ML19332G199

Rev 0 to Radiological Safety Analysis for Spent Fuel Storage & Handling.

ML19332G199
Person / Time
Site:	Shoreham File:Long Island Lighting Company icon.png
Issue date:	09/18/1989
From:	Gillett T, Kascsak R, Paccione R LONG ISLAND LIGHTING CO.
To:
Shared Package
ML19332G153	List: ML19332G152 ML19332G156 ML19332G199 ML19332G607
References
NED-4170024, NED-4170024-R, NED-4170024-R00, NUDOCS 8912200302
Download: ML19332G199 (37)

Text

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i SHOREHAM NUCLEAR POWER STATION l RADIOLOGICAL SAFETY ANALYSIS FOR SPENT FUEL STORACE AND HANDLING NUCLEAR ENGINEERING DEPARTMENT September 1989 -

, O krepared By: [

T. C. Gillett

. f!/[8 L Acting Manager-Radiation Protection Division Reviewed Byt h' P-er-/f

~It . ' 3 . Paccione

', Acting Manager-Nuclear Analysis Division Approved 3y: /M *, .ds.4.ed R. M. K's s e s a k.

9!//8 Manager, Nuclear ErigineeringD /e// partment 8912200302 892225

.j.DR ADOCK 05000322 PDC

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TABLE OF CONTENTS ,

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RADIOLOGICAL SAFETY ANALYSIS FOR SPENT FUEL STORACE AND HANDLING

1. EXECUTIVE

SUMMARY

II. PLANT CONFIGURATION

SUMMARY

III. Sar?.TY ANALYSIS i

A. Radioactive Invantory #

B. Heat Generation Analysia C. Evaluation of Cooling Requirements (Passive / Active)

D. Fuel Criticality Analysis Reviev . . , ;'

E. Accident Analysis

1. Review of Chapter 15 Applicability

2. Fuel Bundle Drop

3. Radvaste Tank Rupture '

4 Worst Case Fuel Damage Event IV. REFERENCES l

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1. EXECUTIVE

SUMMARY

) The purpose of this report is to provide a radiological safety analysis for the storage and handling of  !

Shoreham'a low burnup first cycle spent fuel.

This analysis report is based on the fact that the 560  :

fuel bundles comprising the Shoreham core are stored i under water in the Shorehan spent fuel pool. The fuel ,

bundles are held in a Seismic Category I spent fuel rack within the stainless steel lined spent fuel pool. The ,

spent fuel pool is located in the secondary containment of the Shoreham reactor building. The structures are designed to withstand seismic loads.

It is important to understand that the Shoreham spent fuel is in a low burnup condition. The Shoreham Nuclear "

Power Station operated during low power testing at power levels not exceeding 5% of rated power. The effective  :

burnup of the fuel is approximately 2 full power days.

This results in an estimated total core-wide heat generation rate of approximately 550 watts as of June 1989. The estimated fuel heat load will reduce to approximately 250 watts by June 1991. Figure I-1 depicts the fuel heat load versus time. Based on this low heat generation rate systems for active cooling are not required, and only minimal capacity systems are

/' required for pool makeup to handle evaporation.

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The Shoreham spent fuel contains limited quantities of

• radioactive materials that are available for release.

It has been calculated that approximately 176.000 curies of radioactivity reside in the 560 fuel assemblies. The radioactive inventory estimation is based on a two year decay after the last burnup period. The total gaseous y activity is primarily Krypton-85 (a noble gas with a  :

L half life 10.7 years) and consists of approximately 1560 ,

! curies. Krypton-85 is the only isotope in the fuel that 1

exists in significant quantities and is available for release-in gaseous form during postulated accidents.

Other sources of radioactivity outside the core are minor. -

A spectrum of accidents were identified for radiological analysis. The accidents were identified by reviewing the Shoreham USAR for those events that apply to the A - 1_1 -

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i storage and handling of spent fuel. Based on this l review, the following events were identified for analysist

1. Fuel Randling Accident (Fuel Bundle Drop) ',

, 2. Radwaste Tank Rupture In addition, a worst case radiological event was ,

postulated in which the entire gaseous activity of the '

whole core is released to the reactor building. This event was postulated to conservatively bound any possible situation involving large-scale mechanical ,

damage to the fuel.

l l- The results of the radiological analysis indicate that integrated doses are very small in comparison with 10CFR100 limits. The resules of the radiological '

analysis indicate that integrated doses are very small in comparison with 10CTR100 limits. For the fuel handling accident and the worst case scenario, a spectrum of cases was analyzed, as follows: operation of the standby ventilation system, operation of the  ;

L normal ventilation system, and no ventilation (modelled as a puff release). The results of the fuel handling accident analyses indicate that the integrated offsite whole body and skin doses, with the Reactor Building Normal Ventilation System operational, are approximately

( 0.00005% or less of 10CFR100 limits. For the worst case  ;

scenario, under the same HVAC conditions, the doses are l approximately 0.03% or less of 10CFR100 limits. The I results of the radiological analyses are depicted graphically in Figures I-2A and 1-2B, for the fuel handling accident and worst case scenario, respectively.

In particular, it was demonstrated that the reactor building standby ventilation system operation does not provide an important filtering or ventilation safety function and is therefore no longer required now that fuel is located in the pool.

Based on this analysis, it has been found that the spent  !

fuel pool provides a high degree of passive safety protection for Shoreham spent fuel. Active safety systems are not required to mitigate postulated accidents; however, support systems are required ts meet the intent of the requirements of 100FR50 Appendix A, General Design Criteria; and Regulatory Guide 1.13. j Supporting systems are required to provide for radiation monitoring, fuel pool makeup, fuel pool cleanup, radwaste, and normal support systems to maintain building services.

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11. PLANT CONTICURATION

SUMMARY

)

[' This analysis is based on the fact that the Shoreham i-s initial core spent fuel will be stored for some interim period in the spent fuel pool contained within the SNPS reactor building. ,

The configuration of the plant is summarized as follows: I

1. All 560 fuel bundles have been removed from the '

reactor and are being stored in the spent fuel storage pool. The tocal decay heat power of the j entire core has been determined to be approximately 550 watts as of June 1989.

2. The spent fuel storage pool water level will be  !'

maintained at its normal water level. Makeup will be furnished from the condensate transfer system or '

the domineralized and makeup water system. The fuel pool cooling system is not in service due to  !

the low heat load in the pool. Water quality will '

be maintained by the fuel pool cleanup system. The spent fuel pool transfer canal gates will remain installed.

3. The steam separator and dryer has been placed back in the reactor and the reactor vessel head has been r- placed on the reactor flange, but the studs have not been tensioned.

4 The drywell head has been re-installed and the reactor cavity and dryer / separator pools have been drained. t

5. All reactor protection, nuclear steam supply shutoff and emergency core cooling systems are to ,

be de-energized and isolated.

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6. The reactor building normal ventilation system will be operated to provide suitable environmental conditions and to allow for radiological monitoring of building releases.

7. Radwaste systems will be maintained as required by the above.

'1I1. SATITY ANALYSIS '

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( ,) A. Radioactive Inventory

1. Tuel Sources The Shoreham reactor core has undergone three periods of low power (0-5%) testing over the past four years. The low power tests are summarised belows i

Specific Burnup Power Test Period Duration MWD /MT Ranae 1 7/7-10/7/85 93 days 27.8 0.0 - 3.3 8/5-8/30/86 26 days 13.8 0.0 - 4.0 5/26-6/6/87 12 days 6.7 0.0 - 3.5 Total 48.3 The detailed profiles of the above three low power test periods have been input to the ORIGEN2 (Reference A) burnup code, along with the physical characteristics of the reactor fuel and bundle structural elements. Results of this  ;

analysis (Reference B) are given in Table III.A - 1. The activities correspond to two years decay after the last burnup period, and reflect total core inventories for those isotopes with greater than 10 curies.

As can be seen from the Table III.A-1 only long-lived isotopes remain from the original actinides and

• fission / activation products created, along with their equilibrium daughters. By'far the most radiologically significant, from a gamma dose rate standpoint, are the t Cs-137/Ba-137m pair; about 80% of the whole body dose rate from a spent fuel bundle is due to the Ba-137m photon '

(Reference C). For dose assessment of accidental gaseous releases (e.g., a postulated fuel handling accident), only Kr-85 is meaningful (Reference E).

2. Non-Tuel Sources

a. Liquid Sources With the possible exception of liquid radwaste streams, ,

reactor water would be expected to have the highest concentration of radionuclides of any liquid stream in the plant. At SNPS as of June 1989, the concentration of all radionuclides in reactor water is less than the lower limit

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of detection (LLD). per Reference F. It is concluded that  ;

(j) the liquid streams outside the radwaste system are j s insignificant sources of radioactivity.

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b. Caseous Sources ,

i There are no detectable gaseous sources at SNPS. either j present or anticipated. This statement is supported by the fact that the most recent Semi-Annual Radiological Effluent Release Report (Reference D) indicates there were no  !

detectable releases during the six-month period, either from the offgas system or the various building exhaust systems.

c. Redweste Sources I l

With the exception of the low burnup fuel, radwaste is the l only area of the plant with sessurable activity. The maximum ,, i whole body gamma dose rate in the plant is about 3.5 area /hr, i near the Spent Resin Tank.  !

l l Current-(6/30/89) isotopic concentrations above LLD levels in 1 the Radwaste System are indicated in Table III.A - 2, from J j Reference G, H and I.

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( d. Activated Materials Sources I l '

l f- There are no significant out-of-core radioactive materials j ( sources activated at SNPS. While the low power testing i program may have activated some materials inside the RPV.

l these are not considered significant compared to spent fuel sources.

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- III.A 5

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l 7_x TABLE III.A - 1

' s, Fuel Source Terms ISOTOPE CURIES HALF-LIFE H-3 1.77E+02 1.23E+01 years Mn-54 3.36E+01 3.13E+02 days ,

Fe-55 8.06E+02 2.70E+00 years Co-60 5.64E+02 $.27E+00 years Ni-63 4.28E+01 1.00E+02 years I

Kr-85 l'.56E+03 1.07E+01 years Sr-89 1.54E+01 5.05E+01 days Sr-90 1.37E+04 2.86E+01 years ,

Y-90 1.37E+04 6.41E+01 hours '

Y-91 6.81E+01 5.85E+01 days Zr-95 1.48E+02 6.40E+01 days Nb-95 3.49E+02 3.51E+01 days Ru-106 5.98E+03 3.68E+02 days Rh-106 5.98E+03 2.99E+01 seconds Sn-119m 3.30E+02 2.93E+02 days l Sb-125 1.45E+03 2.77E+00 years Te-125m 3.53E+02 5.80E+01 days e Te-127 1.49E+01 9.35E+00 hours I Te-127m 1.52E+01 1.09E+02 days ,

Cs-134 1.33E+02 2.06E+00 years Cs-137 1.48E+04 3.02E+01 years Ba-137m 1.40E+04 2.55E+00 minutes ,

Ce-144 3.55E+04 2.84E+02 days Pr-144 3.55E+04 1.73E+01 minutes Pr-144m 4.26E+02 7.20E+00 minutes Pm-147 2.95E+04 2.62E+00 years Sm-151 3.60E+02 9.00E+01 years .

Eu-154 1.18E+01 8.80E+00 years ,*

Eu-155 4.47E+01 4.96E+00 years U-234 1.02E+02 2.45E+05 years Th-234 3.38E+01 2.41E+01 days Pa-234m 3.38E+01 1.17E+00 minutes U-238 3.38E+01 4.47E+09 years Pu-239 2.77E+02 2.41E+04 years Pu-241 5.58E+01 1.44E+01 years Total 1.76E+05 Note: Only' isotopes with activity greater than 10 curies are listed.

- III.A . -- -_ -

TABLE III.A - 2

) Radwaste Sources Creater than LLD Spent Resin Tank, Radwaste Filter, & Floor Drain Filter

. The activity concentration is assumed to equal the maximum in the most recent MIC shipment (Nov-Dec 1988) is (from Reference G):

Activity Isotope Concentration, oCi/ec  % of Activity Ci-51 9.84E-04 58.46%

Mn-54 2.17E-05 1.29%

Fe-55 4.19E-04 24.88%

Co-57 7.92E-07 0.05% '

Co-58 6.43E-06 0.38%

Co-60 1.09E-04 6.51%

Fe-59 4.57E-05 2.71%

Ni-63 6.41E-06 0.38%

Sb-124 3.25E-06 0.19%

Zn-65 1.89E-05 1.12%

H-3 6.21E-06 0.37%

C-14 3.94E-07 0.02%

St-90 1.69E-07 0.01%

Zr-95 1.52E-05 0.91%

i Nb-95 2.55E-05 1.51%

Tc-99 4.79E-09 0.00%

I-129 7.32E-10 0.00%

Cs-137 1.34E-06 0.08%

Ce-144 2.95E-06 0.18%

Pu-241 1.59E-05 0.95%

Discharge Supply Tanks The activity concentration in these tanks is assumed to equal the maximum concentration measured in the past 12 months (from Ref.

H):

Activity Isotope Concentration, uC1/cc  ! of Activity Co-60 7.83E-08 100.0%

Note: The remaining radwaste tanks (floor drain collector tanks, vaste collector tanks, and recovery sample tanks) were all determined in Reference I to have isotopic concentrations less than LLD.

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I 111.5 EEAT GENERATION ANALYSIS J

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'O One result from the OR1 GEN 2 calculation 1s a tabu,lation of decay heat or thermal power (in watts), as a function of time. Results of this analysis are presented in i

i Tigure 111.5-1. The calculated decay heat load as of l June 1989 is approximately 0.55 kw.

J 1t must be recognised that there are some limitations in  :

the ORIGEN2 model, and potential inaccuracies in the J calculational processes of the code and its supporting  ;

data sets. For instance. ORICEN2 is a " point reactor"  ;

model, and cannot deal conveniently with the spatial .

variations in fuel enrichment and burnup. In addition, '

there are uncertainities associated with averaging of nuclear cross-section data within the thermal, resonance, and fission neutron energy ranges. ,

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I III.C EVALUATION OF SPENT FUEL POOL COOLING REQUIREMENTS

) An analysis (Reference K) was performed to determine the rate of water loss from the spent fuel pool through evaporation under the worst case scenario described below. The time it took to uncover the spent fuel based I on the calculated evaporation rate was then determined. '

The following assumptions are used in the analysis to maximize the calculated pool evaporation rate and hence nintaire the time required to uncover the Shorehas low burnup spent fuel:

1) The spent fuel pool temperature is conservatively  ;

kept at 110'F.

2) The ambient temperature above the spent fuel pool is conservatively assumed with zero relative humidity.

j 3) The reactor building air flow exists due to normal ,

j ventilation system operation to maximize l evaporation.

• The result of the calculation shows that the maxinua evaporation rate from the pool is approximately 0.6 gpa which translates to a pool level depletion rate of one '

r~s foot per eleven days. Technical Specifications require

( that the water level above the spent fuel be at least ,

twenty-one feet. In addition, it should be noted that pool water level is alarmed in the control room and ,

L alara response procedures exist to provide appropriate operator action.

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a, 1 III.D FUEL CRITICAL.ITY ANALYSIS

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( J The Shoreham Spent Fuel Rack (SFR) is of a stainless steel.and water neutron flux trap design which uses no

-additional poison. The criticality analysis of this rack deeign is des:ribed in detail in Appendix 9A of the Shoreham USAR. The reactivity results which are l 4

summarized in USAR-Table 9A-4 of the same document remein valid for the conditions existing at Shoreham after defueling.- Furthermore, due to the differences in U-235' enrichment between-the: designed ~and the current fuel in the core,-a large negative reactivity credit shculd be taken into account. This,is explained as follows:

9 The Shoreham SFR desi&n isebased on a maximum U-235 enrichment of 3.1 w/o. The resulting basic cell k is calculated to be 0.9129 without any uncertainty "

, and model adjustments-(Table 9A-4, Appendix 9A, c Shorehan USAR). The Shoreham Cycle 1 fuel loading-i uses three (3) enrichments. of the 560Lfuel aesemblics in the core, 340 bundles have the highest bundle average U-235 enrichment of 2.19 w/o, 144 bundles of 1.76 w/o and 76 remaining bundles use natural uranium.

If the six ft.ch natural uranium segments at the top

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and bottom of the fuel are excluded, the average E

enrichment of a 2.19 w/o bundle becomes-2.33 v/o.

Using this, enrichment and linearly extrapolating E 'the reactivity vs. U-235 enrichment-results given in Figure 9A-5 of Appendix 9A, Shoreham USAR, the o reactivity difference between the design enrichment '

L of 3.1 w/o and the current maximum loading enrichment of 2.33 w/o is found to be about -6.0%

in ak(ak =-0.060). This brings the basic cell k.

under nominal storage conditions for the current-fuel in the core down to -0.85, wh,1ch is well below the regulatory acceptance criterion of km 6 0.95.

All the corrective and uncertainty adjustments listed in Table 9A-4 of the Shoreham USAR' remain applicable.

During the period frem July. 1985 co June, 1987, 3 Shoreham went through three separate stages of low power

. testing (less than 5% of rated power), which resulted in L a total core exposure of approximately 48 mwd /MT as determined by a series of core-follow analyses. The net effect of the core exposure is a slight decrease in reactivity ( -0.002 in A k) mainly due to the of f setting

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contributions from the formation'oi Sm-149 and the l M -4 slight depletion of the burnable Gd poison in the fuel lQ '

bundles. In light of the large reactivity margin-described previously (k .~ 0.85), no additional credit will be claimed here, 5

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III.I.1 -0VERVIEW OF USAR CHAPTER'15 EVENTS 1

Introduction Chapter,15 of SNPS USAR provides the results of analyses l of the spectrum of transient and accident events which '

are postulated to-occur with the plant operating i initially at maximum power. The purpose of this J analysis is to identify USAR transients and accidents that apply-to the storage and handling of the low burnup l

. fuel.-  :

The analysis is based on the fact that the fuel is removed from the core and is stored in the spent fuel pool. The total decay heat is approximately 550 watts, which is small enough that it could'be removed by passive cooling and would not require the fuel pool cooling system. Normal and emergency makeups are available.

l p The design basis of the spent fuel storage excludes fuel L '

uncovery under any postulated loss of coolant (Reference Section II.D. Regulatory Position 6).

As-the reactor is not operating.and the fuel is not in L the core, most.of the USAR Chapter 15 events cannot Q

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occur.

r Analysis 1

The safety parameter which is evaluated for each transient of.USAR Chapter 15 is Minimum Critical Power Ratio (MCPR) which is a measure of the fuel cladding integrity. Maximum Average Planar Linear. Heat Generation Rate (MAPLHGRS is the safety perameter for the reactor LOCA-related accidents, and indicates g whether the cladding temperature and the zircenium-water l reaction is below the specified limits. As the decay L power level is extremely low during spent fuel storage, MCPR nnd MAPLHGR are of no concern. Most of the ll transients and. accidents of USAR Chapter 15 occur at operating con '4 tions and are therefore not applicable.

Those transients and accidents of USAR Chapter 15 which pose radiological release outside the primary containment barrier are of primary concern.

The USAR Chapter 15 events are assigned to one of six L analytical categories. The next section presents those L analytical categories and discusses all events one-by-one in each analytical category.

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L III.E-1 -

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.si Decrease in Core: Coolant Temperature

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This analytical' category of USAR Chapter 15 events includes the-following events:

15.1.5 Pressure Regulator Failure - Open )

. 15.1.7 Feedwater Controller Failure - Maximum 1 Demand j 15'.1.8 Loss of Feedwater Heating ,

15.1.9 Shutdown Cooling (RHR) Malfunction -

Decreasing Temperature.

In the spent fuel storage' condition, the pressure '

regulator, feedwater controller, feedwater heating system and RHR system era not operating and all'four j-trenstante are, therefore, not applicable.

Increase in Reactor Pressure

Since generator, turbine, main steam isolation valve, pressure regulator, feedwater system, condenser and RER-systems are not operating, the following transients are not applicable:

15.1.1 Generator Load Reduction 15.1.2 Turbine Trip 15.1.3 Turbine Trip with Failure of Generator n/

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• 15.1.4 Breakers to Open Main Steam Isolation Valve Closure 15.1.6 Pressure Regulator Failure - Closed 15.1.18 Loss of Feedwater Flov 15.1.21 Loss of-Condenser Vacuum 15.1.26 Core Coolant Temperature Increase

-The transient of this category applicable to spent fuel storage is the following:

15.1.19 Loss of'AC Power A loss of AC power condition can be postulated that will affect normal support systems. However, because of the very low heat generation rate (550 watts) and large r thermal-capacity of the pool, loss of normal cooling and makeup systems will result only in.a very slow evaporation of the pool water. This evaporation rate is so slow that ample time exists to restore normal pool makeup sources so.that pool level can be quickly restored. Thus, the passive protection provided by the spent fuel pool eliminates the need for active cooling requirements. The rece of evaporation is discussed in Section III.C.

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Decrease in Reactor-coolant Flow Rate I '

Recirculation pump and recirculation flow controller are I not operating and therefore all the transients of this category are not applicable:

a 15.1.20- Recirculation Pump Trip 15.1.22 Recirculation-Pump-Seizure .i 15.1.23 Recirculation Flow Control Failure -

Decreasing Flow Reactivity and Power Distribution Anomalies -,

t L Events included in this category are those which cause.

rapid-increase in power.' Since the reactor is not E fueled, the following events are not applicable

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15.1.11 Continuous control Rod Withdrawal during h

Power Range Operation i 15.1.12 Continuous Control Rod Withdrawal during ,

Reactor Startup

• 25.1.13 Control Rod Removal Error during

, Refueling L 15.1.14 Fuel Assembly Insertion Error during l

Refueling f,s 15.1.15 off-Design operational Transient due to ,

LjA -) ~ , Inadvertent. Loading of a Fuel Assembly-1

=into an Improper Location '

15.1.16 Inadvertent Loading and Operation of a Fuel Assembly in Improper Location 15.1.24 Recirculation Flow-Control Failure with increasing Flow 15.1.25 Abnormal Startup of Idle Recirculation Pump 15.1.33 Control Rod Drop Accident Increase in Reactor Coolant Inventory Since the HPCI is not operating the following transient is not. applicable:

15.1.10 Inadvertent HPCI Pump Start 1:) - 111.E , ,

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K Decrease in Reactor Coolant Inventory Events Not Applicable te Spent Fuel Storage Safety relief valve and feedwater system are not operating; therefore the following events are not applicables 15.1.17 Inadvertent Opening of a Safety Relief Valve 15.1.37 Feedvater System Piping Break The following event is not a design basis event and is applicable only to power operation 15.1.27 Anticipated transient without Scram (ATWS) l The single failure-proof polar crane design eliminates '

the following event:

15.1.28 Cask Drop .\ccident

. Instrument line, coolant line and steam line breaks 1 present no consequences due to their inoperable status '

L and therefore.the following eventa are not applicable:

)

'15.1.30 off-Design Operational Transient as a Consequence of Instrument Line Failure L 15.1.34 Pipe Breaks Inside the Primary L

Containment (Loss-of-Coolant Accident) 15.1.35 Pipe Breaks Outside Primary Containment

-(Steam Line Break Accident) 1.

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Events Without Fuel Damage 1( )

15.1.27 Miscellt.neous Smal3. Release Outside Primary Containment Releases that could result from piping failures outside '

the primary containment include the pipe breaks in the fuel pool cleanup system. The offsite dose resulting fros'this will be negligible and is bounded by.the Radwaste Tank Rupture accident.

15.1.31 Main condenser Gas Treatment Systek Failure As the mai~ condr- , is e~s s at ng, tL^ offsit 'osa-resultin- of t . v . .'41 * .tigtele. _,

15. )~ te Tank Rupture

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' k~s' Events with Fuel Damage I 15.1.36 Fuel Handling Accident The fuel handling accident is assumed to occur as a consequence of a failure of the fuel assembly lifting

mechanism. . This could cause fuel damage and radioactivity release to the seccndary containment. ,

This event is analyzed in Section III.E.2.

Other Events Seinmic Event Because the spent fuel pool structure _and fuel racks _

meet seismic Category I requirementw, a seismic event is not postulated to create a radiological release.

L' CONCLUSION ,

The following events were idu.tified for radiological analysis:

f -

111.1 -

o - . . _ _ , . _ . . _ -_ _

~! . >\

.:1 I

i Der.ign-Basis Accident Analyses:

1:st >

1. Fuel. Handling Accident (Regulatory Guide 1.25 assumptions) -

2. Radwaste. Tank Rupture In addition, a worst case fuel damage accident was analyzed involving the release of the total gaseous

' inventory af the fuel. The remainder of Section III:

provides a detailed analysis of the above ident'ified events.-

4 l'

x d,Ps I~

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n.

l:

1.

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III.E -

0--.-_-__.--.---...-----..__.-_-----___--__-.____.-_-__--- - , ' -

.1 H l

d

- III.E.2 FUEL HANDLING ACCIDENT . 1

\

)

~

.( III.E.2.1 Identification of Causes

.1 The. fuel handling accident is< assumed to occur as_a  ;

l consequence of a failure of the fuel assembly lifting-  !

I mechanism, resulting in the dropping of a raised fuel i H . assembly onto the top of the core or the spent fuel L racks.

III.E.2.2' Starting-Conditions and Assumptions

? Accidents that result-in the release of radioactive L untarials directly to the secondary containment can

-occur when the fuel is being handled. In this case, radioactive material released as a result of fuel damage is available for transport-directly to the secordary l- "

containment. Table III.E.2-1 presents the parameters t

used in'this: analysis.  ;

y III.E.2.3 Accident Description The most severe' fuel handling accident from a .;

radiological viewpoint is the dropping of a fuel assembly. The sequence of events is as follows:

Approximate Event Elapsed Time

1. LFuel assembly.is being handled by- 0+

refueling equipment. The assembly drops.

2. Some of the fuel rods in both the .

1 min.

dropped assembly- and- another assembly are damaged,'resulting in the release of' gaseous fission products to the reactor coolant and eventually.to the secondary containment atmosphere.

3. The reactor building _ refueling floor 1 min.

exhaust radiation monitoring system may alarm to alert plant personnel.

4. Operator actions begin. 5 min.

q:

j j - III.E . . - . , . . .- - - . -. . . .- . - - . .-

- - . - - -. - -. . . .. _ . - ~ - .

)

s III.E.2.4 Identification of Operator Act$ons (k 1. The-operator may initiate the evacuation of the secondary containment.

2. The fuel handling foreman may instruct personnel to go immediately to the radiation protection personnel decontamination area.

m 3. The fuel handling foteman will make the operator aware.of the accident.

~4. The operator may initiate action to determine the extent of potential radiation doses by measuring the radiation levels in the vicinity of or close to ,

the secondary containment.

5. If RBSVS were to be used, the operator or a delegate would determine whether the RBSYS is performing as designed. (See Section III.E.2.5) l 6. The HP technican wil1~ post the appropriate L radiological control signs at the entrance to the  ;

h secondary containment.

L

7. Before-entry to the secondary containment is made, j a careful study of1 conditions, radiation levels, 6\ etc., will be performed.

III.E.2.5 HVAC Scenarios Considered

, As vill.be seen in Section III.E.2.6.2, the quantity of gaseous fission products in the fuel's gap which is released is not large (2.52.C1 of Kr-85 only).

Calculations. indicate that the reactor building refueling floor exhaust radiation monitoring system would not alarmLand consequently the RBSVS vill not be actuated (i.e., the.RBNVS continues to operate). As e result, analyses were performed assuming either RBSVS or RBNVS system operation. Secondary containment discharge l rates are 167 and_6580 percent / day for the RBSVS and l

RBNVS cases, respectively. As's comparison case,.a

" puff" release over a short period of time (2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, as suggested by Regulatory Guide 1.25), has~been analyzed. '

l Although this is not'a design basis case, it is useful to compare it with the two HVAC cases. Results for all three cases (assuming RBSVS, RBNVS, and puff release)

L are given in the following sections. -

l-i :

- III.E -. ss

-n... .

~

III.E.2.6 Analysis of Effects and Co:isequences III.E.2.6.1 Evaluation Methods The analytical methods.and associated assumptions used to evaluate the consequences of this accident are ,

consistent with Regulatory Guide 1.25, and are quite conservative. The assumptions and parameters are given in Table III.E.-2.

III.E.2.6.1.1 Methods,-Assumptions, and Conditions The assumptions used in the analysis of this accident are listed below:

1. The fuel assembly is dropped from the maximum height allowed.by the fuel handling equipment, ,

2. The entire amount of potential energy,  !

referenced to the top of the spent fuel racks, is'available for application to the fuel assemblies involved in the accident. -This E -assumption neglects the dissipation'of some of' the' mechanical energy of the falling fuel assembly in the water above the racks and p ' requires the complete detachment of the E assembly from the' fuel hoisting equipment.

L/$/ 'N This.is possible if fuel assembly handle, the L fuel grapple, or the grapple cable breaks.

p

3. None of the energy associated with the dropped L fuel. assembly is' absorbed by the fuel material L (uranium dioxide). ,

L III.E.2.6.1.2 Results and Consequences III.E.2.6.1.2.1 Fuel Damage The analysis of USAR' Set' ion 15.1.36.5.1.2.1 applies'to this accident. In that section of

.the USAR, it was assumed that 125 fuel rods would fail as a result of dropping the fuel assembly into the reactor vessel.

III.E.2.6.1.2.2 Fission Product Release From Fu,el

~ Fission product releases for the fuel handling accident are determined from the inventory in Table III.A-1. Specifically, it is seen that

,/ .

J ). -

III.E i _ _ . _ ____ ____ _ __ _ _ _ _ _ _ _ _

y. .

e only Kr-85.is of any significance with respect

.v s to gaseous releases.' The only other gaseous

( [- isotope in this table is H-3, which would add, at most, 0.1% to the skin dose from Kr-85.

Using-the above number of failed rods, and the assumptions given in Regulatory' Guide 1.25, the quantity of Kr-85 released, from Reference E, 1s.as follows:  ;

Release = 1.56E+03Ci x 125 damaged rods ,

62 rods / bundle x 560 bundlen in core x 1.5 peaking. factor x 30% in gap = 2.52 Ci III.E.2.6.1.3 Radiological Effects ,

Offsite Radiological exposures have been evaluated for '

the meteorological conditions, parameters, and assumptions given in Table III.E.2-1. The results are given in Table III.E.2-2.

Control Room Because the amount of radioactivity released

[]

(J is so small, the control room air intake monitors will not alarm..and the HVAC system-will continue to function in its normal operating mode. The resultant whole body and skin 30-day integrated doses are, at aost, 9.59E-08 and 2.08E-04 mrem, respectively, well below the 10CFR50 GDC 19 limits (Reference L).

Discus,sion Ic is seen in Table III.E.2-2 that the (0-2 hour) EAB and (0-30 day) LPZ-integrated doses are many orders of magnitude below 10CFR100 limits. Results are graphically shown in Figure III.E-1. Furthermore, ~ the maximum (t=0) dose rates (whole-body and skin) are very low and, with the exception of the RBNVS case, below Technical Specifications (see Figure III.E.-1A). This indicates that the HVAC system in use in the reactor building has no meaningful effect on radiological

., consequences to members of the public during a fuel

[ handling accident, with the present fuel source terms.

() - III.E l y -- -

3 .w- *,w. , , , . , . , , - - .

5 i t

r i

TABLE III.E.2-1 I'S -. FUEL HANDLING ACCIDENT - PARAMETERS "k- s FOR POSTULATED ACCIDENT. ANALYSES Conservative (NRC)

Assumptions ,

I. Data and assumptions used to estimate radioactive source from postulated accidents A. . Power level See Section III.A ,

B. Peeking factor. 1.5 C. Fual damaged 125 rode D. Release.of activity

, from feel 30% Kr-85 L E. Iodine fractions t- (1) Organic N/A l

(2) Elemental N/A g (3) Particulate N/A II. Data and assumptions used to ,

estimate activity released r

g %p- A. Secondary contain- See Section '

-:( j' ment discharge-rate (%/ day) III.E.2.5 B. Adsorption and filtra-tion efficiencies (1) Elemental iodine N/A C. Recirculation system parameters (1)' Flow rate N/A (2) Mixing efficiency N/A III. Dispersion data A. EAB and LPZ distances (meters) 311/3,220 B. X/Qs (sec/m )

EAB (0-2 hr) 1.36E-03 LPZ (0-8 hr) 2.50E-05 "

(8-24 hr) 1.75E-05 (1-4 days) 7.80E-06 (4-30 days). 2.45E-06 IV. Dose data A. Method of dose calculation Regulatory Guide 1.25 B. Dose conversion assumptions Regulatory Guide 1.25

' [') - C. Doces and Dose Rates Table III.E .2-2

()

- III.E ,. _ _

I t

TABLE III'.E.2-2

) FUEL HANDLING ACCIDENT RADIOLOGICAL CONSEQUENCES Whole,yody Dose, rem Skin Dose, rem HVAC 10CFR100 100FR100 Scenario EAB LPZ Limit EAB LPZ Limit

• RBSVS 1.14E-07 1.22E-08 2.50E+01 9.90E-06 1.06E-06 3.00E+02 Operates Maximun (t = 0) Dose Rates, arem/hr Whole Body Gamma Skin Tech. Spec Tech. Spec.

EAB LPZ Limit EAB LPZ Lieit 6.10E-05 1.12E-06 5.70E-02 5.301-03 9.74E-05 3.42E-01 Whole Body Dose, rem Skin Dose, rem

! RBNVS 10CFR100 10C'ET66 Operates EAB LPZ Limit EAB LPZ Limit

• 1.74E-06 3.22E-08 2.50E+01 1.52E-04 2.80E-06 3.00E+02 1

Mav; mum (t = 0) Dose Rates, mrem /hr

(}

Whole Body Gamma Skin Tech. Spec Tech. Spec

'EAB LPZ Limit EAB __LPZ Limit 4.79E-03 8.81E-05 5.70E-02 4.17E-01 7.66E-03 3.42E-01 Whole Body Dose, rem Skin Dose, rem Puff IDCFR100 10CFR100 Release EAB LPZ Limit EAB LPZ Limit

• d 1.75E-06 3.22E-08 2.50E+01 1.52E-04 2.80E-06 3.00E+02 Maximum (t *, 0) Dose Fates, arem/hr Whole Body , Gamma Skin Tech. Spec Tech. Spec

EAB LPZ Limit EAB LPZ Limit 8.75E-04 1,61E-05 5.70E-02 7.61E-02 1.40E-03 3.42E-01

,

• The skin dose limit ie set equal to the thyroid limit.

(

l-

- III.E . . - . . _ . _ _ . . . .

.r.

, m- -

FIGURE III.E-1 4 2-Design ' Basis . Fuel Handling Accident Exclusion Area Boundary Results  !

{ RBMVS. HVAC Syst.em' in Operation l

t 1 .

1000 i i - -

W# Od YOOUG '

EPA PAG

.. s ,

! D ,z,4,4 ur 10CFR100 M :i e ** 1 r 8.

t I-i 11 -

r

~' '

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• • n y

\  ;

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cosa i r - ~;

i.ws-=  ;

m .-

Whole Body Skivi

\ t

\ . - . . . - . _ . . . . - . _ , , _ . , _ _ _ _ _ _ _ , . _ _ _ . _ _ _ _ _ =. . , _ , ._

FIGURE"III.E-1A

~

T Fuel Handling Accident:

Exclusion Area.~ Boundary Results

!' RBNVS HVAC System in Operation

, 1000 ;_

=

~

Calculated .i 100

!! Tech Spec E

8 '

g  :: 2.ne ap --

mE i ::

l 0.417 o.34a o.342 EE

.h 7- .os?

.01 - g .

4.79

.oot Whole Body . Skin Whole Body Skin DBA ' Worst Case'  ;

.. - _ . __ _ _ . ._ . . _ _ .. ._ _ J

^

nY r

'III.E.3 -Radwaste Tank Rupture

'The accident acenario postulated in the USAR Sections

D, 11.2.3.4.2 through 11.2.3.4.4 is considered here.

o ,

1. .A: conservative partition factor of 1.0E-03 is  !

,assuerd for all isotopic activities-listed in Table  :

III.A J, with the. exception of,H-3, for which it is assumed all activity is evolved.

2. A two hour release duration is-assumed.

3. Ground release atmospheric dispersion factors are assumed, as given in Table III.E.2-1, for the EAB.

Note that the EAB is limiting insofar.as 10CFR100 l p dose limits are concerned, because the release duration is two hours.

4. .The breathing rate of persons offsite is assumed to be 3.47E-04 cubic meters per second, consistent with Regulatory Guides 1.3 and 1.25. For ether' age groups the breathing rate was obtained frch the I L ratio of the maximum age group rates given in j Regulatory Guide 1.109 (Reference J). ,

L 1 L The' doses resulting from the analysis described above .;

are as follows: j Dose, millirem

[ [')-

Whole-Body Beta Maximum L ;(,/ - Source Gamma

• Skin Organ **

o l i

0 ' Spent Resin 1.8E-05 2.7E-06 1.3E-03 h Tank 1

Radwaste Filters 1.2E-07 1.7E-08 8.3E-06 Discharge Sample 3.1E-08 1.4E-08 7.7E-06 Tanks e~ .

3 o Totals 1.8E-05 2.8E-06 1.3E-03 I l.

~

The consequences of the above postulated accident are

clearly very low. The whole body gamma, skin, and Q thyroid doses are 7.2E-08, 9.3E-10, and 4.3E-07%,

l~ respectively, of the 10CFR100 dose limits. Furthermore, these projected doses are far.below those which justify

. Quality Group D, non-seismic. qualification of radwaste equipment (i.e., 500 mrem whole body, or its equivalent to parts of the body).

J l

• External & internal pathways; child is the limiting

. f- age group

(' ** Teen is the limiting age group, and lung is the 11riting organ b

=== w ea

i

. c. ,

?, '

I

, III.E.4 WORST' CASE TUEL DAMACE EVENT Sr~j.

Scenario.

Several " worst case", extremely conservative scenarios were examined. Specifically, for the three reactor +

building BVAC cases analyzed in Section III.E.2 (RBSVS operating, RBNVS operating, and. puff release), instead of assuming the gap activity from 125 fuel rods to released (2.52 Ci Kr-85), it is assumed that all gaseous activity from the entire core in the spent fuel pool is released-(1.56E+03 Ci~Kr-85). This can only occur if all the fuelfis postulated to be mechanically, damaged and there is a complete release of gaseous isotopes. '

The assumption of.a complete release o' the gaseous inventory is also very. conservative with respect to th:a Regulatory Guide 1.25 assumption of a 30% release ,'

fraction.given the low burnup condition of Shoreham spent fuel.- Doses and dose rates are thus a factor of 617 higher than for the corresponding Regulatory Guide 1.25 cases.

All other conditions and parameters indicated in Table III.E.2-1 apply to these cases. Results are given in -

Table III.E.4-1.

Discussion

}

Even with the highly conservative release quantity.

postulated above, the calculated whole body and skin i 0- ' doses at the EAb and LPZ are very small' fractions pl

( < 0. 031%) of the 10CFR100 dose limits. Results are E -graphically shown in. Figure III.E-2. Dose rates for the postulated worst case scenario are above current

' Technical Specification limits (see. Figure _III.E-1A),

but the duration of the high dose rates-in the RBNVS and ,

L puff release cases is quite short (two hours or less).

L l

l:

l l

l l:

' \_[) - III.E l~

h

. o e

.c

• l TABLE III.E.4-1 .

y

" WORST CASE" FUEL DAMAGE ACCIDENT RADIOLOGICAL CONSEQUENCES Whole Body Dose, rem Skin Dose, rem I

EVAC 10CFR100 10CFR100

Scenario EAB LPZ Limit EAT LPZ Limit RBSVS 7.03E-05 7.50E-06 2.50E+01 6.11E-03 6.52E-04 3.00E+02 l Operates .,

1 Maximum (t= 0 Dose Rates, arem/hr l-l Whole Body Gamma Skin Tech. Spec Tech. Spec EAB LPZ Limit EAB LPZ Limit .,

l 3.76E-02 6.92E-04 5.70E-02 3.27E+00 6.01E-2 3.42E-01 l

I Whole Body Dose, rem Skin Dose, rem RBNVS Operates 10CFR100 10CFR100 EAB LPZ Limit EAB LPZ Limit 1.08E-03 1.99E-05 2.50E+01 9.25E-02 1.73F-03 3.00E+02

) Maximum (t = 0) Dose Rates, arem/hr Whole Body Gamma Skin Tech. Spec Tech. Spec EAB LPZ Limit EAB LPZ Limit I

l 2.96E-00, 5.44E-02 5.70E-02 2.57E+02 4.73E+00 3.42E-01 Puff Whole Body Dose, rem Skin Dose, rem Release 10CFR100 10CFR100 EAB LPZ Limit EAB LPZ Limit L 1.08E-03 1.99E-05 2.50E+01 9.39E-02 1.73E-03 3.00E+02 Maximum (t = 0) Dose Rates, arem/hr Whole Body Gsame Skin l Tech. Spec Tech. Spec EAB LPZ Limit EAB LPZ Limit 5.40E-01 9.93E-03 5.70E-02 4.70E+01 8.63E-01 3.42E-01

• Skin dose limit set equal to thyroid limit

- III.E o

1-.

o

^=

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t I'

.j

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FIGURE III.F-2

~

j Worst Case Fuel Damage Accident.

l Exclusion Area Boundary Results RBNVS HVAC System in Operetten -l tooo i see.. .

#6 Cokmleted Value

'co i r

~

g ,

~~

VM ,oem,oo we l ** ! !

l

- ~

~

.j me !  !  !

3.

I t

.aos !  ! q 1

i

,gg , _ . . . . .

whose sody skin

- 4 l

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~. - - .. . ~ . - . . -_ . . - . .

i

'.}

I'... . IV. REFERENCES M

ggs .,,) -Ceneral  !

Updated Safety Analysis Report (USAR) Shoreham Nuclear Power Station Revision 1. December 1987.

Section III

~

A. ORIGEN2,. Isotope Generation and Depletion Code,  ;

9/78.

B. LILCO calculation C-RPD-476, rev 0, 10/21/88, f C. LILCO calculation C-RPD-530, rev 0, 5/19/89.

D-D. " Semiannual Radioactive Effluent Release Report -

r First and Second Quarter 1989", transmitted by- " !

p letter SNRC-1619, 8/29/89.

l E. LILCO calculation C-RPD-529, rev 0, 6/7./89 -

l. .

F. Gamma Spectrometer-Scan of Reactor Coolant Sample, 6/5/89. '

G. "SNPS HIC Package Data for November / December 1988",

f~ 6/5/89, Memorandum-L. Hall to T. Gillett S/b E. Memo, P. Lynch to M. Beer, " Transmittal of Data for Dose: Projection", S/16/89.

I. Gamma Spectrometer ~ Scan of Floor Drain Collector Tanks, Waste Collector Tanks, and Recovery Sample Tanks, 6/15/89, Memorandum M. Ma to T. Gillett. ..

J. LILCO calculation C-RPD-534, rev 0, 6/22/89.

K. LILCO calculation C-NAD-532, rev 0, 6/26/89.

i~

~

L. LILCO calculation HPG-036177, rev 0, 9/15/89.

i I - IV.  ;

,. - . _ , - -.