case At the beginning of the accident, the EGTS starts exhausting filtered fission products to the environs (see Tables 15.5-8 and 15.5-8A). At approximately 114 seconds for the loss of one EGTS train the Annulus pressure becomes less than -0.25 inches w.g.

and the effluents are filtered for the duration of the accident. At approximately 60 seconds (for the single failure scenario which results in one pressure control train in full exhaust to the shield building exhaust stack while the other train remains functonal) the annulus pressure becomes less than minus 0.25 inches w.g., and the effluents are filtered for the duration of the accident. All of the primary containment leakage going to the shield building is assumed to be uniformly mixed in 50% of the annulus free volume.

Emergency Gas Treatment System Filter Efficiencies The EGTS takes suction from the annulus, and the exhaust gases are drawn through two banks of impregnated charcoal filters in series. Sufficient filter capacity is provided to contain all iodines, inorganic, organic, and particulate available for leakage. Since the air in the annulus is dry, filter efficiencies of greater than 99% are attainable as reported in ORNL-NSIC-4[6]. Heaters and demisters have been incorporated upstream of the filters resulting in a relative humidity of less than 70% in the air entering the filters which further ensures high filter efficiency.

In the Regulatory Guide 1.4 analysis however, an overall removal efficiency of 99% for elemental, organic, and particulate iodine is assumed for the two filter banks in series.

ENVIRONMENTAL CONSEQUENCES OF ACCIDENTS 15.5-9

WATTS BAR Discussion of Results The gamma, beta, and thyroid doses for the LOCA at the exclusion area boundary and the low population zone are given in Table 15.5-9. These doses are calculated by the FENCDOSE computer code[1 6]. The doses are based on the atmospheric dilution factors and dose models given in Appendix 15A. The doses for this accident are less than 25 rem whole body, 300 rem beta, and 300 rem thyroid. The doses are well within the 10 CFR 100 guidelines and reflect the worst case values in consideration of both single failure scenarios.

Loss of Coolant Accident - Environmental Consequences of Recirculation Loop Leakage Component leakage in the portion of the emergency core cooling system outside containment during the recirculation phase following a loss of coolant accident could result in offsite exposure. The maximum potential leakage for this equipment is specified is Table 6.3-6. This leakage refers to specified design limits for components and normal leakage is expected to be well below those upper limits. Recirculation is assumed in the analysis to start at 10 minutes after the loss of coolant accident. At this time the sump temperature is approximately 1601F (Figure 6.2.1-3). The enthalpy of the sump is approximately 130 BTU/Ib. The enthalpy of saturated liquid at 1.0 atmosphere pressure and 212°F is greater than 130 BTU/lb. Therefore, there will be no flashing of the leakage from recirculation loop components, and an iodine partition factor of 0.1 is assumed for the total leakage.

The analysis of the environmental consequences is performed as follows:

Core iodine inventory given in Table 15.1-5 is used. The water volume is comprised of water volumes from the reactor coolant system, accumulators, refueling water storage tank, and ice melt. All the noble gases are assumed to escape to the primary containment. Ninety-seven percent of trituim was assumed to remain liquid and accumulate in the sump, while 3% was assumed to go airborne to the containment. An alternate analysis was also performed assuming 100% of the tritium goes airborne into the containment. Radioactive decay was taken into account in the dose calculation.

The major assumptions used in the analysis are listed in Table 15.5-12. The offsite doses at the exclusion area boundary and low population zone for the analysis are given in Table 15.5-13 and reflect the worst case values in consideration of 3%

airborne tritium or 100% airborne tritium. The atmospheric dilution factors and dose models discussed in Appendix 15A are used in the dose analysis. The whole body, beta, and thyroid doses to control room personnel from the radiation sources discussed above are presented in Table 15.5-13. The doses are calculated by the COROD computer code [17]. Parameters for the control room analysis are found in Table 15.5-14. The dose to whole body is below the GDC 19 limit of 5 rem for control room personnel, and thyroid dose is below the limit of 30 rem.

Dose equations in TID-14844 [23] were used to determine the dose. Dose conversion factors in ICRP-30 [25] were used to determine thyroid doses in place of those found in TID-14844.

15.5-10 ENVIRONMENTAL CONSEQUENCES OFACCIDENTS

WATTS BAR Loss of Coolant Accident - Control Room Operator Doses In accordance with General Design Criterion 19, the control room ventilation system and shielding have been designed to limit the whole body gamma dose during an accident period to 5 rem, the thyroid dose to 30 rem and the beta skin dose to 30 rem.

The doses to personnel during a post-accident period originate from several different sources. Exposure within the control room may result from airborne radioactive nuclides entering the control room via the ventilation system. In addition, personnel are exposed to direct gamma radiation penetrating the control room walls, floor, and roof from:

(1) Radioactivity within the primary containment atmosphere (2) Radioactivity released from containment which may have entered adjacent structures (3) Radioactivity released from containment which passes above the control room roof Further exposure of control room personnel to radiation may occur during ingress to the control room from the exclusion area boundary and during egress from the control room to the exclusion area boundary.

In the event of a radioactive release incident, the control room is isolated automatically by a safety injection system signal and/or by radiation signal from beta detectors located in the air intake stream common to the air intake ports at either end of the Control Building. These redundant signals are routed to redundant controls which actuate air-operated isolation dampers downstream of the beta detectors. Operation of the emergency pressurizing fans with inline HEPA filters and charcoal adsorbers is also initiated by these signals. Simultaneously, recirculation air is rerouted automatically through the HEPA filters and charcoal adsorbers. Approximately 711 cfm of outside air, the emergency pressurization air, flows through a duct routed to the emergency recirculation system upstream of the HEPA filters and charcoal adsorbers.

This flow of outside air provides the control room with a slight positive pressure relative to the atmosphere outside and to surrounding structures. In addition, the equivalent of 51 cfm of unfiltered outside air enters through the main control room doors and other sources. Isolation dampers located in each intake line may be selectively closed by control room personnel. The selection between the two would be based on the objective of admitting a minimum of airborne activity to the control room via the makeup airflow.

The control room ventilation flow system is shown in Figure 9.4-1.

To evaluate the ability of the control room to meet the requirements of General Design Criterion 19, a time-dependent model of the control room was developed. In this model, the outside air concentration enters the control room via the isolation damper bypass line and the HEPA filters and charcoal absorbers. The concentration in the room is reduced by decay, leakage out, and by recirculation through the HEPA filters ENVIRONMENTAL CONSEQUENCES OF ACCIDENTS 15.5-11

WATTS BAR and charcoal absorbers. Credit for filtration is taken during two passes through the charcoal absorbers. Using these assumptions, the following equations for the rate of change of the control room concentrations are obtained:

dMc dM _ C0 (1 _KI)L/V_(L/V)M_- M- xM (1) dt dN Rc dt -V (I-K2) M-(L/V)N-XN (2)

C(t) = M(t) + N(t) (3)

Where:

M(t) = Once-filtered time-dependent concentration N(t) = Twice-filtered (or more) time-dependent concentration C(t) = Total time-dependent concentration in control room Co = Concentration of isotope entering air intake K1 = Filter efficiency for a particular isotope during first pass K2 = Filter efficiency for a particular isotope during second pass L = Flow rate of outside air into control room and leakage out of control room Rc = Recirculated air flow rate through filters A = Decay constant V = Control room free volume These equations are readily solvable if Co is constant or a simple function of time during a time interval. Since Co consists of a number of terms involving exponentials, it was assumed to be constant during particular time intervals corresponding to the average concentration during each interval as described below. Solving equations (1), (2), and (3) yields:

C~Wt) x[(I - K2) (I - e-Wmt) + ,,-n;*( (e e-Wn) )oLCeWtewt I 4 LI-K1I-K2C°]

WmVL 15.5-12 ENVIRONMENTAL CONSEQUENCES OFACCIDENTS

WATTS BAR Where:

XV)

Wm = (L + Rc +

V (L +V)

W V

The value of Co used in equation (4) is determined as follows:

ti +

J R dt (5)

C0 i = (X/Q)i t--L i+l -ti Coi= Average concentration of activity outside control room during ith time period (Ci/m 3 ).

(x/Q)i= Atmospheric dilution factor (sec/m3 ) during the ith time period.

R= Time dependent release rate of activity from containment (Ci/sec).

The atmospheric dilution factors were determined using the accumulated meteorological data on wind speed, direction, and duration of occurrence obtained from the Watts Bar plant site applied-to a building wake dilution model. The dilution factors are calculated by the ARCON96 methodology[81 and are the maximum values for each time period. The worst case is Unit 1 exhaust to intake 2. These factors are applied for the first 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , at which time it is assumed that the operator selects intake 1 which has more favorable dilution factors. The values used in the analysis are given in Table 15.5-14.

Equation (4) is used to determine the concentration at any time within a time period and upon integrating and dividing by the time interval gives the average concentration during the time interval due to inflow of radioactivity with outside air as shown:

T Ci(t)dt (6) f~ T-O0 0

Where:

ENVIRONMENTAL CONSEQUENCES OF ACCIDENTS 15.5-13

WATTS BAR T =t - ti_1 t = Time after accident ti_1 = Time at end of previous time period Further contributions to the concentration during the time period are due to the concentrations remaining from prior time periods. These contributions are obtained from the following equations:

CR(i+j) = MR(i+j) +NR(i+j) (7) dMR(i+j) = -(L/V+Rc/V+X)MR(i+j) (8) dt dNR(i +j) = (Rc/V)(1 - K 2 )MR(i +j) - (L/V + X)NR(i +j) (9) dt With initial conditions:

MR(i+j) (0) = MR0(i = (Once-filtered concentration at end of the ith time period.)

NR(+/--j) (0) = NR 0(i) = (Twice-filtered, or more, concentration at end of the ith time period.)

Solving equations (8) and (9) and substituting certain initial condition relations, equation (7) becomes:

CR(i +J) = CRO(i)ew N(t-u)_MR0(i)K 2(e-WN(t-ti) - eWM(t ti)) (10)

Integrating equation (10) for each of the prior time periods gives the contribution from these time periods to the present time period. The average concentration is determined for these contributions using the method of equation (6).

Filter efficiencies of 95% for elemental and particulate iodine and 95% for organic iodine were deemed appropriate for the first filter pass. Since the concentrations of iodine in the main control room are such reduced as a result of this filtration, the efficiencies were reduced for the second pass to 70% for elemental and particulate iodine, and 70% for organic iodine.

To account for the unfiltered inleakage, a bypass leak rate (BPR) of 51 cfm was added to the makeup flow (L in equation (1)) of 711 cfm, and the filter factor for the first pass was decreased by the ratio L/(L+BPR). The filter efficiencies for the second pass are not affected by the unfiltered inleakage.

The filter efficiency for noble gases was taken as zero for all cases.

15.5-14 ENVIRONMENTAL CONSEQUENCES OFACCIDENTS

WATTS BAR The above equations were incorporated into computer program COROD[ 171 together with appropriate equations for computing gamma dose, beta dose, and inhalation dose using these average nuclide concentrations and time periods. The whole body gamma dose calculation consists of an incremental volume summation of a point kernel over the control room volume. The principal gammas of each isotope are used to compute the dose from each isotope. The dose computations for beta activity were based on a semi infinite cloud model. Doses to thyroid were based on activity to dose conversion factors. (The equations and various data are given below.) The doses from these calculations are presented in Table 15.5-9. Gamma dose contributions from shine through the control room roof due to the external cloud and from shine through the control room walls from adjacent structures and from containment are computed using an incremental volume summation of a point kernel which includes buildup factors for the concrete shielding. For the calculation of shine through the control room roof, an atmospheric, rectangular volume several thousand feet in height and several control room widths was used. The control room roof is a 2 foot 3-inch-thick concrete slab and is the only shielding considered in this calculation. The average isotope concentrations at the control bay for each time period were used as the source concentrations. For the shine from adjacent structures, the shielding consists of the 3-foot-thick (5 feet in certain areas) control room walls. The doses are calculated similarly to the shine dose through the roof. The average isotope concentrations at the control bay intake for each time period are also used for these calculations.

The shine from the spreading room below the control room is also computed in the same manner as adjacent structures.

Shielding for this computation consists of the 8-inch-thick concrete floor. The summation of the incremental elements is performed over the volume of each room or structure of interest.

In addition to the dose due to shine from surrounding structures and from the passing cloud, the shine from the reactor containment building also contributes to the gamma whole body dose to personnel. This contribution is computed in the same manner as the methods used above. Due to the location of the Auxiliary Building between the Reactor Buildings and the control room and the thicker control room auxiliary building wall near the roof, the minimum ray path through concrete from the containment into the control room below 10 feet above the control floor, is 8 feet. All nuclides released to containment are assumed uniformly distributed and their time-dependent concentrations were used to compute the dose. The dose computed from this source is small.

Several doors penetrate the control room walls, and the dose at these areas would be larger than the doses calculated as described above. The potential shine at these doors and at other penetrations has been evaluated. As a result, hollow steel doors filled with no. 12 lead shot have been incorporated into the design of the shield wall between the control room and the Turbine Building. These doors provide shielding comparable to the concrete walls. Shine through other penetrations was found to be negligible.

ENVIRONMENTAL CONSEQUENCES OFACCIDENTS 15.5.15

WATTS BAR Another contribution to the total exposure of control room personnel is the exposure incurred during ingress from and egress to the exclusion area boundary. The doses due to ingress and egress were computed based on the following assumptions:

(1) Five minutes are required to leave the control room and arrive at car or vice versa.

(2) The distance traveled on the access road to the site exclusion boundary is estimated to be 1500 meters. The average car speed is assumed to be 25 mph.

(3) One one-way trip first day, one round-trip/day 2nd through 30th days.

The control room occupancy factors used in this calculation were taken from Murphy and Campe[91 . They are:

100% occupancy 0-24 hours 60% occupancy 1-4 days 40% occupancy 4-30 days.

All atmospheric dilution factors were conservatively based on 5th percentile wind velocity averages.

It was also assumed that initially the makeup air intake would be through the vent admitting the highest radioisotope concentration, but that the main control room personnel would switch intake vents 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months after the accident in order to admit a lower amount of airborne activity to the MCR via the makeup air flow.

The whole body, beta, and thyroid doses from the radiation sources discussed above are presented in Table 15.5-9. The dose to whole body is below the GDC 19 limit of.5 rem for control room personnel, and the thyroid dose is below the limit of 30 rem.

Dose Equations, Data, and Assumptions The dose from gamma radiation originating within the control room is given by:

exp (-ga,I4~ +Yn +zj).Alf Dr =1.69 X 10 Z YTC01 ik Ekfki p (11)

SI m=1 n=l q=i Where:

D7 = Absorbed dose in flesh in mrads 153.5-16 ENVIRONMENTAL CONSEQUENCES OF ACCIDENTS

WATTS BAR 3

TCOTik=Total concentration integrated over time period i of isotope k in curies/m Eke = Energy of gamma e from isotope k in MeV fke = Number of e gammas of isotope k given off per disintegration D--) = Mass attenuation coefficient for flesh determined at the energy of gamma in P cm 2/gram P,,=Linear attenuation coefficient for air determined at the energy of gamma i in inverse meters xm,Yn,zq=Coordinate distances from the dose point to the source volume element (mn,q) in meters Ax,Ay,Az=Dimensions of source element (m,n,q) a=Number of time periods P3=Number of isotopes y=Number of gammas from an isotope E=Number of intervals in the x direction w=Number of intervals in the y direction O=Number of intervals in the z direction The control room radiation dose from gamma radiation originating outside of the control room and penetrating concrete walls is given as:

a~~ ~~) (x exp(-g' lqxm~*z) q

[

I*- I=* Z'_-' +ye+z+ Z2) g= 1.69 x I04 Coi EkXfk, ,C, + 1 tcSeC6)

C4)exp(-p~

i=W k=e I m=1 n=1 q=1 Where:

ENVIRONMENTAL CONSEQUENCES OF ACCIDENTS 15.5-17

WATTS BAR Ipcl = Linear attenuation coefficient of concrete determined at the energy of gamma in inverse meters tc= Concrete shield thickness in meters 8 = Angle between a vector normal to the shield and a vector from the dose point to the source point Bc(picltcsecG) = Buildup factor for concrete Coik = Average concentration of isotope k outside the control room during time period i in 3

curies/m ti1 ,ti = Times at the beginning and end of time period i in hours Other parameters are defined as previously noted.

The dose from beta radiation is given by the semi-infinite cloud immersion dose:

DB = (0.230) (X/Q)LI Q Y, Eikfik] (12)

Li=1 k=1 Where:

DB=Dose due to beta in rem 3

X/Q=Atmospheric dispersion factor during time period in sec/mi Qi=Accumulated activity release of isotope i during time period Eik=Average energy of beta k of isotope i fik=Number of k betas of isotope i per disintegration For beta dose in the control room, equation (12) becomes:

DB = (0.230) iij ,_Eikfik(tj -t- 1 )

i= 1 1 I=1 15.5-18 ENVIRONMENTAL CONSEQUENCES OFACCIDENTS

WATTS BAR Where:

Cii =Average concentration of isotope i during time period j Inhalation Dose (Thyroid)

The inhalation dose for a given period of time has the general form:

DI (X/Q)(B) (Qij)(DCF) (tj -tj_]) (13)

Where:

D1=Thyroid inhalation dose, rem X/Q=Site dispersion factor during time period, sec/m3 B=Breathing rate during time period, m 3/hr Q1j=Average activity release rate during time period j of iodine isotope i DCFi=ICRP-30 Dose conversion factor for iodine isotope i, rem/microcurie inhaled tj=Total time at end of period j, hours For inhalation dose within the control room, equation (13) becomes:

DI = (B) Cij(DCFi (ti-tii_)

In this expression Cij, the average concentration of isotope i during time period j, has replaced the following factor:

(X/Q) Qij The Cij's are those determined by equations (4) and (6). The breathing rate factor B, was taken to be 3.47 x 10-4 m 3/sec, 1.75 x 10-4 m 3/sec, and 2.32 x 10-4 m3 /sec for the time intervals of 0-8 hours, 8-24 hours, and 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months - 30 days, respectively.

ENVIRONMENTAL CONSEQUENCES OF ACCIDENTS 15.5-19

WATTS BAR 15.5.4 Environmental Consequences of a Postulated Main Steam Line Break The postulated accidents involving release of steam from the secondary system will not result in a release of radioactivity unless there is a leakage from the reactor coolant system to the secondary system in the steam generator. An acceptable primary-to-secondary leakage rate for the main steam line break (MSLB) accident is 1 gallon per minute (gpm) for the faulted steam generator loop and 150 gallons per day (gpd) for each unfaulted steam generator.

A calculation determines the offsite and main control room doses resulting from a MSLB incorporating the above primary-to-secondary criteria. The calculation determined that 1 gpm (at standard temperature and pressure) primary-to-secondary leakage in the faulted steam generator results in site boundary doses within 10CFR100 guidelines and control room doses within the 10CFR50, Appendix A, General Design Criteria (GDC)-19 limit. The calculation uses TVA computer codes STP, FENCDOSE and COROD. The STP output is input to COROD, which determines control room operator dose and FENCDOSE, which determines the 30-day low population zone (LPZ) and the 2-hour exclusion area boundary (EAB) dose.

Two methods for determining the resultant dose for a MSLB in accordance with the Standard Review Plan 15.1.5, Appendix A methodology are:

1. A pre-accident iodine spike where the iodine level in the reactor coolant spiked upward to the maximum allowable limit of 14 pCi/gm 1-131 dose equivalent just prior to the initiation of the accident.

2. The reactor coolant at the maximum steady state dose of equivalent 1-131 of 0.265 pCi/gm with an accident initiated iodine spike consisting of a 500 times increase on the rate of iodine release from the fuel.

In both cases, the primary-to-secondary side leak is assumed to be 1 gpm in the faulted steam generator loop and 150 gpd in each faulted loop. The primary side activity release uses the Technical Specification (TS) limit design reactor coolant activities, and the secondary side activity uses the Technical Specification limit of

_<0.1pCi/gm 1-131 dose equivalent.

The steam releases to the atmosphere for the MSLB are in Table 15.5-16.

The gamma, beta and thyroid doses for the MSLB accident at the EAB and LPZ are in Table 15.5-17. The doses from this accident are less than the reference values as listed in 10CFR100 (25 rem whole body and 300 rem thyroid).

The whole body, beta and thyroid doses to control room personnel from the radiation sources discussed above are in Table 15.5-17. The doses are calculated by the COROD computer code. [17] Parameters for the control room analysis are found in Table 15.5-14. The dose to whole body is below the GDC limit of 5 rem for control room personnel, and the thyroid dose is below the limit of 30 rem.

15.5-20 ENVIRONMENTAL CONSEQUENCES OF ACCIDENTS

WATTS BAR Dose equations in TID-14844 [23] determine the dose. Dose conversion factor in ICRP-30 [25] determine thyroid doses in place of those found in TID-14844.

Assumptions for the MSLB accident:

1. RCS letdown flow of 124.39 gpm.

2. RCS letdown demineralizer efficiency is 1.0 for iodines.

3. ANSI/ASN-18.1-1984 spectrum scaled up to 0.265 or 14 pCi/gmequivalent iodine.

4. Two cases were used. In the first, pre-accident iodine spike of 14 pCi/gm 1-131 dose equivalent in the RCS. In the second case, an accident initiated spike which increases the iodine concentration at the equilibrium into the reactor coolant from the fuel rods.

5. Primary side to secondary side leakage of 150 gpd (standard temperature and pressure) per steam generator in the intact loops.

6. The primary-to-secondary leakage mass release to the Environment is 1 gpm (standard temperature and pressure) from the faulted loops.

7. Steam generator secondary inventory released as steam to the atmosphere:

a) total from the non-defective steam generators (0-2 hrs), 433,079 Ibm b) total from the non-defective steam generators (2-8 hrs), 870,754 Ibm c) total from the faulted steam generator (0-30 mins), 96,100 Ibm

8. Iodine partition coefficients from steaming of steam generator water:

i. non-defective steam generators initial inventory and primary-to-secondary leakage, 100.

ii. faulted steam generator initial inventory and primary-to-secondary leakage, 1.0

9. Atmospheric dilution factors, x/Q, are in Table 15A-2 for offsite and Table 15.5-14 for control room personnel.

10. Main control room related assumptions are in Table 15.5-14.

15.5.5 Environmental Consequences of a Postulated Steam Generator Tube Rupture Thermal and hydraulic analyses determine the plant response for a design basis steam generator tube rupture (SGTR), and the integrated primary to secondary break flow and mass releases from the ruptured and intact steam generators (SGs) to the condenser and the atmosphere (Section 15.4.3). An analysis of the environmental consequences of the postulated SGTR has also been performed, utilizing the reactor coolant mass and secondary steam mass releases determined in the base thermal and ENVIRONMENTAL CONSEQUENCES OF ACCIDENTS 15.5-21

WATTS BAR hydraulic analysis (See Reference [38] in Section 15.4). Table 15.5-18 summarizes the parameters used in the SGTR analysis.

The SGTR thermal and hydraulic analysis documents use WBN specific parameters and actual operator performance data, as determined from simulator exercises utilizing the appropriate emergency operating procedures (EOPs). Two cases were analyzed.

Case 1: The primary side activity release use the maximum Technical Specification (TS) limit design reactor coolant activities and an iodine spike immediately after the accident that increases the iodine activity in the reactor coolant by a factor of 500 times the iodine production rate necessary to maintain a steady state concentration of 0.265 pCi/gm of 1-131 dose equivalent. Case 2: The initial reactor coolant activity is at 14 pCi/gm of 1-131 dose equivalent due to a pre-accident iodine spike caused by an RCS transient. For both cases, the secondary side releases uses expected secondary side activities, based on ANSU/ANS-18.1-1984[ 141 as modified for WBN, and on a 150 gpd/steam generator primary-to-secondary-side leakage. Credit was taken for flashing of the primary coolant (References [34] and [35] of Section 15.4), but "scrubbing" of the iodine in the rising steam bubbles by the water in the steam generator was conservatively neglected. A partition factor of 100 applies to iodine in the remaining unflashed coolant which will boil.

The atmospheric diffusion coefficients (X/Q) for the exclusion area boundary (EAB) and offsite dose determination are the same as those used for the LOCA'analysis (Appendix 15A). The X/Q values for the control room operator were determined in the analysis. The LOCA X/Q values are for a release from the shield building vent, whereas the SGTR release is from the top of the main steam valve vault. The methodology for determination of the WBN control room X/Q values is based on computer code ARCON96.

The whole body, beta, and thyroid doses to control room personnel from the radiation sources discussed above are in Table 15.5-19. The doses are calculated by the COROD computer code [17]. Parameters for the control room analysis are in Table 15.5-14. The dose to whole body is below the GDC 19 limit of 5 rem for control room personnel, and thyroid dose is below the limit of 30 rem.

Dose equations in TID-14844 [23] were used to determine the dose. Dose conversion factors in ICRP-30 [25] were used to determine thyroid doses in place of those found in TID-14844.

The gamma, beta, and thyroid dose for the SGTR event are in Table 15.5-19. The doses at the EAB and the low population zone are less than 10% of the 10 CFR 100 limits.

15.5.6 Environmental Consequences of a Postulated Fuel Handling Accident The analysis of the fuel handling accident considers three cases. The first case is for a Fuel Handling Accident inside containment with the containment closed and the Reactor Building Purge System operating. This analysis is discussed in Section 15.5.6.1 and is based on Regulatory Guide 1.25[11] and NUREG 5009[241. The second case is for an accident in the soent fuel nool area located in the Auxiliary 15.5-22 ENVIRONMENTAL CONSEQUENCES OFACCIDENTS

WATTS BAR Building. This case is discussed in Section 15.5.6.2 and is evaluated using the Alternate Source Term based on Regulatory Guide 1.183181, "Alternate Source Terms." The third case considered is an open containment case for an accident inside containment where there is open communication between the containment and the Auxiliary Building. This evaluation is discussed in Section 15.5.6.2 and is based on Regulatory Guide 1.183.The a.alysk., of a postulated fuel handling acident is based on Regulatey Guide 1 .25- and NUPEG/CR 5009.0 T I-hp mpr-."4.,.

cters used for this analysis Arp I;ý+^A ;- Tr- i 1* r- r- 00 The bases for. the R..ulat Guide

_,Y 1.25 evaluations ar.:

(1) In the Rcgulatory Guide 1.25 analysiS the accidlent occurs 100 hourS aftcr plant shutdown. Radioactive decay of the fission produet inventorFy dur~ing the-mintcr~'al between shutdcwn and placement of the firSt spent fuel assembly into the spent fuel pit is taken iRne account.

(2) In the Regulator.y Guide 1.25 analysis damage was au fra, rods in eRe asembey-(3) The assembly damaged is the highest powered assembly in thcoe rgo to be discharged. The values for individual fission product inventories in thc damagd assembly ae rccalculated assuming full power operation. at the *nd Of eorc life immediately prceeeding s~hutdo~wn. Nuclcar eorc ehwarateristiGss used in the analysis are givcn in Table 15.5 21. In thc R.gulate.y Guide 1.25 analysis, a r-adial pcaking factor Of 1.65 is uscd.

(4,) For the Regulatory Guidc 1.25 analysis all of the gap actfivity in the dar-agcd rods *S releascd to the spRnt fuel pe*l and consists of 105 of the total noble gaSc. and -adi"activcocn tcry in the rods at thc time ef the accidcnt with the followingAA a cren~tagc c"~" ption s which arc based On

,,6EG!CR IaS 5 appropriate: 141; of the Kr 85, 57 f the-0-6 Xc.

e 133,,2%*

of the Xe 136, and 12-,%oef h i131 (5) Noble gases released to the spent fbuel pool are r .leased through the Shield Building vent to the env;r)oanment.

(6) Inthe Reeaulatorv Guide 1.25 analvsis thec io~dine eaa inventorY Is eemposed Of iorFganic species (99,75-07) and organic species (0.25%17).

(7) in the Regulato.y Guide 1..25 analysis the spent fuel pool decotam.ination factorFs forF the inorganic and organic iodine arc 133 and 1, respectively.

(8) AllIodine escaping frolm the peol is exhausted to the envIronmnIt through chalrcFal f;ite*rs (9)

ABGTS filters and 902%fo inrancidine and 30-%for orFganic iodine for the purge air exhaust filters.

ENVIRONMENTAL CONSEQUENCES OF ACCIDENTS 15.5-23

WATTS BAR (10) No ro;dit is taken for noturol decay ;ithr du-e to holdup in the A^,xili;-y Bu~ilding OF ofter the activity has boon rsleoscd to thc atmosphers.

(11) The short teFrm (i.e., 0 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months ) atmosphecri dlu.tion fctors at the exclusion area boundory and loW pGpulotien zonRe given ;n Table 15A 2 orc used. Thee-thyoid dcsc utilizes CRP-30. G odinc do' e ccn'-,veSion factors.Docs aore based on the dose models prcsented in Appendix 15A.

The thyroid, gamma, and bcta doses for FHAs in the Auxiliary ond ReaDtor BuildingS are givcn in Tablc 15.5 23 for the excluslion rco boundory and low. population zone.

These doses ore less than 25-% of the 10CFRI00 11 limits of 300 rem to the thyroid, and 25 rem gmF*.o t* the whole bod. These dioses ore -olcu.ot-d by using RvisioRn 4 of the computer FEN'DOS'a*;58FE44

oed The ventilation function of the re.ctor. building purge vontilsting system. (RBPVS) is not a safety reloted function..Howevr, the filtrotion units ond ossocioted exhoust dutW.ork do providn a safety roloted filtrotion path feloWing a fuel hondling occ.ident prior to out~mo-tic closurWe of the ossocioted isolotion volves. The RBPVS contoins oir cle.nup utsre With prIfilt HEPA lrs, filters, ond 2 i-n h thick charcool odmorber. . This system is similor to the

.uxiliry bu.ilding gas treatm.nt system except that the o.tteris equipped with 4 inch thick chorcool odSrebrFS. Anytime fuel hondling operations or, being crcned On inside the pFrimarF, cntoinment, eithnmthe cnctainment is isoloted oe the rooctor bulilding purge filtrotion system is oporotional. The ossumptions listed oboveore, therefore, rppliogblo to a fuel hntdling occident nside p*ri*mlay *c*t*inM*Rt

..Xcept that the ossignod filter efficioncy is 90-0% for inr)gonic io~dine ond 30% for oa ione since n) relot.ie humidity is pr.vid.d.

.ontrol The rod i.t..n dose results of the Regulaton Guide 1/.25 fuel .ocidenthandling (F HA) imgiveninToble 15.5 23. For a FHA inside andtocrnment, no allswone hcthbont modt fo p.....b.. holdup OrFmin in the primarypg crntnmycntor soltion oell f the m-r,'

con~tainmenPt as a rosult of a high radiotion signal fromA the oiosi h etlto systems fE the ease whers contoinment pcnetrotions ore closod to the Au~xiliary Building. However, the Gontoinment purge filters ore creditod. For a FHA inside containment when containment penetrations and/er the annulus arc open to the uilingABCE paesthe G~tfiFetis iselated by ahigh radiation signal Auxliay fromF mneontors in the ventilation system and no credit is assumed forF the conRtaiment A ylar ulin BC of spaces,... **, the*T*,,

purge filters. -The rosult a FHA isde prim~ary eontainmont is well below the limits of 10 CFR 100.

The whole body, beta, ond thyroid doses to control roomF personnel fromA the rodiotion sourco.s *o-r discussed abovo . presnted in Table 15.5 23. The doses aFr calculoted by the CODRO.D computer co..de ,-7-. The gomma and beta doses a. o based on a.n.

tiome bur of a TPC fuel element, wheFers the thyroid dos. is based On a three times burned clement. This selection of sources produces higher doses. Parometcrc for thc control *ram analysis a r found iR Table 15.5 11 . The deoe to whole body is, b.low the GDC 19 lim.it of 5 rem f. control ro.m personnel, and the thyroid dose is below the l;imnit of 30 rem.

15.5-24 ENVIRONMENTAL CONSEQUENCES OFACCIDENTS

WATTS BAR Dose

. .u-tis - ;i TID 118442-3.1 cr uod to dtormino the dos.. D*o* convorsion-ftor,-,S in I'RP 30 - were -u-sd t, dctorminc thyroid dosos in plaGc of those feund in TD--14844-15.5.6.1 Fuel Handling Accident Based on Regulatory Guide 1.25 The parameters used for this analysis are listed in Table 15.5-20.

The bases for the Regulatory Guide 1.25 evaluation are:

(1) In the Regulatory Guide 1.25 analysis, the accident occurs 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months after plant shutdown. Radioactive decay of the fission product inventory during the interval between shutdown and placement of the first spent fuel assembly into the spent fuel pit is taken into account.

(2) In the Regulatory Guide 1.25 analysis damage is assumed for all rods in one assembly.

(3) The assembly damaged is the highest powered assembly in the core region to be discharged. The values for individual fission product inventories in the damaged assembly are calculated assuming full-power operation at the end of core life immediately preceding shutdown. Nuclear core characteristics used in the analysis are given in Table 15.5-21. A radial peaking factor of 1.65 is used.

(4) For the Regulatory Guide 1.25 analysis all of the gap activity in the damaged rods is released to the spent fuel pool and consists of 10% of the total noble gases and radioactive iodine inventory in the rods at the time of the accident with the following gap percentage exceptions which are based on NUREG/CR 5009 [241 as appropriate: 14% of the Kr-85, 5% of the Xe-133, 2% of the Xe-1 35, and 12% of the 1-131.

(5) Noble gases released in the containment are released through the Shield Building vent to the environment.

(6) In the Regulatory Guide 1.25 analysis the iodine gap inventory is composed of inorganic species (99.75%) and organic species (0.25%).

(7) A filter efficiency of 90% for inorganic iodine and 30% for organic iodine for the purge air exhaust filters is used since no relative humidity control is provided.

(8) No credit is taken for natural decay after the activity has been released to the atmosphere.

(9) The short-term (i.e., 0-2 hour) atmospheric dilution factors at the exclusion area boundary and low population zone given in Table 15A-2 are used. The thyroid dose utilizes ICRP-30 [251 iodine dose conversion factors. Doses are based on the dose models presented in Appendix 15A.

ENVIRONMENTAL CONSEQUENCES OF ACCIDENTS 15.5-25

WATTS BAR 15.5.6.2 Fuel Handling Accident Based on Regulatory Guide 1.183 The analysis of a postulated fuel handling accident in the Auxiliary Building refueling Area is based on Regulatory Guide 1.183. i.e., Alternate Source Terms (AST). The parameters used for this analysis are listed in Table 15.5-20.a.

The bases for evaluation are:

(1) In the Regulatory Guide 1.183 analysis, the accident occurs 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months after plant shutdown. Radioactive decay of the fission product inventory during the interval between shutdown and placement of the first spent fuel assembly into the spent fuel pit is taken into account.

(2) Inthe Regulatory Guide 1.183 analysis, damage was assumed for all rods in one assembly.

(3) The assembly damaged is the highest powered assembly in the core region to be discharged. The values for individual fission product inventories in.the damaged assembly are calculated assuming full-power operation at the end of core life immediately preceding shutdown. Nuclear core characteristics used inthe analysis are given in Table 15.5-21. A radial peaking factor of 1.65 is used.

(4) The Regulatory Guide 1.183 analysis assumes all of the gap activity in the damaged rods is released to the spent fuel pool and consists of 8% 1-131, 10% Kr-85, and 5% of other noble gases and other halogens.

(5) Noble gases released to the Auxiliary Building spent fuel pool are released through the Auxiliary Building vent to the environment.

(6) In the Regulatory Guide 1.183 analysis, the iodine gap inventory is composed of inorganic species (99.85%) and organic species (0.15%).

(7) In the Regulatory Guide 1.183 analysis, the overall inorganic and organic iodine spent fuel pool decontamination factor is 200.

(8) Inthe Regulatory Guide 1.183 analysis, all iodine escaping from the Auxiliary Building spent fuel pool is exhausted unfiltered through the Auxiliary Building vent.

(9) No credit is taken for the ABGTS or Containment Purge System Filters in the analysis.

(10) No credit is taken for natural decay either due to holdup in the Auxiliary Building or after the activity has been released to the atmosphere.

(11) The short-term (i.e., 0-2 hour) atmospheric dilution factors at the exclusion area boundary and low population zone given in Table 15A-2 are used. The thyroid dose utilizes ICRP-30 [25] iodine dose conversion factors. Doses are based on the dose models presented in Appendix 15A.

15.5-26 ENVIRONMENTAL CONSEQUENCES OF ACCIDENTS

WATTS BAR 15.5.6.3 Fuel Handling Accident Results The radiation dose results of the Regulatory dGide 1.25 with the containment closed fuel handling accident (FHA) are given in Table 15.5-23. For a FHA inside containment, no allowance has been made for possible holdup or mixing in the primary containment or isolation of the primary containment as a result of a high radiation signal from the monitors in the ventilation systems for the case where containment penetrations are closed to the Auxiliary Building. However, the containment purge filters are credited. Dose equations in TID-14844 [231 were used to determine the dose. Dose conversion factors in ICRP-30 [25] were used to determine thyroid doses in place of those found in TID-14844.

The ventilation function of the reactor building purge ventilating system (RBPVS) is not a safety-related function. However, the filtration units and associated exhaust ductwork do provide a safety-related filtration path following a fuel-handling accident prior to automatic closure of the associated isolation valves. The RBPVS contains air cleanup units with prefilters, HEPA filters, and 2-inch-thick charcoal adsorbers. This system is similar to the auxiliary building gas treatment system except that the latter is equipped with 4-inch-thick charcoal adsorbers. Anytime fuel handling operations are being carried on inside the primary containment, either the containment is isolated or the reactor building purge filtration system is operational. The assumptions listed above are, therefore, applicable to a fuel handling accident inside primary containment.

The thyroid, gamma, and beta doses for FHAs for the closed containment are given in Table 15.5-23 for the exclusion area boundary and low population zone. These doses are less than 25% of the 10CFR100.11 limits of 300 rem to the thyroid, and 25 rem gamma to the whole body. These doses are calculated by using Revision 5 of the computer code FENCDOSE [161.

The whole body, beta, and thyroid doses to control room personnel from the radiation sources discussed above are presented in Table 15.5-23. The doses are calculated by the COROD computer code [171. Parameters for the control room analysis are found in Table 15.5-14. The dose to whole body is below the 10 CFR 50 Appendix A, GDC 19 limit of 5 rem for control room personnel and the thyroid dose is below the limit of 30 rem.

The radiation dose results of the Regulatory Guide 1.183 fuel handling accident (FHA) are given in Table 15.5-23. Alternate source term (AST) described in RG 1.183 was selectively used to evaluate the FHA due to an event in the spent fuel pool located in the Auxiliary Building or in the containment when the equipment hatch or both doors in a personnel air lock are open. As Dart of this selective implementation of AST, the following assumptions are used in the analysis:

" The total effective dose equivalent (TEDE) acceptance criterion of 10 CFR 50.67(b)(2) replaces the previous whole body and thyroid dose guidelines of 10 CFR 100.11.

" The gap activity is revised to be consistent with that required by RG 1.183.

ENVIRONMENTAL CONSEQUENCES OF ACCIDENTS 15.5-27

WATTS BAR

" The decontamination factors were changed to be consistent with those required by RG. 1.183..

" No Auxiliary Building isolation is assumed.

2 No filtration of the release from the spent fuel pool to the environment by the ABGTS is assumed.

The evaluation for the FHA at the spent fuel pool is a bounding analysis for a dropped assembly in containment when the containment is open. The release point for the containment purge system is the Unit 2 shield building stack. The X/Qs are lower for this release point than for the normal auxiliary building exhaust. In addition, any release from the shield building stack would go through the purge system HEPA and Charcoal filter assemblies prior to release. Currently, when the purge lines isolate on high radiation, the auxiliary building also isolates and ABGTS is actuated. The release point for ABGTS is the shield building stacks and the releases are filtered through HEPA and Charcoal assemblies. Thus AST analysis for the FHA in the Auxiliary Building that considers no filtration is conservative and an acceptable as the basis for the containment open evaluation.

The thyroid, gamma, and beta doses for FHAs in the Auxiliary and the open containment are given in Table 15.5-23 for the exclusion area boundary and low population zone. These doses are less than 25% of the 10 CFR 100.11 limits of 300 rem to the thyroid, and 25 rem gamma to the whole body and less than the 10 CFR 50.67 limit of 25 rem TEDE. These doses are calculated by using Revision 5 of the computer code FENCDOSE [161.

The whole body, beta, and thyroid doses to control room personnel from the radiation sources discussed above are presented in Table 15.5-23. The doses are calculated by the COROD computer code [171. Parameters for the control room analysis are found in Table 15.5-14. The dose to whole body is below the 10 CFR 50 Appendix A, GDC 19 limit of 5 rem for control room personnel, and the thyroid dose is below the limit of 30 rem and the 10 CFR 50.67 limit of 5 rem TEDE.

15.5.7 Environmental Consequences of a Postulated Rod Ejection Accident This accident is bounded by the loss-of-coolant accident. See Section 15.5.3 for the loss-of-coolant accident.

REFERENCES (1) Styrikovich, M. A., Martynova, 0. I., Katkovska, K. YA., Dubrovski, I. YA.,

Smrinova, I. N., "Transfer of Iodine from Aqueous Solutions to Saturated Vapor," translated from Atomnaya Energiya, Vol. 17, No. 1, pp. 45-49, July 1964.

15.5-28 ENVIRONMENTAL CONSEQUENCES OFACCIDENTS

WATTS BAR (2) Regulatory Guides for Water Cooled Nuclear Power Plants, Regulatory Guide 1.24, "Assumptions Used for Evaluating the Potential Radiological Consequences of a Pressurized Water Reactor Gas Storage Tank Failure,"

Division of Reactor Standards, U.S. Atomic Energy Commission, March 23, 1972.

(3) Regulatory Guide 1.4, "Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss of Coolant Accident for Pressurized Water Reactors," Directorate of Regulatory Standards, U.S. Atomic Energy Commission, June 1974.

(4) D. D. Malinowski, "Iodine Removal in the Ice Condenser System,"

WCAP-7426, April 1970.

(5) NAA-SR 10100, Conventional Buildings for Reactor Containment.

(6) ORNL-NSIC-4, Behavior of Iodine in Reactor Containment Systems, February 1965.

(7) Branch Technical Position CSB 6-2, "Control of Combustible Gas Concentrations in Containment Following a Loss-of-Coolant Accident."

(8) Ramsdell, J. V. Jr. and C. A. Simonen, "Atmospheric Relative Concentrations in Building Wakes." Prepared by Pacific Northwest laboratory fo the U. S.

Nuclear Regulatory Commission, PNL-10521, NUREG/CR-6331, Revision1, May 1997.

(9) K. G. Murphy and Dr. K. M. Campe "Nuclear Power Plant Control Room Ventilation System Design for Meeting General Criterion 19," 13th AEC Air Cleaning Conference, August 1974.

(10) Deleted by Amendment 80.

(11) Regulatory Guides for Water Cooled Nuclear Power Plants, Regulatory Guide 1.25, "Assumptions Used for Evaluating the Potential Radiological Consequences of a Fuel Handling Accident in the Fuel Handling and Storage Facility for Boiling and Pressurized Water Reactors," Division of Reactor Standards, U.S. Atomic Energy Commission, March 23, 1972.

(12) Regulatory Guides for Water Cooled Nuclear Power Plants, Regulatory Guide 1.77, "Assumptions Used for Evaluating a Control Rod Ejection Accident for Pressurized Water Reactors," Directorate of Regulatory Standards, U.S. Atomic Energy Commission, May 1974.

(13) D. B. Risher, Jr., "An Evaluation of the Rod Ejection Accident in Westinghouse Pressurized Water Reactors Using Spatial Kinetics Methods,"

WCAP-7588, Revision 1, December 1971.

ENVIRONMENTAL CONSEQUENCES OF ACCIDENTS 15.5-29

WATTS BAR (14) ANSI/ANS-18.1-1984, "Radioactive Source Terms for Normal Operations of Light Water Reactors," December 31, 1984.

(15) WCAP-7664, Revision 1, "Radiation Analysis Design Manual-4 Loop Plant,"

RIMS Number NEB 810126 316, October 1972.

(16) Computer Code FENCDOSE, Code I.D. 262358.

(17) Computer Code COROD, Code I.D. 262347.

(18) Regulatory Guide 1. 183 RO, Alternate Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, US Nuclear Reaulatory Commission, July 2000.Net u-sed (19) Not used (20) NRC Safety Evaluation for Watts Bar Nuclear Plant Unit 1, Amendment 38, for Steam Generator Tubing Voltage Based Alternate Repair Criteria for Outside Diameter Stress Corrosion Cracking (ODSCC) dated February 26, 2002.

(21) NRC Generic Letter 95-05, "Voltage-Based Repair Criteria for Westinghouse Steam Generator Tubes Affected by Outside Diameter Stress Corrosion Cracking", dated August 3, 1995.

(22) TVA Letters to NRC "Technical Specification Change No. WBN-TS-99-014 -

Steam Generator Alternate Repair Criteria for Axial Outside Diameter Stress Corrosion Cracking (ODSCC)," dated April 10, 2000, September 18, 2000, August 22, 2001, November 8, 2001 and January 15, 2002.

(23) J.J. Dinunno, et, al "Calculation of Distance Factors for Power and Test Reactor Sites", TIC-14844, March 1962.

(24) NUREG/CR-5009, "Assessment of the Use of Extended Burnup Fuel in Light Water Power Reactors," February 1988.

(25) International Commission on Radiation Protection (ICRP) Publication 30, Limits for Intakes of Radionuclides by Workers," 1979.

15.5-30 ENVIRONMENTAL CONSEQUENCES OFACCIDENTS

fl Table 15.5-1 Parameters Used In Loss Of A. C. Power Analyses Realistic Analysis Conservative Analysis Core thermal power 3565 MWt 3565 MWt C6, Steam generator tube leak rate 1 gpm 1.0 gpm 00 prior to and during accident C,? Fuel defects ANSI/ANS 18.1 - 1984 Technical Specification limit of 0.1 pCi/gm 1-131 dose equivalent 0 Iodine partition factor in 0.01 0.01 01 steam generator prior to and r¢l during accident Blowdown rate per steam 25 gpm 25 gpm generator prior to accident Duration of plant cooldown by secondary system after 8 hrs 8 hrs accident Steam release from 4 steam 444,875 Ibm (0-2 hrs) 444,875 lbs (0-2 hrs) generators 903,530 Ibm (2-8 hrs) 903,530 lbs (2-8 hrs)

Meteorology See Tables 15A-2 & 15.5-14 See Tables 15A-2 & 15.5-14 CII (Ii

WATTS BAR Table 15.5-2 Doses From Loss Of A/C Power Conservative Analysis (rem) 2HR EAB 30 DAY LPZ CONTROL ROOM Gamma 7.45E-04 4.18E-04 2.102-.4-2E-04 I Beta 4.48E-04 2.52E-04 2.522-.-.-E-03 I Thyroid - ICRP-30 4.57E-02 2.57E-02 2.092-44E-02 2HR EAB 30 DAY LPZ CONTROL ROOM Realistic Analysis (rem)

Gamma 1.80E-08 1.01E-08 5.05&.44-E-09 I Beta 1.66E-05 9.29E-06 1.794-.-91E-04 Thyroid - ICRP-30 1.1OE-06 6.18E-07 5.03-5-09E-07 15.5-32 ENVIRONMENTAL CONSEQUENCES OFACCIDENTS

WATTS BAR Table 15.5-3 Parameters Used In Waste Gas Decay Tank Rupture Analyses Regulatory Guide Realistic Analysis 1.24 Analysis Core thermal power 3565 MWt 3565 MWt Plant load factor 1.0 1.0 Fuel defects ANSI/ANS-1 8.1, 1984 1%

Activity released from GWPS (1) See Table 15.5-4 Time of accident After Tank Fill At end of equilibrium core cycle Meteorology See Table 15.5-14 and Table 15A-2 See Table 15.5-14 and Table 15A-2 (1)Activity based on maximum concentrations of each isotope and actual plant flow rates of the GWPS.

ENVIRONMENTAL CONSEQUENCES OF ACCIDENTS 15,5-33

WATTS BAR Table 15.5-4 Waste Gas Decay Tank Inventory (One Unit) (Regulatory Guide 1.24 Analysis)

Activity Isotope (Curies)

Xe-131m 8.9 x 102 Xe-133 6.8 x 104 Xe-133m 1.0 x 103 Xe-135 9.4 x 102 Xe-135m 4.8 x 101 Xe-137 2.7 x 10-1 Xe-138 3.2 Kr-83m 1.7 x 101 Kr-85 4.2 x 103 Kr-85m 1.3 x 102 Kr-87 2.9 x 101 Kr-88 1.6 x 102 Kr-89 1.0 x 10-1 1-131 4.8 x 10-2 1-132 1-133 3.3 x 10-2 1-134 1-135 1.2 x 102 15.5-34 ENVIRONMENTAL CONSEQUENCES OF ACCIDENTS

WATTS BAR Table 15.5-5 Doses From Gas Decay Tank Rupture Regulatory Guide 1.24 Analysis 2HR EAB 30 DAY LPZ CONTROL ROOM (rem)

Gamma 5.96E-01 1.67E-01 8.43E-01 Beta 1.61E+00 4.51E-01 7.28E+00 Thyroid - ICRP-30 1.29E-02 3.60E-03 6.99E-03 2HR EAB 30 DAY LPZ CONTROL ROOM Realistic Analysis (rem)

Gamma 2.88E-02 8.05E-03 3.81 E-02 Beta 1.10E-01 3.08E-02 5.01 E-01 Thyroid - ICRP-30 1.21 E-02 3.37E-03 6.50E-03 ENVIRONMENTAL CONSEQUENCES OF ACCIDENTS 15.5-35

WATTS BAR Table 15.5-6 Parameters Used In LOCA Analysis (Page 1 of 2)

Regulatory Guide 1.4 Analysis Core thermal power 3565 MWt 3

Primary containment free volume 1.27 x 106 ft 5 3 Annulus free volume 3.75 x 10 ft Primary containment deck (air return) fan flow rate 40,000 cfm Number of deck (containment air return fans) fans assumed 1 of 2 operating Activity released to primary containment and available for release noble gases 100% of core inventory iodines 25% of core inventory Form of iodine activity in primary containment available for release elemental iodine 91%

methyl iodine 4%

particulate iodine 5%

Ice condenser removal efficiency for elemental and See Table 15.5-7 particulate iodine Primary containment leak rate (volume percent) 0.25% per day (0-24 hours) 0.125% per day (1-30 days)

Percent of primary containment leakage to auxiliary building 25%

ABGTS filter efficiencies elemental iodine 99%

methyl iodine 99%

particulate iodine 99%

Delay time of activity in auxiliary building before ABGTS None operation Delay time before filtration credit is taken for the ABGTS 4 min Mean holdup time in auxiliary building after initial 4 minutes 0.3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months ABGTS flow rate 9000 cfm 15.5-36 ENVIRONMENTAL CONSEQUENCES OFACCIDENTS

WATTS BAR Table 15.5-6 Parameters Used In LOCA Analysis (Page 2 of 2)

Leakage from Auxiliary Building to ABGTS downstream 27.88 cfm HVAC (bypass of filters)

Leakage from ABGTS HVAC into Auxiliary Building 8.87 cfm Leakage from Auxiliary Building into EGTS downstream 10.7 cfm HVAC (bypass of filters)

Leakage from Auxiliary Building to environment due to single 9900 cfm (for 4 minutes) failure of ABGTS (from 30 minutes to 34 minutes post-LOCA)

Percent of primary containment leakage to annulus 75%

Emergency gas treatment system flow rates See Table 15.5-8 and Table 15.5-8A Percent of annulus free volume available for mixing of 50%

recirculated activity Number of emergency gas treatment system air handling 1 of 2 units operating Emergency gas treatment system filter efficiencies elemental iodine 99%

methyl iodine 99%

particulate iodine 99%

Shield building mixing model (see Section 15.5.3) 50% mixing Meteorology See Table 15.5-14 and Table 15A-2_ I ENVIRONMENTAL CONSEQUENCES OF ACCIDENTS 15.5-37

WATTS BAR Table 15.5-7 Ice Condenser Elemental And Particulate Iodine Removal Efficiency(1,2)

Time Interval Iodine Removal Post LOCA (Hours) Efficiency 0.0 to 0.156 0.96 0.156 to 0.267 0.76 0.267 to 0.323 0.73 0.323 to 0.489 0.71 0.489 to 0.615 0.60 0.615 to 0.768 0.58 0.768 to 0.824 0.40 0.824 to 720 0.0 (1) The ice condenser removal efficiencies given in the above table are used for the Regulatory Guide 1.4 analysis. The inlet steam/air mixture coming into the ice condenser is greater than 90% steam by volume initially due to the delaying of the operation of the containment deck fans. Without the delay of operation of the deck fans, the amount of steam by volume in the inlet mixture initially would be much lower and the ice condenser iodine removal efficiencies would be reduced.

(2) The ice bed iodine removal efficiency, 01, has been computed on a time dependent basis and is shown in Table 15.5-7. Note that the information presented in Table 15.5-7 has been revised by Westinghouse letter WAT-D-1 0954. The revised efficiency information is associated with the WCAP-1 5699, Revision 1 analysis for reduced ice weight. A comparison of the information presented in Table 15.5-7 and the revised information contained in WAT-D-1 0954 shows that the information in Table 15.5-7 is conservative. Analyses supporting the plant design basis acknowledge the revised efficiency information but shall utilize the information presented in Table.15.5-7.

15.5-38 ENVIRONMENTAL CONSEQUENCES OFACCIDENTS

WATTS BAR Table 15.5-8 EMERGENCY GAS TREATMENT SYSTEM FLOW RATES (Sheet 1 of 1)

Time Interval Time Interval Recirculation Rate Exhaust Rate (sec) (hours) (cfm) (cfh) (cfm) (cfh) 0-30 0-0.0083 0.00 0.OOE+00 0.00 0.OOE+00 30-39 0.0083-0.0108 3600.00 2.16E+05 0.00 0.OOE+00 39-40 0.0108-0.0111 3286.62 1.97E+05 313.38 1.88E+04 40-41 0.0111-0.0114 2352.31 1.41E+05 1247.69 7.49E+04 41-42 0.0114-0.0117 1304.79 7.83E+04 2295.21 1.38E+05 42-43 0.0117-0.0119 362.60 2.18E+04 3237.40 1.94E+05 43-190 0.0119-0.0528 0.00 0.OOE+00 3600.00 2.16E+05 190-191 0.0528-0.0531 537.28 3.22E+04 3062.72 1.84E+05 191-192 0.0531-0.0533 733.23 4.40E+04 2866.77 1.72E+05 192-193 0.0533-0.0536 735.14 4.41 E+04 2864.86 1.72E+05 193-194 0.0536-0.0539 737.51 4.43E+04 2862.49 1.72E+05 194-199 0.0539-0.0553 745.23 4.47E+04 2854.77 1.71E+05 199-207 0.0553-0.0575 764.12 4.58E+04 2835.89 1.70E+05 207-215 0.0575-0.0597 790.80 4.74E+04 2809.20 1.69E+05 215-225 0.0597-0.0625 825.45 4.95E+04 2774.56 1.66E+05 225-245 0.0625-0.0681 892.72 5.36E+04 2707.29 1.62E+05 245-265 0.0681-0.0736 992.80 5.96E+04 2607.20 1.56E+05 265-285 0.0736-0.0792 1102.40 6.61 E+04 2497.61 1.50E+05 285-305 0.0792-0.0847 1217.05 7.30E+04 2382.95 1.43E+05 305-446 0.0847-0.1239 1664.05 9.98E+04 1935.96 1.16E+05 446-601 0.1239-0.1669 2356.72 1.41 E+05 1243.29 7.46E+04 601-602 0.1669-0.1672 2661.35 1.60E+05 938.65 5.63E+04 602-1700 0.1672-0.4722 3600.00 2.16E+05 0.00 0.OOE+00 1700-1701 0.4722-0.4725 3508.13 2.10E+05 91.87 5.51 E+03 1701-1702 0.4725-0.4728 3423.44 2.05E+05 176.56 1.06E+04 1702-1703 0.4728-0.4731 3410.73 2.05E+05 189.27 1.14E+04 1703-1704 0.4731-0.4733 3408.66 2.05E+05 191.34 1.15E+04 1704-1705 0.4733-0.4736 3408.17 2.04E+05 191.83 1.15E+04 1705-1706 0.4736-0.4739 3407.91 2.04E+05 192.09 1.15E+04 1706-1855 0.4739-0.5153 3395.23 2.04E+05 204.77 1.23E+04 1855-2100 0.5153-0.5833 3372.37 2.02E+05 227.64 1.37E+04 2100-30 days* 0.5833-720 3350.00 2.01 E+05 250.00 1.50E+04

Required to maintain annulus pressure when assuming 250 cfm annulus inleakage ENVIRONMENTAL CONSEQUENCES OFACCIDENTS 15.5-39

Table 15.5-8A Emergency Gas Treatment System Flow Rates (Unit 2)

Time Interval Time Interval Recirculation Rate Exhaust Rate

~~1 (sec) (sec) (hours) (hours) (cfm) (cfh) (cfm) (cfh) Co 0 30 0 0.0083 0 0.OOE+00 0 0.OOE+00 30 39 0.0083 0.0108 7200 4.32E+05 0 0.OOE+00 39 40 0.0108 0.0111 6573.24 3.94E+05 626.76 3.76E+04 40 41 0.0111 0.0114 4704.62 2.82E+05 2495.38 1.50E+05 41 42 0.0114 0.0117 2609.58 1.57E+05 4590.42 2.75E+05 42 43 0.0117 0.0119 725.2 4.35E+04 6474.8 3.88E+05 43 71 0.0119 0.0197 0 0.OOE+00 7200 4.32E+05 71 78 0.0197 0.0217 0 0.OOE+00 7200 4.32E+05 78 79 0.0217 0.0219 1062 6.37E+04 6138 3.68E+05 79 80 0.0219 0.0222 4775 2.87E+05 2425 1.46E+05 80 102 0.0222 0.0283 4337 2.60E+05 2863 1.72E+05 102 132 0.0283 0.0367 4188 2.51E+05 3012 1.81E+05 132 165 0.0367 0.0458 3922 2.35E+05 3278 1.97E+05 165 170 0.0458 0.0472 3762 2.26E+05 3438 2.06E+05 170 210. 0.0472 0.0583 3719 2.23E+05 3481 2.09E+05 210 307 0.0583 0.0853 3760 2.26E+05 3440 2.06E+05 307 498 0.0853 0.1383 4050 2.43E+05 3150 1.89E+05 498 602 0.1383 0.1672 4797 2.88E+05 2403 1.44E+05 602 603 0.1672 0.1675 5232 3.14E+05 1968 1.18E+05 603 850 0.1675 0.2361 5137 3.08E+05 1432 8.59E+04 850 1100 0.2361 0.3056 5237 3.14E+05 1332 7.99E+04 1100 1350 0.3056 0.3750 5337 3.20E+05 1232 7.39E+04 1350 1600 0.3750 0.4444 5437 3.26E+05 1132 6.79E+04 1600 1850 0.4444 0.5139 5537 3.32E+05 1032 6.19E+04 1850 2100 0.5139 0.5833 5637 3.38E+05 932 5.59E+04 2100 3600* 0.5833 1.0000 5737 3.44E+05 832 4.99E+04 3600* 30 1.0000 30 days 3455 2.07E+05 604 3.62E+04 days

Reflects operator action to place one EGTS fan in standby mode at 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

WATTS BAR Table 15.5-9 DOSES FROM LOSS-OF-COOLANT ACCIDENT (rem) 2Hr EAB 30 Day LPZ Control Room Gamma 2.12 2.18 1.05 Beta 1.25 2.61 9.10 Thyroid - ICRP - 30 40.4 14.33 3.75 Breakdown of Control Room Personnel Dose (rem) Airborne Shine Ingress/Egress Total

+ 4 Gamma 1.02 0.005 0.027 1.05 Beta 9.04 0.000 0.060 9.10 Thyroid - ICRP - 30 3.66 0.000 0.090 3.75 ENVIRONMENTAL CONSEQUENCES OF ACCIDENTS 15.5-41

WATTS BAR Table 15.5-10 Deleted by Amendment 80 15.5-42 ENVIRONMENTAL CONSEQUENCES OFACCIDENTS

WATTS BAR Table 15.5-11 Deleted by Amendment 80 ENVIRONMENTAL CONSEQUENCES OFACCIDENTS 15.5-43

Table 15.5-12 PARAMETERS USED IN ANALYSIS OF RECIRCULATION LOOP LEAKAGE FOLLOWING A LOCA Regulatory Guide 1.4 Analysis W, Core thermal power 3565 MWt 4 3 Recirculation sump water volume 9.63 x 10 ft Activity mixed with recirculation loop water Noble gases 0.0 lodines 50% of core inventory Tritium 97% to sump (water)

Leakage of ECCS equipment outside containment See Table 6.3-6 Iodine partition factor for leakage 0.1 ABGTS filter efficiencies elemental iodine 99%

methyl iodine 99%

particulate iodine 99%

Meteorology See Table 15.5-14.and Table 15A-2

WATTS BAR Table 15.5-13 Doses From Recirculation Loop Leakage Following A LOCA (rem) 2HR EAB 30 Day LPZ Control Room Gamma 4.14E-03 2.28E-02 1.51E-03 Beta 1.36E-03 8.54E-02 1.62E-02 Thyroid - ICRP - 30 1.40E-03 1.53E-01 3.69E-02 ENVIRONMENTAL CONSEQUENCES OF ACCIDENTS 15.5-45

WATTS BAR WBNP-107 WBNP-1 07 WATTS BAR Table 15.5-14 Atmospheric Dilution Factors At The Control Building 3

DILUTION FACTOR (seclm )

SGTR/MSLB/Loss of AC 0-2 Time Period (hr) LOCAIFHA 1.09E-03 Power LOSS OF NJC POWER 3.85E 03 WGDT 2.56E-03 I

2.*6 E 0;32.59E-03 2-8 9.44E-04 3.22E 03 1.17E 03N/.A I 2.165 092.12E-03 8-24 1.56E-04* N/ANA 7.26E-04N/A 24-96 1.16E-04"- N/AN-A .2-4-E-04N/A 96-720 9.59E-05*** N/AN-A 4.9 N/_4N/A

. GENERAL CONTROL ROOM PARAMETERS Volume 257,198 cu ft Makeup/pr essurization flow 711 cfm Recirculati )n flow 2889 cfm Unfiltered irntake 51 cfm Filter efficie .ncy 95% first pass 70% second pass 0% for noble gases, Tritium Isolation tin Te. T 40 seconds Occupancy factors:

0-24 hr 100%

1-4 days 60%

4-30 days 40%

1. All FHA releases are within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . Thus, only the 0-2 hr X/Q is applicable for the FHA. I

Calculated value for Ul Shield Blda Vent to East MCR Intake 1.26E-04 I Calculated value for Ul Shield 9.53E-05 Blda Vent to East MCR Intake Calculated value for Ul Shield 8.07E-05 I

Blda Vent to East MCR Intake 15.5-46 ENVIRONMENTAL CONSEQUENCES OFACCIDENTS

WATTS BAR Table 15.5-15 Deleted by Amendment 97 ENVIRONMENTAL CONSEQUENCES OF ACCIDENTS 15.5-47

WATTS BAR Table 15.5-16 Parameters Used In Steam Line Break Analysis Analysis Value Steam Generator tube leak rate Faulted Steam Generator 1.0 gpm Per Intact Steam Generator 150 gpd Iodine Partition Factor Faulted Steam Generator 1 Intact Steam Generator 100 RCS Letdown flow rate 124.39 gpm Steam Releases Faulted Steam Generator (0-30 minutes) 96,100 Ibm Three Intact Steam Generators (0-2 hrs) 433,079 Ibm Three Intact Steam Generators (2-8 hrs) 870,754 Ibm 15.5-48 ENVIRONMENTAL CONSEQUENCES OF ACCIDENTS

WATTS BAR Table 15.5-17 Doses From Main Steam Line Break SRP- SRP G'-d4Ree- 30-Day4P*- Guidance 1 gpm Primary- GOA4rz Recm fer-GDG- fm-)SRP for to-Secondary .peFat.F Lieflits- Guidance for Control Room2- IOCFR100 Leakage -re*-) --eff-+30- 10CFR100 HOu EAB (Site Limits (ARCON-96 x/Q) 2 HR EAB Day LPZ Limits (rem) be..d.. .yFe.), (rem)

Pre-Accident Initiated-ledine Spike Case (140466 pCi/gm maximum peaksteady state) I Gamma- 8.10E 032.74E-02 1.11E-025 254-,25E--04 4.32E-0314E 01 2-5 Beta, 6.52E 028.80E-03 4.20E-0330 3003.02E 02 3.96E-022.54E-02 30 Thyroid - 10.4E+002.41E+00 1.21E+0030 3004.78E+00 7.38E+003..9.... 30 ICRP-Pfe-Accident Initiatedled4*e, Spike Case (0.26544 pCi/gm steady statee.= peak) I Gamma- 413-E--31.04E-01 61.25E-01 1.11E 022.5 2.74E 028.00E-03 2-5 Beta+ 4-.OE--02.54E-02 393.02E-02 4.20E 0330 8.80E 036.44E-02 300 Thyroid - 7.44E+,3.09E+00 WO4.78E+00 1.21-E+0030 2.44 E4-01.03E+01 300 ICRP-30_Re.. .atieR.

(1GRP4 + ______ 11 ENVIRONMENTAL CONSEQUENCES OF ACCIDENTS 15.5-49

WATTS BAR Table 15.5-18 Parameters Used In Steam Generator Tube Rupture Analysis Primary Side Activity Technical Specification Limit Secondary Side Activity ANSI/ANS-18.1-1984 (Expected levels, 150 gpd/SG)

Iodine Spiking Factor Case 1: Accident initiated spike of 500 times equilibrium iodine concentration Case 2: Pre-accident spike of 14 pCi/gm 1-131 dose equivalent Iodine Partition Factor 100 Secondary Side Mass Release (Ruptured Steam Generator) 0-2 hours 103,300 Ibm 2 - 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 32,800 Ibm Secondary Side Mass Release (Intact Steam Generator) 0 - 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 492,100 Ibm 2 - 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months 900,200 Ibm Primary Coolant Mass Release (Total) 0 - 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 191,400 Ibm Primary Coolant Mass Release (Flashed) 0 - 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 10,077.2 Ibm AtmozphOric diffuzion cccfficcnts for control room OGpemteF desesMeteorology 2.15 x ,-'-* ag.See Table 15A-2 and 15.5-14 I 15.5-50 ENVIRONMENTAL CONSEQUENCES OFACCIDENTS

WATTS BAR Table 15.5-19 Doses From Steam Generator Tube Rupture Pre-Accident Initiated Spike Case (14 pCi/gm maximum peak)

(rem) 2 HR EAB 30 DAY LPZ CONTROL ROOM Gamma 3.78E-01 1.11E-01 6&-296.22E-02 Beta 2.26E-01 6.92E-02 7--77.01E-01 Thyroid - ICRP-30 1.39E+01 3.79E+00 4-241.23E+01 Accident Initiated Iodine Spike Case (0.265 pCi/gm steady state)

(rem) 2 HR EAB 30 DAY LPZ CONTROL ROOM Gamma 5.46E-01 1.60E-01 &765.71E-02 Beta 2.51 E-01 7.73E-02 &-_796.64E-01 Thyroid - ICRP-30 7.19E+00 2.12E+00 2-032.01E+00 ENVIRONMENTAL CONSEQUENCES OF ACCIDENTS 15.5-51

WATTS BAR Table 15.5-20 Parameters Used In Fuel Handling Accident Analysis Regulatory Guide 1.25 Analysis Time between plant shutdown and accident 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months Damage to fuel assembly All rods ruptured Fuel assembly activity Highest powered fuel assembly in core region discharged Activity release to spent fuel pool Gap activity in ruptured rods(1 )

Radial peaking factor 1.65

Form of iodine activity released to spcnt fucl pool I elemental iodine 99.75%

methyl iodine 0.25%

Dcccntamination fecter in spent fucie peE4 elemental iodine 4-33 Rqethyliedie- 4 4

Filter efficiencies in auxiliary building PrBGTSk2 RBPVS(23) elemental iodine 90%

990- 30%

methyl iodine Amount of mixing of activity in Auxiliary Building None Meteorology See Table 15.5-14 and Table 15A-2 (1) 10% of the total radioactive iodine except for 12% of 1-131 and 10% of total noble gases, except for 14% for Kr-85, 5% for Xe-1 33 and 2% for Xe-1 35 in the damaged rods at the time of the accident.

(2) Auxiliary Building Gas T..atmcnt SysteRc (32) Reactor Building Purge Ventilation System 15.5-52 ENVIRONMENTAL CONSEQUENCES OFACCIDENTS

WATTS BAR Table 15.5-20a Parameters Used In Fuel Handling Accident Analysis Regulatory Guide 1.183 Analysis Time between plant shutdown and accident 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months Damage to fuel assembly All rods ruptured Fuel assembly activity Highest powered fuel assembly in core region discharged Activity release to spent fuel pool Gar activity in ruptured rodsl-'W Radial peaking factor 1.65 Form of iodine activity released to soent fuel nool elemental iodine 99.85%(AST) methyl iodine 0.15%(AST)

Decontamination factor in spent fuel pool AST Overall=200 Filter efficiencies No credit taken Amount of mixing of activity in Auxiliary Building None Meteorology See Table 15.5-14 and Tablel5A-2 M 8% 1-131, 10% Kr-85, and 5% other gasses and other halogens.

I ENVIRONMENTAL CONSEQUENCES OF ACCIDENTS 15.5-53

WATTS BAR Table 15.5-21 Nuclear Characteristics Of Highest Rated Discharged Assembly Used In The Analysis Core thermal power 3565 MWt Number of assemblies 193 Fuel rods per assembly 264 Core average assembly power 18.47 MWt Discharged Assembly Radial peak to average ratio 1.65 15.5-54 ENVIRONMENTAL CONSEQUENCES OFACCIDENTS

WATTS BAR Table 15.5-22 Deleted by Amendment 80 ENVIRONMENTAL CONSEQUENCES OFACCIDENTS 15.5-55

WATTS BAR Table 15.5-23 Doses From A Fuel Handling Accident Regulator'; Guidc 1.25 A*..y.ic Drccz Fr...m A Fuel Hardling A^;idcnt (FHA) (rem)

Doses from Fuel Handling Accident Regulatory Guide 1.183 Analyses I.

FHA in Auxiliary Building (rem) or In Containment - Containment Open (rem) 2 HR EAB 30 DAY LPZ CONTROL ROOM Gamma 3.994 E 01-4.29E-0 1 9.2-7.8E-21 .20E-01 4*.93E--05.86E-01 Beta 4.177E4.001 .19E+00 2.7-4E 013.33E-01 4.6B*004.68E+00 Thyroid - ICRP-30 1.577E4-005.51 E+01 3.63E-O011.54E+01 44640E+901.32E+01 TEDE 2.38E+00 6.66E-01 1.02E-00 Doses from Fuel Handlinq Accident Regulatory Guide 1.25 Analyses FHA in Reactor Building., Containment Closed (rem)

(r-era) 2 HR EAB 30 DAY LPZ CONTROL ROOM Gamma 4402B-014.31 E-01 2.67:7'92.72E-01 Beta 1.1482E+~001 .24E+00 2.746E-043.48E-0 1 2207-E402 .25 E+00 Thyroid - ICRP-30 3942E-4.OO4.1 5E+01 9.468Em+01 .16E+01 6.209E+996.81 E+00 A . .

Ou A R mearf8F KU1191RA IAI*#L, g -ARtam nmew ijeRelFat eAS f% -- +^ Atj-2!jagIIIMI ýnmý!Am" 2-HR-AB 30 DAY LPZ ONTROL ROOM Gamma 4.937E 01 9.378E 02 5.012E01G- I Beta 1.189E+00 2.761 E0 1 4.4 56E+,00 ThYF~id ICRP 30 ~A3.11E+0 6.436E=+900 15.5-56 ENVIRONMENTAL CONSEQUENCES OFACCIDENTS

WATTS BAR Table 15.5-24 Deleted by Amendment 80 ENVIRONMENTAL CONSEQUENCES OF ACCIDENTS 15.5-57

WATTS BAR Table 15.5-25 Deleted by Amendment 80 15.5-58 ENVIRONMENTAL CONSEQUENCESOFACCIDENTS

Enclosure 3 WBN Unit 2 - Revised FSAR Section 15.5 E3-1

WATTS BAR 15.5 ENVIRONMENTAL CONSEQUENCES OF ACCIDENTS 15.5.1 Environmental Consequences of a Postulated Loss of AC Power to the Plant Auxiliaries The postulated accidents involving release of steam from the secondary system will not result in a release of radioactivity unless there is leakage from the reactor coolant system (RCS) to the secondary system in the steam generator. A conservative analysis of the potential offsite doses resulting from this accident is presented with steam generator leakage as a parameter. This analysis incorporates assumptions of a Technical Specification limit of 0.1 pCi/gm 1-131 dose equivalent, and a realistic source term. Parameters used in both the realistic and conservative analyses are listed in Table 15.5-1.

The realistic assumptions that determine the equilibrium concentrations of isotopes in the secondary system are as follows:

(1) Primary coolant activity is associated with 0.125% defective fuel and is given in Table 11.1-7.

(2) The iodine partition factor in the steam generators is:

amount of iodine/unit mass steam 0.01 amount of iodine/unit mass liquid (3) No noble gas is dissolved or contained in the steam generator water, i.e., all noble gas leaked to the secondary system is continuously released with steam from the steam generators through the condenser off gas system.

(4) The 0-2 and 2-8 hour atmospheric dilution factors given in Appendix 15A and Table 15.5-14; the 0-8 hour breathing rate of 3.47 x 10-4 m3/sec are applicable. Doses are based on the dose models in Appendix 15A.

(5) Primary and Secondary side source terms are based on ANSI/ANS-1 8.1-1984.

Assumptions used for the conservative analysis are the same as the realistic assumptions except the Secondary side source terms at the Technical Specification limit of 0.1 pCi/gm 1-131 dose equivalent are assumed.

The steam releases to the atmosphere for the loss of AC power are in Table 15.5-1.

The gamma, beta, and thyroid doses for the loss of AC power to the plant auxiliaries at the exclusion area boundary and low population zone are in Table 15.5-2 for the realistic and conservative analyses. These doses are calculated by the FENCDOSE ENVIRONMENTAL CONSEQUENCES OF ACCIDENTS 15.5-1

WATTS BAR computer code[1 61 . The doses for this accident are less than 25 rem whole body, 300 rem beta and 300 rem thyroid. This is well within the limits as defined in 10 CFR 100.

The whole body, beta, and thyroid doses to control room personnel from the radiation sources discussed above are presented in Table 15.5-2. The doses are calculated by the COROD computer code [171 ] Parameters for the control room analysis are found in Table 15.5-14. The dose to whole body is below the GDC 19 limit of 5 rem for control room personnel, and thyroid dose is below the limit of 30 rem.

Dose equations in TID-14844 [23] were used to determine the dose. Dose conversion factors in ICRP-30 [25] were used to determine thyroid doses in place of those found in TID-14844.

15.5.2 Environmental Consequences of a Postulated Waste Gas Decay Tank Rupture Two analyses of the postulated waste gas decay tank rupture are performed:

(1) a realistic analysis, and (2) an analysis based on Regulatory Guide 1.24 (Reference 2). The parameters used for each of these analyses are listed in Table 15.5-3.

The assumptions for the Regulatory Guide analysis are:

(1) The reactor has been operating at full power with 1% defective fuel for the RG 1.24 analysis.

(2) The maximum content of the decay tank assumed to fail is used for the purpose of computing the noble gas inventory in the tank. Radiological decay is taken into account in the computation only for the minimum time period required to transfer the gases from the reactor coolant system to the decay tank. For the Regulatory Guide 1.24 analysis, noble gas and iodine inventories of the tank are given in Table 15.5-4. For the realistic analysis, source terms are based on ANSI/ANS-1 8.1-1984 methodology 1 1 41 .

(3) The tank rupture is assumed to occur immediately upon completion of the waste gas transfer, releasing the entire contents of the tank through the Auxiliary Building vent to the outside atmosphere. The assumption of the release of the noble gas inventory from only a single tank is based on the fact that all gas decay tanks will be isolated from each other whenever they are in use.

(4) The short-term (i.e., 0-2 hour) dilution factor at the exclusion area boundary given in Appendix 15A is used to evaluate the doses from the released activity. Doses are based on the dose models presented in Appendix 15A.

The gamma, beta, and thyroid doses for the gas decay tank rupture at the exclusion area boundary and low population zone are given in Table 15.5-5 for both the realistic and Regulatory Guide 1.24 analyses.

15.5-2 ENVIRONMENTAL CONSEQUENCES OFACCIDENTS

WATTS BAR (5) The whole body, beta, and thyroid doses to control room personnel from the radiation sources discussed above are presented in Table 15.5-5. The doses are calculated by the COROD computer code [17]. Parameters for the control room analysis are found in Table 15.5-14. The dose to whole body is below the GDC 19 limit of 5 rem for control room personnel, and thyroid dose is below the limit of 30 rem.

Dose equations in TID-14844 [23] were used to determine the dose. Dose conversion factors in ICRP-30 [25] were used to determine thyroid doses in place of those found in TID-14844.

15.5.3 Environmental Consequences of a Postulated Loss of Coolant Accident The results of the analysis presented in this section demonstrate that the amounts of radioactivity released to the environment in the event of a loss-of-coolant accident do not result in doses which exceed the reference values specified in a 10 CFR 100.

The analysis is based on Regulatory Guide 1.4131. The parameters used for this analysis are listed in Table 15.5-6. In addition, an evaluation of the dose to control room operators and an evaluation of the offsite doses resulting from recirculation loop leakage are presented.

Fission Product Release to the Containment Following a postulated double-ended rupture of a reactor coolant pipe with subsequent blowdown, the emergency core cooling system keeps cladding temperatures well below melting, and limits zirconium-water reactions to an insignificant level, assuring that the core remains intact and in place. As a result of the increase in cladding temperature and rapid depressurization of the core, however, some cladding failure may occur in the hottest regions of the core. Thus, a fraction of the fission products accumulated in the pellet-cladding gap may be released to the reactor coolant system and thereby to the primary containment.

In this analysis, based on Regulatory Guide 1.4[3], a total of 100% of the noble gas core inventory and 25% of the core iodine inventory is assumed to be immediately available for leakage from the primary containment. Of the halogen activity available for release, it is further assumed that 91% is in elemental form, 4% in methyl form, and 5% in particulate form. The core inventory of iodines and noble gases is listed in Table 15.1-5.

Primary Containment Model The quantity of activity released from the containment was calculated with a single volume model of the containment.

If it is assumed that there are no sources of activity following the initial instantaneous release of fission products to the containment, the equation which describes the time dependent activity or quantity of material in a component is:

ENVIRONMENTAL CONSEQUENCES OF ACCIDENTS 15.5-3

WATTS BAR dt = -AijAij(t) + Pij(t) (1) where Aij is the activity or quantity of material i in component j. Pij is the rate at which activity or material i is added to component j, and Aij is the rate at which activity or material i is removed or lost from component j. If both A and P are independent of time, then for one material and one component one obtains the solution:

A = Aoe-At + P(I1 - e-At (2) where A0 is the initial activity. However, in general, P is time dependent and in some cases A is also time dependent.

The addition of material to the component, Pij(t), may come from two sources: (1) flow from another component in the system may add material to the component, (2) material may be produced within the component by radioactive decay. Thus, the addition rate for material i to component j can be expressed as:

Pi(t) = p 1)(t) + p(2)(t) (3) where:

n P(t) = I cijj _j(t)Aijj(t);cijj _j(t) is the transfer coefficient jj#j n of i from componentjj to j, and P (t) -- iAij(t); Ti_ i is the rate of production

-='i_

ii of i from ii in component j. Note that yii-i is not normally a function of time or component.

Similarly, the loss from a component can be due to: (1) loss within the component (such as radioactive decay), (2) flow out of the component to other components, and (3) removal from the system. Thus, the loss rate from component j for material i can be expressed as:

Aij(t) = Xi1+A () (t)+ A i(t) (4) where A1 is the removal rate inside the component due to radioactive decay (neither time nor component dependent),

15.5-4 ENVIRONMENTAL CONSEQUENCES OFACCIDENTS

WATTS BAR n

A(j) (t)J = fij -jj(t); fij -jj(t) is the transfer coefficient of material i from component j to jj, and A( 3 )(t) is the removal from the system.

A computer program Source Transport Program (STP) has been developed to solve equation (1) for each isotope and for two halogen forms (i.e., elemental and or organic). From this, the isotopic concentration airborne in the containment as a function of time and the integrated isotopic leakage from the containment for a given time period can be obtained. Parameters used in the loss-of-coolant accident analysis are listed in Table 15.5-6.

Modeling of Removal Process For fission products other than iodine, the only removal processes considered are radioactive decay and leakage.

The fission product iodine is assumed to be present in the containment atmosphere in elemental, organic, and particulate form. It is assumed that 91% of the iodine available for leakage from the containment is in elemental (i.e., 12 vapor) form, 4% is assumed to be in the form of organic iodine compounds (e.g., methyl iodine), and 5% is assumed to be absorbed on airborne particulate matter. In this analysis it was conservatively assumed that the organic form of iodine is not subject to any removal processes other than radioactive decay and leakage from the containment. The elemental and particulate forms of iodine are assumed to behave identically.

The effectiveness of the ice condenser for elemental iodine removal is described in Section 6.5.4. For the calculation of doses, the ice condenser was treated as a time dependent removal process. The time dependent ice condenser iodine removal efficiencies for the Regulatory Guide 1.4 analysis are given in Table 15.5-7.

Ice Condenser The ice condenser is designed to limit the leakage of airborne activity from the containment in the event of a loss-of-coolant accident. This is accomplished by the removal of heat released to the containment during the accident to the extent necessary to initially maintain that structure below design pressure and then reduce the pressure to near atmospheric. The addition of an alkaline solution such as sodium tetraborate enhances the iodine removal qualities of the melting ice to a point where credit can be assumed in the radiological analyses.

The operation of the containment deck fans (air return fans) is delayed for approximately 10 minutes following a Phase B isolation signal resulting from the loss-of-coolant accident.

This delay in fan operation yields an initial inlet steam-air mixture into the ice condenser of greater than 90% steam by volume which results in more efficient iodine removal by the ice condenser.

ENVIRONMENTAL CONSEQUENCES OF ACCIDENTS 15.5-5