ML16355A196

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Redacted - Waterford Steam Electric Station, Unit 3, Revision 309 to Final Safety Analysis Report, Chapter 15, Accident Analyses, Section 15.7
ML16355A196
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WSES-FSAR-UNIT-3 15.7-1 Revision 14 (12/05) 15.7 RADIOACTIVE RELEASE FROM A SUBSYSTEM OR COMPONENT 15.7.1 MODERATE FREQUENCY INCIDENTS

Incidents in this category are postulated as limiting faults. Moderate frequency incidents will have radiological consequences less severe than the corresponding limiting fault described below.

15.7.2 INFREQUENT INCIDENTS

Incidents in this category are postulated as limiting faults. Infrequent incidents will have radiological

consequences less severe than the corresponding limiting fault described below.

15.7.3 LIMITING FAULTS

15.7.3.1 Radioactive Waste Gas System Leak or Failure (DRN 04-704, R14)

Deleted.

(DRN 04-704, R14) 15.7.3.2 Liquid Waste System Leak or Failure (Release to Atmosphere)

(DRN 04-704, R14)

Deleted.

(DRN 04-704, R14) 15.7.3.3 Postulated Radioactive Releases Due to Liquid Containing Tank Failures (DRN 04-704, R14)

START OF HISTORICAL INFORMATION.

(DRN 05-1551, R14)

The original waste concentrator of the Waste Management System contained the highest inventory of radionuclides of any component located outside of the Reactor Containment Building. The isotopic inventory was based on a normal operation source term which was based on NUREG-0017. The current source term for core power uprated conditions (3716 WMt) was developed in accordance with ANS 18.1-1999 as discussed in Chapter 11. The analysis discussed below was not revised for power uprate conditions, however is represents a conservative analysis which would be expected to bound plant

operations at 3716 MWt. (DRN 05-1551, R14)

For purposes of analysis it is conservatively assumed that the entire contents of the waste concentrator tank is instantaneously injected into the Zone 3 aquifer, and is transported as a slug of groundwater without any dispersion or ion-exchange taking place. As indicated in Subsection 2.4.13.3 the maximum groundwater velocity in the site groundwater regime is 0.234 ft/yr in the Zone 3 aquifer. The flow of the groundwater is in a southwest direction away from the Mississippi River. Therefore it would take this contaminated liquid about 1000 years to travel to the nearest boundary (250 ft.) of the restricted area in this direction. It should be noted that there are no potable water intakes or drinking wells within the restricted area. Due to radioactive decay the total radionuclide concentration at this point after about

1000 years was calculated to be less than 10 6 of the Maximum Permissible Concentration as defined in Column 2 of Table II of Appendix B to 10CFR20 ( C/MPC < 10 6). (DRN 04-704, R14)

WSES-FSAR-UNIT-3 15.7-2 Revision 306 (05/12)

(DRN 00-592;04-704, R14)

This analysis was based on a liquid waste system wh ich operated on the principle of evaporation. The configuration presently used is based on the principle of demineralization. The demineralizer system was evaluated using very conservative va lues to determine a maximum saturation activity level (i.e.: system total curies) using FSAR Reactor Coolant Isotopic Inventories to determine a weighted average influent and finally determining a dilution factor representative of the LWM stream. The system spent resin (solid waste) was conservatively treated as a liquid waste. (DRN 00-592;04-704, R14)

Since the groundwater velocity in Zone I is appreci ably less than Zone 3, activity level of the nearest surface water is an unrestricted area would be even less.

(DRN 04-704, R14)

END OF HISTORICAL INFORMATION.

(DRN 04-704, R14) 15.7.3.4 Design Basis Fuel Handling Accidents

15.7.3.4.1 Identification of Causes and Frequency Classification

The possibility of a fuel handling accident is remote because of the many interlocks and administrative controls and physical limitations im posed on the fuel handling operations (refer to Subsection 9.1.4). All refueling operations are conducted in accordance with prescribed procedures under direct surveillance of a supervisor technically trained in nuclear safety and fuel handling.

(DRN 02-1729, R12-A) (DRN 02-1729, R12-A)

Design of the fuel storage racks and handling facilities in both the containment and fuel storage area is such that fuel will always be in a subcritical geomet rical array, assuming zero boron concentration in the fuel pool water. The spent fuel pool and refueli ng pool water contains boron at the refueling boron concentration. Natural convection of the surrounding water provides adequate cooling of fuel during

handling and storage. Cooling of the water is provided by the spent fuel pool cooling system. At no time during the transfer from the reactor core to the spent fuel storage rack is a fuel assembly removed from

the water. Fuel failure during refueling, as a result of inadvertent criticality or overheating, is not possible. (EC-14275, R306)

Interlocks and mechanical stops prevent the spent fuel cask handling crane from moving the cask over stored irradiated fuel in fuel st orage racks and limit cask movement. The single-failure-proof main hook on the cask handling crane will move over irradiated fuel in the spent fuel transfer cask, but a fuel handling accident due to a cask drop is not credible si nce a single-failure-proof handling system is being used. (EC-14275, R306)

(DRN 00-1479, R11-A; 02-1729, R12-A) (DRN 00-1479, R11-A; 02-1729, R12-A)

(DRN 03-179, R12-C; EC-28875 R305)

During fuel handling operations, the containment equipment door, personnel airlock doors, and penetrations may be open provided the equipment door, a minimum of one door in the airlock, and other

penetrations are capable of being closed in the event of a fuel handling accident. Each penetration providing direct access to the outside atmosphere s hall be either capable of being closed by an isolation valve, blind flange, manual valve or equivalent or capable of being closed on a containment purge isolation signal (CPIS) initiated by redundant area and airborne radiation monitors. Should a fuel handling accident occur inside containment, the equipm ent door, a minimum of one door in the airlock, and the open penetrations will be closed to minimize the escape of any radioactivity. The containment purge lines are automatically closed upon a CPIS if the fuel handling accident releases activity above prescribed levels.

WSES-FSAR-UNIT-3 15.7-3 Revision 307 (07/13)(DRN 02-1729, R12-A)

For this evaluation, dropping of a fuel assemb ly is assumed to occur breaching the cladding and releasing the volatile fission products in the gas gap of the fuel pins. In addition to the area radiation monitor located in the spent fuel cask area, portable radiation monitors capable of emitting audible alarms are located in this area during fuel handling operations. Doors in the Fuel Handling Building are closed to maintain controlled leakage characteristics in the spent fuel pool region during refueling operations involving irradiated fuel. Should a fuel assembly be dropped in the fuel tr ansfer canal or in the spent fuel pool and release radioactivity above a prescribed level, the airborne radiation monitors sound an alarm, alerting personnel to the problem. (DRN 02-1729, R12-A) 15.7.3.4.2 Sequence of Ev ents and System Operation 15.7.3.4.2.1 Design Basis Sequence of Events and System Operation The refueling procedure is described in Subsection 9.1.

4. The earliest anticipated time at which a spent fuel assembly could be handled is three days after shutdown. (DRN 00-996, R11; EC-5000081470, R301; EC-38571, R307)

For the design basis accident, the failure of two fuel a ssemblies results in 472 rod failures. The failure of 472 fuel rods is the largest number of fuel rods that could fail from the worst postulated assembly drop. (DRN 00-996, R11; EC-5000081470, R301; EC-38571, R307)15.7.3.4.2.2 Structural Ev aluation of Fuel Assembly In this analysis, dropping of a fuel assembly is assumed. Interlocks and procedural and administrative controls make such an event highly unlikely. Howeve r, if an assembly were damaged to the extent that one or more fuel rods were broken, the accumulated fission gases and iodines in the fuel rod gaps would be released to the surrounding water. Release of the solid fission products in the fuel would be negligible

because of the low fuel temperature during refueling. (DRN 00-996, R11; EC-38571, R307)

The fuel handling accident analysis evaluates the case of a fuel bundle being dropped from the fuel handling device and impacting one fuel bundle located in t he spent fuel rack or one or more fuel bundles located in the reactor core. This analysis is performed for the CE 16 NGF and CE 16x16 standard design

fuel assemblies that include the weight of com ponents (e.g., CEA and neutron source), handling grapples and some extra weight for margin. The impacted fuel bundles are considered to be struck by either (1) a vertically dropped fuel bundle, or (2) a horizonta lly dropped fuel bundle. In addition, the analysis considers that the dropped fuel bundle may tip over a fter impact; the impacted bundle in the reactor core may also tip over after impact. In all cases, ener gy balance theory is employed to determine the number of damaged fuel rods resulting from the postulated events.

Since the fuel bundle being transported and the im pacted bundle(s) are submerged in water, all scenarios take place in water at an assumed temper ature of 150°F. As the drop height may vary during movement of a fuel bundle, impact velocities for the vertical and horizontal drop scenarios are conservatively calculated as the terminal velocity of the dropped bundle. In the case of a bundle tipping over, the rotational energy at impact is assumed to be a conservative percentage of the kinetic energy of the horizontally dropped bundle at terminal velocity. T he terminal velocity and the resulting impact energy are dependent on many design parameters including:

fuel bundle weight and buoyancy, component weight, hydraulic drag, Reynolds number. These fa ctors are all considered in determining the impact velocity and impact energy for each scenario.

In determining the number of fuel rods that fail as a result of each fuel handling accident scenario, all the kinetic energy developed by the dropped, or rotating, fuel assembly is absorbed in the form of fuel rod and guide tube deformation and associated strain ener gy. Energy absorption by the guide tubes and (DRN 00-996, R11; EC-38571, R307)

WSES-FSAR-UNIT-3 15.7-4 Revision 307 (07/13)

(EC-38571, R307) instrument tube are accounted for by equating each tube is equivalent to 4 fuel rods; this results in the energy being absorbed by 256 fuel rods per assembly. Note that 256 represents the equivalent number of rods for energy absorption; the ac tual number of rods is 236. This approach is conservative in that it does not take into account the absorption of any impact energy by the upper or lower end fitting components of the dropped and impacted fuel assemblies.

The fuel handling accident analysis also evaluates t he case of a 2,000 Ib object dropped over the spent fuel rack and impacting one or more fuel assemblies.

The terminal velocity of the dropped object and the impact kinetic energy are calculated using the same methodology as that used for dropped fuel bundles.

The calculated impact energy is then used to determine the number of failed rods in the impacted fuel bundles.

For the bundle drop scenarios at both the Spent Fuel Pool and Reactor Core locations, the results

demonstrate that the maximum total number of fuel rods predicted to fail in the dropped and impacted fuel bundles is 472 rods.

(DRN 00-996, R11)

(DRN 00-996, R11)

(DRN 02-1729, R12-A) (DRN 02-1729, R12-A)

(EC-38571, R307)

WSES-FSAR-UNIT-3 15.7-5 Revision 307 (07/13) 15.7.3.4.3 Core and System Performance

This subsection is not applicable for a fuel handling accident.

15.7.3.4.4 Barrier Performance

This subsection is not applicable for a fuel handling accident.

15.7.3.4.5 Radiological Consequences (DRN 02-1729, R12-A) 15.7.3.4.5.1 Design Basis Analysis

(DRN 00-996, R11;04-704, R14; EC-5000081470, R 301; EC-38571, R3-7)

The worst fuel handling accident (assembly drop) that re sults in failure of 472 fuel rods with an assembly power peaking factor of 1.65 is assumed for t he radiological consequences of this event. The assumptions and parameters used in evaluating the fuel handling accident are consistent with Regulatory Guide 1.25 (3/23/72) and Regulatory Guide 1.183 reco mmendations. All the gap activity in the damaged fuel rods is assumed to be released to the env ironment within two hours, consistent with the recommendations of RG 1.183, without any filtration. This assumption e liminates the need for filtration of the activity released to the environment from fuel hand ling building or containment building. The offsite and control room doses are calculated using t he RADTRAD computer code and the dose consequences for this event are found to be well below 10CFR50.67 limits. (DRN 00-996, R11; 02-1729, R12-A;04-704, R14; EC-5000081470, R301; EC-38571, R307)

(DRN 02-1729, R12-A) (DRN 02-1729, R12-A)

(DRN 00-996, R11; 02-1729, R12-A) (DRN 00-996, R11; 02-1729, R12-A)

WSES-FSAR-UNIT-3 15.7-6 Revision 307 (07/13)(DRN 02-1729, R12-A)

Input and Assumptions: (DRN 04-704, R14)

The evaluation for the offsite and control room r adiological consequences of a FHA used the RADTRAD computer code. The analysis incorporated the appropr iate conservative assumptions provided in RG 1.25 and RG 1.183. The following provides the conservative input and assumptions used in the calculations: (DRN 04-704, R14)(EC-38571, R307)Number of Failed Fuel Rods: 472 fuel rods in the dropped and impacted bundles are assumed to fail based on the worst postulated assembly drop (Design Basis Accident). (EC-38571, R307)(DRN 04-704, R14)Gap Inventory: For each isotope the most conser vative calculated activity (i.e., the highest value) for the 30, 40, 50, 60, and 70 GWD/MTU burnups is assumed. Power Level: 3735 MWt = 3716

Radiological source term:

The radiological source term used in the fuel handling accident

is provided in Table 15.7-6. (DRN 02-1729, R12-A) (DRN 04-704, R14)

The release fractions are based on NUREG/CR-5009.

Decontamination Factor:

RG 1.25 states that, "For release pressure greater than 1200 psig, the iodine decontamination factor will be less than those assumed in this guide." For Waterf ord it was determined that the fuel rod pressure exceeds 1200 psig at 49 GWD/MTU at a spent fuel pool temperature of 177°F. However, it was shown that the radiological source term used in Waterford FHA analysis bounds this conditions. (DRN 04-704, R14)(EC-5000081470, R301)

A minimum control room volume is assumed. FHA Ma in Control Room dose has negligible sensitivity to control room volume assumptions. (EC-5000081470, R301) 15.7.3.4.6 Results a)Offsite Doses(DRN 02-1729, R12-A;04-704, R14; EC-38571, R307)

The potential radiological consequences result ing from the occurrence of a postulated fuel handling accident have been conservatively anal yzed, using inputs, assumptions provided in Table 15.7-6 and models described in the preceding subsections. The TEDE dose at exclusion area boundary (EAB) and low population zone (LPZ) is provided in Table 15.7-7.(DRN 02-1729, R12-A;04-704, R14; EC-38571, R307)

WSES-FSAR-UNIT-3 15.7-7 Revision 306 (05/12)(DRN 02-1729, R12-A) b)Dose to Main Control Room Personnel(DRN 04-704, R14)

The radiological consequences TEDE doses to ma in control room personnel following a fuel handling accident are provided in Table 15.7-7.

The control room doses are well within the acceptable limits. (DRN 02-1729, R12-A;04-704, R14) 15.7.3.5Spent Fuel Cask Drop Accidents (DRN 00-996, R11; EC-14275, R306) 15.7.3.5.1 Cask Drop Into Spent Fuel Pool, Cask Pit Storage Area, Cask Decontamination Area, and Rail Bay (EC-30504, R305)

As discussed in Subsection 9.1.4, the cask handling crane carrying a spent fuel transfer cask or other heavy load (i.e., loads in excess of the weight of a f uel assembly plus handling tool) is prohibited from traveling over the spent fuel pool. Only the singl e-failure-proof main hook on the cask handling crane is permitted to carry a spent fuel transfer cask or other heavy load over the cask storage area or any unprotected safety related equipment. Thus, an accident resulting from a drop of a spent fuel transfer cask or other heavy load into the spent fuel pool, t he cask storage area, the ask decontamination area, or the rail bay is not credible. (DRN 00-996, R11; EC-30504, R305) 15.7.3.5.2 Cask Drop to Flat Surface As discussed in Subsection 9.1.4, loaded spent tr ansfer fuel casks are only handled using the single-failure-proof main hook on the cask handling crane as part of a single-failure-proof handling system, and thus a drop of a spent fuel transfer cask is not credibl e and is not postulated. Since a drop of a spent fuel transfer cask is not postulated, the radiological c onsequences of a cask drop a ccident are not evaluated. (EC-14275, R306)

WSES-FSAR-UNIT-3 15.7-8 Revision 14 (12/05)SECTION 15.7: REFERENCES 1.Love, A. E. H., A Treatise on the Mathematical Theory of Elasticity, 4th Edition, Dover Publications, New York, New York, October 1926.2.Gabrielson, V. K., SHOCK - A Computer Code for Solving Lumped Mass Dynamic Systems, SCL-DR- 5-35, January 1966.

WSES-FSAR-UNIT-3 TABLE 15.7-1 Revision 14 (12/05) (DRN 04-704, R14)

TABLE INTENTIONALLY DELETED. (DRN 04-704, R14)

WSES-FSAR-UNIT-3 TABLE 15.7-2 Revision 14 (12/05) (DRN 04-704, R14)

TABLE INTENTIONALLY DELETED. (DRN 04-704, R14)

WSES-FSAR-UNIT-3 TABLE 15.7-3 Revision 14 (12/05) (DRN 04-704, R14)

TABLE INTENTIONALLY DELETED. (DRN 04-704, R14)

WSESFSARUNIT3(DRN 04-704, R14)

START OF HISTORICAL INFORMATION (DRN 04-704, R14)TABLE 15.74 Revision 14 (12/05)

MAXIMUM ISOTOPIC INVENTORIES IN EQUIPMENT CLASSIFIED ASNONSAFETY, NON SEISMIC (CURIES)WASTE MANAGEMENT SYSTEM(DRN 01-1250, R11-B)

Waste Cond.Waste Tanks Waste Laundry Tank Tank Waste Waste Cond.

Total Isotope A&B (each)TankC (each)

(each)

Demineralizer Ion Exchanger For System(DRN 01-1250, R11-B) I131 5.7(0) 5.7(1) 7.1(3) 4.9(3) 2.8(+1) 4.3(2) 9.7(+1) I132 1.6(0) 1.6(1) 2.0(3) 1.4(3) 8.0(0) 1.4(4) 2.7(1) I133 7.1(0) 7.1(1) 8.9(3) 6.1(3) 3.6(+1) 5.8(3) 1.2(2) I134 6.9(1) 6.9(0) 8.7(4) 6.0(4) 3.5(0) 2.4(5) 1.2(1) I135 3.1(0) 3.1(1) 3.9(3) 2.7(3) 1.6(+1) 8.1(4) 5.3(1) BORON MANAGEMENT SYSTEMPreConcentrator Boric Acid Condensate Ion Exchanger Ion Exchanger Boric Acid Isotope (both)(both)ConcentratorKr85m 0.0 0.09.1(2)Kr85 0.0 0.0 9.1(0)Kr87 0.0 0.01.7(2)Kr88 0.0 0.01.1(1)I131 4.0(+1)1.9(2)7.6(0)I1322.6(3)1.3(6)4.1(2)I1337.2(1)3.5(4)1.3(0)I1341.8(4)8.9(8)7.7(3)11353.7(2)1.8(5)2.0(1)Xe131m 0.0 0.0 2.7(0)Xe133 0.0 0.0 2.4(+2)Xe135m 0.0 0.03.4(3)Xe135 0.0 0.06.6(1)Xe138 0.0 0.01.6(3)(DRN 04-704, R14)

END OF HISTORICAL INFORMATION (DRN 04-704, R14)

WSES-FSAR-UNIT-3(DRN 04-704, R14)

START OF HISTORICAL INFORMATION (DRN 04-704, R14)

TABLE 15.7-5 Revision 14 (12/05)

RADIOLOGICAL EXPOSURES AS A RESULT OF LIQUID WASTE SYSTEM FAILURE (CURIES)Boron Management Waste Management Isotope System System Kr-85m 9.1(-2)0.0 Kr-85 9.1(0)0.0 Kr-87 1.7(-2)0.0 Kr-88 1.1(-1)0.0 I-131 4.8(-1)9.7(-1)I-132 4.4(-4)2.7(-1)I-133 2.0(-2)1.2(0)I-134 7.9(-5)1.2(-1)1-135 2.4(-3)5.3(-1)Xe-131m 2.7(0)0.0 Xe-133 2.4(+2)0.0 Xe-135m 3.4(-3)0.0 Xe-135 6.6(-1)0.0 Xe-138 1.6(-3)0.0 Dose Boron Management Waste Management (rem) System System EAB Thyroid 1.6 (-1) 1.6 (-1)

Whole body 1.2 (-3) 2.9 (-4)

LPZ Thyroid 1.8 (-2) 1.8 (-2) Whole body 1.3 (-4) 3.2 (-5) (DRN 04-704, R14)

END OF HISTORICAL INFORMATION (DRN 04-704, R14)

WSES-FSAR-UNIT-3 TABLE 15.7-6 (Sheet 1 of 2) Revision 307 (07/13)

PARAMETERS USED IN EVALUATING THE RADIOLOGICAL CONSEQUENCES OF A FUEL HANDLING ACCIDENT (DRN 04-704, R14)

Core Power Level:

3735 MWt Core Inventory: (Ci/MWt) Kr-85 Kr-85m Kr-87 Kr-88 I-131 I-132 I-133 I-134 I-135 Xe-131m Xe-133 Xe-133m Xe-135 Xe-135m 4.520E+02

1.378E-01

5.849E-14

2.071E-04

9.867E+01

3.985E+01

9.764E+00

9.488E-23

4.991E-02

1.551E+02

1.667E+04

3.713E+02

2.223E+02

1.625E+00 Fission Product Gap Fractions:

I-131 Kr-85 Other Noble Gases Other Halogens Alkali Metals 12%

14%

5%

5%

12% (EC-38571, R307)

Fuel Rods Failing (maximum): (EC-38571, R307) 472 rods Iodine Chemical Form *:

Elemental

Organic Particulate 4.85%

0.15%

95% *The releases from the pool are conservative ly modeled as 99.85% elemental and 0.15% organic iodine.Control Room Parameters: (EC-5000081470, R301)

Volume 168,500 ft 3 (EC-5000081470, R301)

Recirculation Flow Rate 3800 CFM Iodine Filter Efficiency 99% (elemental/particulate/organic)

Pressurization Flow 225 CFM (EC-5000081470, R301) Unfiltered Inleakage 100 CFM (EC-5000081470, R301)

Breathing Rate 3.47E-04 m 3/sec. Control Room Occupancy Factors 0-24 hours 1.0 (DRN 04-704, R14)

WSES-FSAR-UNIT-3 TABLE 15.7-6 (Sheet 2 of 2) Revision 301 (09/07)

PARAMETERS USED IN EVALUATING THE RADIOLOGICAL CONSEQUENCES OF A FUEL HANDLING ACCIDENT (DRN 04-704, R14; EC-5000081470, R301)

Main Control Room /Q Assumed:

Time Unfiltered Inleakage East MCR Intake Pressurization Flow

West MCR Intake 0-2 hr 2.77E-03 3.90E-04 *

WSES-FSAR-UNIT-3 TABLE 15.7-7 Revision 14 (12/05) (DRN 02-1729, R12-A)RADIOLOGICAL CONSEQUENCES OF A POSTULATED FUEL HANDLING ACCIDENT(DRN 02-1729, R12-A)(DRN 00-996, R11; 02-1729, R12-A;04-704, R14)

FHA Acceptance Criteria EAB (worst two hour dose)

< 6.3 6.3 Rem TEDE LPZ (duration)

< 6.3 6.3 Rem TEDE MCR < 5 5 Rem TEDE (DRN 00-996, R11; 02-1729, R12-A;04-704, R14)

WSES-FSAR-UNIT-3TABLE 15.7-8INFORMATION NEEDED TO EVALUATE CONTAINMENT ISOLATIONCAPABILITY DURING REFUELING ACCIDENT 1.2.

3.4.

5.6.

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