ML20236M978

Safety Evaluation Supporting Amend 144 to License NPF-38

ML20236M978
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Site:	Waterford
Issue date:	07/10/1998
From:	NRC (Affiliation Not Assigned)
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UNITED STATES j

j NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 30666-0001 l

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATEDTO AMENDMENT NO.144 TO FACILITY OPERATING LICENSE NO. NPF-38 ENTERGY OPERATIONS. INC.

WATERFORD STEAM ELECTRIC STATION. UNIT 3 DOCKET NO. 50-382

1.0 INTRODUCTION

By application dated March 27,1997, as supplemented by letters dated April 3, July 21, October 23, November 13, and December 12,1997, January 21, January 29, March 23, May 1, May 19, May 21, May 28, and June 12,1998, Entergy Operations, Inc. ( EOl, the licensee),

submitted a request for changes to the Waterford Steam Electric Station, Unit 3, ( Waterford 3)

Technical Specifications (TSs). The requested changes would reflect the removal of the existing fuel storage racks from the spent fuel pool (SFP) and the installation of the new higher density racks in the SFP. The new racks would accommodate an increase in spent fuel assemblies beyond the existing storage capacity of the spent fuel pool and would allow an increase in the maximum initial nominal enrichment of stored fuel from 4.9 weight percent (w/o) U-235 to 5.0 w/o U-235.

The Waterford 3 SFP currently contains 1088 storage cells in 16 spent fuel racks and full core off-load capability would be lost in the year 2000. Under the proposed raracking, the 16 existing racks, which contain Boraflex as the neutron absorber, would be removed and replaced by new high density modules. The new racks are grouped into two design styles, designated as Region 1 and Region 2. Both rack styles will contain Boral as the neutron absorber. Region 1 racks will be located in the cask storage pit and will have the capability to store 255 fuel assemblies in four separate rack modules. Region 2 will have the capability to store 1849 fuel assemblies in 16 separate rack modules in the spent fuel pool and, after permanent plant shutdown,294 fuel assemblies in five separate modules in the refueling canal. The new design will provide a full core off-load discharge capability through the end of Cycle 19 (Year 2018).

The fresh fuel storage racks have not been analyzed for the enrichment increase. Waterford 3 procedure RF-002-001, " Refueling Procedure Fuel Receipt," currently states that new fuel assemblies shall not be placed in the new fuel storage area. The higher enriched fresh fuel will initially be stored in the Region 1 spent fuel storage racks. Future revisions to this procedure will be controlled by 10 CFR 50.59 procedures.

The July 21, October 23, and December 12,1997, January 21, January 29, March 23, May 1, May 19, May 21, May 28, and June 12,1998, letters provided additional information that did not change the initial proposed no significant hazards consideration determination.

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2.0 EVALUATION 2.1 Criticaldy The analysis of the reactivity effects of fuel storage in the Waterford 3 spent fuel pool was performed with the three-dimensional NITAWL2-KENO 5a computer code using the 238 group

~ SCALE cross section library. Independent verification calculations were performed with the MCNP4a three-dimensional Monte Carlo code. Since the KENO 5a code package does not have bumup capability, depletion analyses were made with the two-dimensional integral transport theory code CASMO3. CASMO3 was also used for the determination of small reactivity increments due to manufacturing tolerances. These codes are widely used for the analysis of fuel rack reactivity and have been benchmarked against results from numerous critical experiments.' These experiments simulate the Waterford 3 spent fuel racks as realistically as possible with respect to parameters important to reactivity such as enrichment, assembly spacing, and neutron absorber worth. A sufficient number of neutron histories (at least 2 million) were accumulated in each calculation to minimize the statistical uncertainty of the KENO 5a calculations. The staff concludes that the analysis methods used ars acceptable and capable of predicting the reactivity of the Waterford 3 storage racks with a high degree of confidence.

The criticality analyses were performed with several assumptions that tend to maximize the rack reactivity. These assumptions included (1) assuming unborated moderator at a temperature (4*C) that results in the highest reactivity, (2) neglecting radial neutron leakage (except in certain calculations where neutron leakage is inherent), (3) neglecting neutron absorption in minor structural members such as spacer grids, and (4) assuming the most conservative operating conditions for the depletion calculations. The design basis fuel assembly was a Combustion Engineering (CE) 16x16 assembly containing UO at a maximum initial enrichment of 5.0 w/o j

2 U-235.

i Based on the above, the staff concludes that appropriately conservative assumptions were made.

Generai Design Criterion 62 of Appendix A to 10 CFR Part 50 states that criticality in the fuel storage and handling system shall be prevented by physical systems or processes, preferably by use of geometrically safe configurations. This requirement is met by conforming to the Nuclear Regulatory Commission (NRC) acceptance criterion for criticality, which states that the effective neutron multiplication factor (1 ) in the spent fuel pool storage racks, if fully flooded by unborated 9

water, shall be no greater than 0.95, including uncertainties at a 95/95 probability / confidence level.

l For the Region 1 analysis, biases due to the calculational method, the enrichment extrapolation from critical experiments, boron particle self-shielding, and, where applicable, a temperature correction to 4'C were included. Uncertainties due to the KENO 5a statistics and the KENO 5a j

method were statistically combined with mechanical tolerance uncertainties. These uncertainties were appropnately determined at least at the 95/95 probability / confidence level. These biases and uncertainties meet the previously stated NRC requirements and are, therefore, acceptable.

The licensee's analysis of the Region 1 storage racks showed that fuel with a nominal 5.0 w/o U-235 enrichment (maximum 5.0 + 0.05) results in a maximum k, of 0.9284 including biases and 95/95 uncertainties.

. For the Region 2 analysis, biases due to the calculational method and uncertainty in the depletion calculations were included. Uncertainties due to the KENO 5a statistics and the KENO 5a method, as well as bumup, where applicable, were statistically combined with mechanical tolerance uncertainties. Those uncertainties were appropriately determined at least at the 95/95 probability / confidence level.- These biases and uncertainties meet the previously stated NRC requirements and are, therefore, acceptable.

The concept of bumup reactivity equivalencing was also used in order to store fuel with nominal enrichment up to 5.0 w/o U-235. This concept is based on the reactivity decrease associated with fuel depletion and has been previously found acceptable by the NRC for use in pressurized water reactor (PWR) fuel storage analysis. A series of reactivity calculations is performed to generate a set of enrichment versus bumup ordered pairs which yield an equivalent y for fuel stored in the Waterford 3 racks. The results of the bumup reactivity equivalencing shows that fuel with an initial U-235 enrichment of 5.0 w/o and irradiated to 42 MWD /KgU results in a maximum Q of 0.9450 including biases and 95/95 uncertainties.

In addition, calculations were made for checkerboard configurations. Fresh fuel with a 5.0 w/o U-235 enrichment checkerboarded with empty cells results in a maximum 4 of 0.8309, including biases and 95/95 uncertainties. Fuel with an initial 5.0 w/o U-235 enrichment and irradiated to 27 MWD /KgU checkerboarded with fuel assemblies irradiated to 55 MWD /KgU results in a maximum 4 of 0.9480, including biases and 95/95 uncertainties.

The results of these analyses, using the acceptable methods discussed above, meet the NRC criterion of 4 no greater than 0.95, including all uncertainties at the 95/95 probability / confidence level and are, therefore, acceptable. The results show that Region 1 can accommodate fuel with enrichment as high as 5.0 w/o U-235 without restriction. Storage of fuel assemblies with initial enrichment of 5.0 w/o U-235 in Region 2 requires either fuel bumup of at least 42 MWD /KgU or placement in storage locations which have face adjacent storage cells containing water. Fuel with initial enrichment of 5.0 w/o U-235 irradiated to 27 MWD /KgU can ce stored in Region 2 if checkerboarded with fuelirradiated to 55 MWD /KgU.

Although not included in the bumup dependent criticality analyses, subsequent decay of Pu-241 and growth of Am-241 with long-term tbrage results in a significant decrease in reactivity. This will provide an increasing subcriticality margin and further compensate for any uncertainty in the depletion calculations.

Most abnormal storage conditions will not result in an increase in the 4 of the racks. However, it is possible to postulate events, such as the inadvertent mistoading of an assembly with a bumup and enrichment combination outside of the acceptable areas in TS Figures 5.6-1,5.6-2, or 5.6-3, which could lead to an increase in reactivity. However, for such events, credit may be taken for the presence of at least 1720 parts per million (ppm) of soluble boron required in the pool whenever a fuel assembly is moved, since the staff does not require the assumption of two unlikely, independent, concurrent events to ensure protection against a criticality accident L

'(Double Contingency Principle). The reduction in 4 caused by the boron more than offsets the reactivity addition caused by credible accidents. In fact, calculations show that for the most severe accident condition, a soluble boron concentration of 700 ppm boron would be adequate to maintain 4 less than 0.95.

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Reactivity calculations were also performed for the fuel transfer carriage, which is capable of accommodating two fuel assemblies at a time, and the temporary storage racks located in containment, which are capable of accommodating five fuel assemblies in a linear arrangement on an 18-inch spacing. Both no mal and abnormal conditions result in 4 values less than 0.95, thus meeting the regulatory guidelines.

In July 21,1997, letter the licensee clarified thet the word " normal" was added in TS 5.6.1(a) to differentiate between normal operation when 4 is less than 0.95 with unborated water and accident conditions when credit for soluble boron in the water is allowed. This is acceptable to the staff.

The following TS changes have been proposed as a result of the requested spent fuel pool reracking. Based on the above evaluation, the staff finds these changes acceptable as well as the associated Bases changes.

(1) TS 5.3.1 would change the maximum nominal enrichment from 4.9 weight percent U-235 to 5.0 weight percent U-235.,

(2) In TS 5.6.1(a) the word " normal" will be added to differentiate between normal operation and accident condition 4 l

(3) TS 5.6.1(b) would change the nominal storage center-to-center distance of 10.38 inches between fuel assemblies in the spent fuel storage racks to 10.185 inches between fuel assemblies placed in the Region 1 (cask storage pit) spent fuel storage racks.

(4) TS 5.6.1(c) would be added to state a nominal 8.692 inch center-to-center distance between fuel assemblies placed in the Region 2 (spent fuel pool and refueling canal) spent fuel

' storage racks, except for the four southem-most racks in the spent fuel pool, which would have an increased center-to-center distance of 8.892 inches in the north-south direction.

(5) TS 5.6.1(d) would be added to state that new or partially spent fuel assemblies may be allowed unrestricted storage in Region 1 racks.

(6) TS 5.6.1(e) would be added to state that new fuel assemblies may be stored in the Region 2 racks provided that they are stored in a checkerboard pattem as illustrated in TS Figure 5.6-1.

(7) TS 5.6.1(f) would be added to state that partially spent fuel assemblies with a discwge bumup in the acceptable region of TS Figure 5.6-2 would be allowed unrestricted storage in the Region 2 racks.

l (8) TS 5.6.1(g) would be added to state that partially spent fuel assemblies with a discharge l

bumup in the unacceptable range of TS Figure 5.6-2 may be stored in the Region 2 racks provided that they are stored in a checkerboard pattem as illustrated in TS Figure 5.6-1, with spent fuel in the acceptable range of TS Figure 5.6-3.

(9) TS 5.6.2 would be revised to reference the new (fresh) fuel storage racks and both the 4 less than or equal to 0.95 criterion when flooded with unborated water and the 4 no greater than 0.98 criterion when aqueous foam moderation is assumed.

. (10)

TS 5.6.3 would be revised to add the statement that when fuel is being stored in the cask storage pit and/or the refueling canal, these areas will also be maintained at +40.0 MSL (mean sea level).

(11)

TS 5.6.4 would be revised to increase the storage capacity from 1088 fuel assemblies to 1849 assemblies in the main pool,255 assemblies in the cask storage pit, and after permanent plant shutdown,294 assemblies in the refueling canal.

2.1.1 Criticality Conclusion Based on the review described above, the staff finds the criticality aspects of the proposed increase in the storage capacity of the Waterford 3 spent fuel pool storage racks are acceptable and meet the requirements of General Design Criterion 62 for the prevention of criticality in fuel storage and handling.

Although the Waterford 3 TS have been modified to specify the above-mentioned fuel as acceptable for storage in the spent fuel racks, evaluations of reload core designs (using any enrichment) will, of course, be performed on a cycle by cycle basis as part of the reload safety evaluation process. Each reload design is evaluated to confirm that the cycle core design adheres to the limits thei exist in the accident analyses and TS to ensure that reactor operation is acceptable.

2.2 Soent Fuel Pool Coolina and Cleanuo System J

The spent fuel pool cooling and cleanup (SFPCC) system is designed to remove the decay heat from spcnt fuel assemblies (SFAs) stored in the SFP, and to clarify and purify the water in the i

SFP. The SFP cooling portion of the SFPCC system is a seismic Category I closed loop system consisting of two half capacity pumps and one full capacity heat exchanger. Heat is removed from the SFP heat exchanger by the component cooling water system. A backup fuel pool heat exchanger with a lower heat removal capacity is available for use when the SFP primary heat exchangeris out of service' or in an emergency when the temperature monitor system alarm goes off due to high SFP temperature. The SFP cooling loop is designed to:

i a) Maintain the SFP water temperature during planned refueling outages at or below 140'F with one SFP pump operating to remove decay heat from approximately 45% of a core discharged to the SFP in addition to the decay heat from 11 previous refueling batches, and b) Maintain the SFP water temperature during a full core off-load outage at or below 155'F with two SFP pumps operating to remove decay heat from a full core discharge in addition to the decay heat from 10 previous refueling batches.

I The plant system operating procedure, " Fuel Pool Cooling and Purification", has

provisions established to ensure that the SFP primary heat exchanger will not be taken out of service during plant operation whenever the SFP heat load is higher than the backup heat exchanger design heat removal capacity of 15.4x10' Btu /hr.

. The proposed increase in SFA storage capacity will result in an increase of SFP cooling loop heat load for any specific fuel discharge scenario. The licensee re-performed the thermal-hydraulic analyses to evaluate the effects of increased SFA storage capacity on the SFP cooling loop heat loads and SFP water temperatures. The following summarizes the SFP heat exchanger loads and their corresponding peak temperatures in the SFP, cask storage pit and refueling canal for the proposed SFA storage capacity of 2398 SFAs (2485 SFAs were l

conservatively assumed in the thermal hydraulic analysis):

l Case in-core Max. Pool Temp.

Coincident Coincident Hold Time

('F)

Time (Hrs Hx Loads (hrs) after (MBlu/Hr)

SFP Cask Refuel shutdown) l Pit Canal 1

72 139.41 108 33.73 2

72 151.61 155.5 158.5 131 50.41 I

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Case 1:

Planned Refueling - Based on 2369 SFAs stored in the SFP prior to the last cycle, a batch of 116 SFAs being discharged at a rate of 4 SFAs per hour into the SFP 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after reactor shutdown and cooling the SFP using 1 pump and the primary heat exchanger.

1 Case 2:

Unplanned Full Core Off-load - Based on 2268 SFAs stored in the SFP and cask pit areas at the end of the final cycle of plant operation, a full core of 217 SFAs being transferred at a rate of 4 SFAs per hour into the SFP 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> after reactor 8

shutdown and cooling the SFP using 2 pumps and the primary heat exchanger.

For the SFP heat load during the planned refueling outage (Case 1), the increase in SFA storage capacity will result in an increase in the SFP heat load from 22.2x10' Btu /hr to approximately 33.73x10' Btu /hr. The previously calculated 22.2x10 Btu /hr SFP heat load was based on a 8

partial core of 96 SFAs discharged five days after reactor shutdown plus 992 SFAs (total of 1088 SFAs) stored in the SFP. The SFP heat load resulting from the proposed SFA storage capacity is significantly higher because it is based on a reactor shutdown time of only three days for the partial core discharge of 116 SFAs plus 2369 SFAs stored in the SFP. In addition, the design basis values (enrichment and bumup) that were used in the rerack thermal-hydraulic analysis are also much more conservative (higher) than the values that were used for the current analysis.

The peak SFP water temperature resulting from this heat load and with a single active failure is 139.41*F, which is below the design basis and the guidance of Standard Review Plan (SRP)

Section 9.1.3 for SFP water temperature during planned refueling.

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Waterford 3 does not routinely conduct full core off-load during planned refueling outages.

The bulk temperature of the combined three regions for an unplanned full core off-load with a single active failure of one SFP cooling pump is 163.49'F.

. For an unplanned full core off-load, the combined heat load in the SFP, cask pit storage pit and refueling canal for the proposed SFA storage capacity is 50.41x10e Stu/hr, which is lower than 8

the previously calculated SFP heat load of 55.7x10 Stu/hr for the current storage capacity. In the thermal-hydraulic analysis for the current SFA storage capacity of 1088 SFAs, a full core of 217 SFAs were assumed to be instantaneously discharged to the SFP three days after reactor shutdown. In the thermal-hydraulic analysis for the proposed SFA storage capacity of 2398 SFAs,217 SFAs were assumed to be discharged at a rate of four SFAs per hour starting at three days after reactor shutdown. Thus, average SFAs will have longer cooling time in the core prior to being discharged. The longer in reactor cooling time used in the thermal-hydraulic analysis for the proposed SFA storage capacity also outweighs the higher bumup and enrichment values which would, otherwise, result in a higher heat load. The staff has reviewed the input parameters used in the thermal-hydraulic analysis for the proposed SFA storage capacity and finds them acceptable.

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Also, in the thermal-hydraulic analysis for an unplanned full core off-load for the proposed SFA storage capacity, the licensee assumed that: (a) 217 freshly discharged SFAs and 38 previousig discharged SFAs were stored in the cask storage pit providing a bounding heat load of 42.5x10 8

Stu/hr in the cask storage pit; and (b) only old fuel (294 SFAs generating 1.72x10 Btu /hr) from the SFP were stored in the refueling canal. The licensee has proposed the TSs to ensure that 8

the heat load in the refueling canal will not exceed 1.72x10 Btu /hr. The staff finds the proposed TSs acceptable.

The calculated peak bulk temperatures for the proposed SFA storage capacity resulting from a full core off-load with two SFP pumps operating are 151.61*F,155.50*F and 158.50*F in the SFP, cask storage pit and refueling canal, respectively. The licensee also performed an analysis to determine the bulk temperature of the combined three regions for a full core off-load assuming

- a single active failure of one of the two SFP cooling pumps. The calculated peak bulk temperature is 163.490*F. These temperatures are well below the guidance of SRP Section 9.1.3 for the SFP water temperature limits (pool boiling) during an unplanned full core off load outage.

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As indicated before, the licensee has proposed to rerack the SFP with high density storage racks, install new racks in the cask loading pit, and install additional storage racks in the refueling canal for use after permanent plant shutdown.

Natural circulation of spent fuel pool coolant will be relied upon to maintain the design temperatures in the cask pit and refueling canal regions. These regions are connected to the spent fuel pool by gates that are isolated except during refueling operations. The licensee stated i

l that the temperatures in these regions will equalize within a few degrees when the gates are L

removed and natural circulation between the regions occurs. Decay heat from the stored fuel in the refueling canal and cask pit will heat the coolant causing it to flow to cooler regions of the spent fuel storage system where it will be removed by the spent fuel pool cooling system. The

' licensee will prevent isolating stored fuel assemblies from the natural circulation flow with administrative controls.

The licensee used FLUENT 4.32, a general purpose computational fluid dynamics code (CDF), to determine whether adequate cooling was provided by natural circulation in the cask pit and refueling canal. The analysis modeled the spent fuel pool, cask pit and refueling canal in two dimensions. A single FLUENT run was submitted to support the analysis.

. The staff performed an independent evaluation of the licensee's calculations to ensure the results were conservative. The staff used FLUENT /UNS 4.2, an unstructured version of the FLUENT code, for the independent evaluation. The code uses a different algorithm from the version used in the licensee's analysis. This version of FLUENT allowed the staff to make modeling improvements for the description of the facility, computational grid, and modeling options. The staff performed analyses to verify that the licensee's analysis was representative of j

the thermal-hydraulic system, and to ensure that the results were bounded by the spent fuel storage system design. The staff performed: 1) a replication of the Hollec analysis using l

FLUENT 4.32,2) a First Law of Thermodynamics hand calculation,3) an independent two

{

l dimensional analysis using revised modeling and FLUENT /UNS 4.2,4) sensitivity studies using l

ihe two dimensional model, and 5) a three dimensional model analysis using FLUENT /UNS 4.2 i

The results of the staffs calculations indicate that, although bulk coolant temperature and hot channel temperature are slightly higher than those predicted in the licensee's model, the bulk L

coolant temperature in excess of the temperature limit of 150'F for concrete structure described in the American Concrete Institute code (ACl-349) is of short duration. While the staff identified concems with the licensee's assumptions and their use of a two dimensional analytical program to estimate temperatures in a three-dimensional system, the results of the licensee's analysis l

were comparable to the results of the staff's analysis.

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The licensee stated that prior to each refueling or felt core off-load outage, an engineering review of the Reload Safety Analysis Ground Rules Document will be routinely performed to ensure that the limiting parameters (i.e.,2485 SFAs, fuel bumup, peaking factors, etc.) used in the thermal-hydraulic analyses for rerack are not exceeded.

In addition, the SFP has a water temperature monitoring system, which alarms in the control l

room when the SFP water temperature reaches 135'F. In the event that the alarm goes off due to high SFP temperature, plant procedure, " Spent Fuel Pool Cooling Malfunction," and site directive, " Corrective Action," list the probable causes and corrective actions (i.e., to start the second SFP pump, operating two heat exchangers in parallel, stop fuel movement, etc.) to be taken. This will provide addit lonal s';surance that the above limitations (140'F during planned refueling outages and 155'F dunng unplanned full core off-loads) are not exceeded.

Based on its review and the license's provisions described above, the staff finds that the design I

and operation of the SFPCC system meets the intent of the SRP for SFPs. The staff, therefore, l

concludes that the peak calculated water temperatures in the SFP, cask storage pit and refueling canal during routine refueling or unplanned full core off-load outage resulting from the increase of l

SFA storage capacity at Waterford 3 are acceptable.

2.2.1 Effects of SFP Boilina in the event that there is a complete loss of cooling capability of using SFP heat exchangers to remove heat from the SFP, the SFP water temperature will begin to rise and eventually will reach

' the boiling temperature.

The licensee performed analysis to demonstrate the time to boil and the boil-off rate. The calculated minimum time from the loss-of-pool cooling at peak SFP temperature until the pool boils based on the heat load for the full core off-load is 2.89 hours0.00103 days <br />0.0247 hours <br />1.471561e-4 weeks <br />3.38645e-5 months <br /> with a maximum boil-off rate of 96.48 gpm. The licensee stated that this boil-off rate of 96.48 gpm would not result in the spent fuel being uncovered until after an additional 34 hours3.935185e-4 days <br />0.00944 hours <br />5.621693e-5 weeks <br />1.2937e-5 months <br /> which is 168 hours0.00194 days <br />0.0467 hours <br />2.777778e-4 weeks <br />6.3924e-5 months <br /> after reactor shutdown.

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This duration allows sufficient time for the operators to intervene. Makeup water for the SFP can l

be provided from either the refueling water storage pool (via the refueling water storage pool l

purification pump at 150 gpm), the condensate storage pool (via the component cooling water l

makeup pumps at 600 gpm) or the fire water system (at approximately 100 gpm).

Based on its review, the staff finds that cooling the SFP by allowing the SFP to boll and by I

adding makeup water in the event of a complete loss of cooling capability (using SFP heat exchangers to remove heat from the SFP) conforms with the guidance described in the SRP and, therefore, is acceptable.

2.2.2 SFP Coolina System Conclusion Based on its review of the licensee's rationale, the staff's confirmatory analysis, and licensee's assurance that the plant's Final Safety Analyses Report (FSAR) will be updated and procedures and/or administrative conkols will be established to reflect the above information regarding the SFP and to control SFAs being discharged to the SFP, cask storage pit and r3 fueling canal, the staff concludes that the proposed TS changes to increase the SFA storage coacity from 108S to 2398 SFAs at Waterford 3 with respect to the SFP cooling capacities are acceptable.

2.3 Handlina of Heavy Loads and Soent Fuel Assemblies Reracking to increase the plant's spent fuel pool storage capacity involves the handling of heavy loads, including removal of the existing spent fuel storage racks and installation of new high density racks in the SFP; installation of a support platform and high density spent fuel storage racks in the cask storage pit; removal and installation of the gate enclosing the north side of the cask storage pit; and after permanent plant shutdown, installation of high density storage racks in the Refueling Canal. The licensee also plans to install an offset fuel handling tool to reach and move spent fuel assemblies to storage locations that are adjacent to the pool walls.

2.3.1 Technical Specification and Bases 3/4.9.7 Crane Travel-Fuel Handlina Buildina The proposed changes to the Limiting Condition of Operation (LCO), the Surveillance Requirements (SRs), and the Bases clearly show that heavy loads (loads in excess of 2000 pounds) will be prohibited from travel over irradiated fuel assemblies in the Fuel Handling Building. The present TS prohibits the transfer of heavy loads over fuel assemblies in the spent fuel pool only. The proposed changes will ensure that only the spent fuel handling machine will be used for handling irradiated fuel assemblies in the fuel handling building. These changes correspond to the licensee's proposal to use the areas within the Fuel Handling Building, including the spent fuel storage pool, the cask storage pit, and the refueling canal, to optimize its storage of irradiated fuel, and transport fuel assemblies according to the guidelines in NUREG-0612, " Control of Heavy Loads at Nuclear Power Plants." Therefore, the proposed changes to the TS are acceptable to the staff.

2.3.2 Evaluation of Postulated Load Droo Accidents in the proposed rerack, the licensee evaluated the consequences of an accidental drop of a spent fuel assembly from the highest lifting point of the fuel handling machine unto the top of the fuel racks and into the spent fuel pool. The licensee also evaluated the consequences of an accidental drop of a fully loaded spent fuel shipping cask into the cask storage pit, and a spent fuel rack into the SFP Based on the evalua' ions, the licensee found that the consequences of l

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I I an accidental drop of a spent fuel assembly, a spent fuel cask, or a rack would not exceed the guidelines in NUREG-0612 Section 5.1,

• Recommended Guidelines."

The licensee determined that a dropped fuel assembly would cause releases of radioactive material that would produce doses that are below the 10 CFR Part 100 limits. Damage to fuel would not result in an increase in the suberiticality margin and k,, would remain at less than or equal to 0.95; damage to the spent fuel pool would not result in water leakage that could uncover the fuel; and damage to safe shutdown systems is not credible. Therefore, the guidelines in NUREG-0612 would be satisfied.

The evaluation of a postulated cask drop into the cask storage pit revealed that the pool walls and liner would sustain minimum damage, including some local crushing of the concrete and limited damage to the liner plate. Minimum water leakage would occur through the leak chase system; however, it would be collected at the sump. To offset the loss of wat.er due to leakage, makeup water would be obtained from the condensate storage pool and/or the refueling water storage tank. The licensee also found that the integrity of the Fuel Handling Building would not be compromised; therefore, radioactive releases to the environment would not occur. Thus, the result of the evaluation is that the consequences of a dropped cask would not exceed the guidelines in NUREG-0612.- The staff finds that this is acceptable.

Based on its evaluation, the licensee stated that a dropped rack is unlikely to have radiological consequences because removal and installation of the racks involve the use of safe load paths that prevent heavy loads from being transported over stored spent fuel and over safety systems.

A dropped rack would also result in limited damage to the pool liner, which, according to the licensee, would be easily repairable. The consequences of a dropped rack also satisfy the guidelines of NUREG-0612 and are acceptable to the staff.

2.3.3 Evaluation of Holstina Systems The licensee stated that the operability of all cranes and lifting devices will be verified before the reracking operation. The rigging and Fuel Handling Building crane / hoists system would be verified for compliance with design and testing requirements of CMAA Specification No. 70,

- ANSI B30.2," Overhead and Gantry Cranes," ANSI B30.9," Slings," and ANSI B30.11 " Monorail

- Systems and Underhung Cranes." In addition, the licensee will use the procedures that assure compliance with NUREG-0612. The staff finds that these provisions are acceptable.

l 2.3.4 Heavy Loads and Soent Fuel Assemblies Handlina Conclusion Based on the preceding discussion, the staff finds that the changes to the TS and Bases are adequate. We also agree with the licensee that the consequences of postulated load drops involving spent fuel storage rack, spent fuel assembly, and shipping cask would satisfy the

- guidelines in NUREG-0612. We also accept the licensee's plan to verify the operability of the crane and lifting systems in accordance with the requirements for design and operation before proceding with the SFP modifications.

2.4 Structural Enaineerina This section addresses the adequacy of the structural aspects of the licensee's proposal to increase the SFP capacity at Waterford 3. The primary purpose of this evaluation is to assure the structuraiintegrity and functionality of the racks, the stored fuel assemblies and the spent fuel L

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l 1 pool structure subject to the effects of the postulated loads (Appendix D of SRP Section 3.8.4) and fuel handling accidents.

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2.4.1 Storsoe Racks I

l The 23g8 storage cells will be contained in 25 fuel storage racks, which are seismic Category 1 i

equipment and are required to remain functional during and after a safe shutdown earthquake 1

l (SSE). Among 25 storage racks,16 racks will be placed in the SFP, four racks will be placed in l

the cask storage pit (CSP) and five racks will be placed in the refueling canal (RC). All 25 racks are new, and are designed and manufactured by Holtec intemational, Inc. (Holtec). EOl with its contractor, Holtec, performed structural analyses for the racks for the requested license I

amendment.

EOl used a computer program, DYNARACK, for dynamic analysis to demonstrate the structural adequacy of the Waterford 3 spent fuel rack design under the combined effects of earthquake and other applicable loading conditions. The proposed spent fuel storage racks are free-standing and self-supporting equipment, and they are not attached to the floor of the storage pool. A nonlinear dynamic model consisting of inertial mass elements, spring elements, gap elements and friction elements, as defined in the program, was used to simulate three dimensional l

dynamic behavior of the rack and the stored fuel assemblies including frictional and hydrodynamic effects. The program calculated nodal forces and displacements at the nodes, and then obtained the detailed stress field in the rack elements from the calculated nodal forces.

1 Two model analyses were performed: the 3 D single-rack (SR) model analysis and the 3-D l

whole pool multi-rack (MR) analysis. For the 3-D SR model analysis, a 10x11 cell rack was considered for the calculation of stresses and displacements. The dimension of the rack was 8 ft (W) x 7.3 ft (L) x 15.4 ft (H), where W, L and H are defined as width, length and height of the rack, respectively.. The rack was considered fully loaded and half loaded with two different i

coefficients of friction between the rack and the pool floor (p=0.2 and 0.8) to identify the worst

- case response for rack movement and for rack member stresses and strains. In the MR model 1

analysis,16 free standing racks in the SFP were considered to investigate the fluid-structure interaction effects between racks and pool walls as well as those among the racks.

The seismic analyses were performed utilizing the direct integration time-history method. Two sets (one set for OBE and the other set for a safe shutdown earthquake (SSE)) of artificial time histories were generated from the design response spectra defined in the FSAR. Each set j

consisted of three artificial time histories (two horizontal and one vertical acceleration time histories). EOl demonstrated the adequacy of the artificial time history sets used for the seismic analyses by satisfying requirements of both enveloping design response spectra as well as matching a target power spectral density (PSD) function compatible with the design response spectra as discussed in Standard Review Plan (SRP) Section 3.7.1.

A total of eight 3-D SR analyses were performed. The results of the analyses show that the maximum vertical displacement of the racks is about 3.6 inches, assuring that there is adequate safety margin against overtuming of the racks and, thereby, the structural integrity and stability of the racks is maintained. In addition, the calculated stresses in tension, compression, bending, combined flexure and compression, and combined flexure and tension were compared with l

corresponding allowable stresses specified in ASME Boiler and Pressure Vessel Code (1986 i

edition), Section lil, Subsection NF. The results show that allinduced stresses under an SSE loading condition are smaller than the corresponding allowable stresses specified in the ASME l

. Boiler and Pressure Vessel Code, Section lil, Subsection NF indicating that the rack design is adequate.

In the 3-D MR analyses,16 fully loaded racks were considered and were subjected to the ssrvice, upset and faulted loading conditions (Level A, B and D Service Limits). The results of the MR analysis indicate that the calculated stresses on a rack are higher than those obtained from the single rack analyses. However, all calculated stresses for the MR analyses are smal!ar i

than the corresponding allowable stresses of the ASME Code.

EOl also calculated the weld stresses of the rack at the connections (e.g., baseplate-to-rack, baseplate-to-pedestal and cell-to-cell connections) under the dynamic loading conditions. EOl demonstrated that all the calculated weld stresses are smaller than the corresponding allowable stresses specified in the ASME Code Section Ill, Subsection NF, indicating that the weld connection design of the rack is adequate.

The staff concludes that the rack modules will perform their safety function and maintain their structural integrity and functionality under the service, upset and faulted loading conditions and, I

therefore, are acceptable based on: (1) the EOl's comprehensive parametric study (e.g., varying coefficients of friction, different geometries and fuel loading conditions of the rack), (2) the large safety margins of the induced stresses of the rack when they are compared to the corresponding allowables provided in the ASME Boiler and Pressure Vessel Code, Section lil, and (3) EOl's l

overall structural integrity conclusions supported by both 3-D SR and MR analyses.

l 2.4.2 Spent Fuel Storace Pool EOl analyzed the SFP, CSP and RC structures using the finite element computer program, EBS/NASTRAN, to demonstrate the adequacy of the structures under fully loaded fuel racks with all storage locations occupied by fuel assemblies. The fully loaded structures were subjected to the load combinations specified in the Waterford 3 FSAR.

Table 8.1 of Reference 1 shows the predicted factors of safety varying from 1.11 to 4.0 for shear force and bending moments of the concrete walls and stab. In view of the calculated factors of safety, the staff concludes that the Waterford 3 structural analysis demonstrates the adequacy and integrity of the structures under full fuel loading, thermal loading and SSE loading conditions.

Thus, the storage fuel pool design is acceptable.

j 2.4.3 Fuel Handlina Accident

)

The following two refueling accident cases were evaluated by EOl: (1) drop of a fuel assembly I

with its handling tool, which impacts the baseplate (deep drop scenario) and (2) drop of a fuel assembly with its handling tool, which impacts the top of a rack (shallow drop scenario).

i The analysis results of Accident Case (1) show that the load transmitted to the liner through the rack structure is property distributed through the bearing pads located near the fuel handling area; therefore, the liner would not be ruptured by the impact as a result of the fuel assembly drop through the rack structure. The analysis results of Accident Drop Case (2) show that I

damage will be restricted to a depth of 22 inches below the top of the rack, which is above the 1

active fuel region. The staff reviewed EOl's analysis results presented in the March 27,1997, letter and concurs with its findings. This.is acceptable based on the EOl's structural integrity conclusions supported by the parametric studies.

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• 2.4.4 Structural Enaineerina Conclusion Based on the review and evaluation of the EOl's submittals, the staff concludes that the EOl's structural analysis and design of the spent fuel rack modules and the spent fuel pool structures are adequate to withstand the effects of the applicable loads including that of the SSE. The analysis and design are in compliance with current licensing basis set forth in the FSAR and applicable provisions of the SRP and, therefore, are acceptable, 2.5 Material Enaineenna All the newly installed fuel racks will use Boral neutron absorber. This material is compatible with the spent fuel pool environment and the licensee expects to have very few operational problems with this material. Nevertheless, licensee is planning to develop a surveillance program for verifying the condition of Boral after its exposure to the spent fuel pool environment. This program will be implemented should there be evidence of Boral degradation in ongoing tests carried out by several utilities. The NRC staff has evaluated the information addressing the compatibility.of structural materials and Boral with the spent fuel pool environment.

The spent fuel racks to be installed in the spent fuel pool are manufactured by Holtec Intemational. They are of a free-standing, self-supporting type. The racks are designed to the stress limits and analyzed in accordance with Section Ill, Division 1, Subsection NF of the ASME Boiler and Pressure Vessel Code.

l 2.5.1 Structural Materials l

The following structural materials are used in the spent fuel racks:

Storage cell structures and intemally threaded support legs are made from 304L stainless steel, ASME Standard SA240, Extemally threaded support spindles are made from precipitation hardened stainless steel l

l (heat treated to 1100'F) ASME Standard SA564-630, and Weld materialis ASME Type SFA 5.9 ER308L.

l All these materials have been previously used in many similar applications. They have been l

exposed to environments similar to or more severe than those existing in the spent fuel pool at Wateiford 3 without experiencing any observable corrosion damage. They are, therefore, acceptable for their present application.

2.5.2 Neutron Absorbino Material in the spent fuel racks Boralis utilized as neutron absorbing material. Boralis a cermet composite material made of Type 1100 aluminum and boron. The composite panel consists of boron carbide particles imbedded in a Type 1100 aluminum matrix clad in Type 1100 aluminum l

sheets. The 1100 aluminum materialimparts sufficient pitting and general corrosion resistance l

by forming an aluminum oxide layer on its surface when exposed to oxidizing environments. The oxide is stable in environments with a pH range 4.5 to 8.5. The boron carbide particles in Boral panels have been shown to have good structural compatibility with the Type 1100 matrix material. Despite these corrosion preventive properties of Boral, some corrosion is expected.

Although this usually will not result in a significant depletion of boron and resulting degradation of its neutron absorbing properties, some generation of hydrogen from corrosion of aluminum can occur when Boral is exposed to the spent fuel pool water. This effect is more pronounced in new

l l 1 panels which have not yet well formed a protective oxide film. This hydrogen, when not vented, could cause sw6!!ing of the sheathings holding Boral panels and resultant deformation of storage cells. In order to prevent this from occurring, the racks manufactured by Holtec will have vented Boral sheathings, allowing the generated hydrogen to escape. Production of this hydrogen will significantly decrease as aluminum surfaces develop a protective oxide film.

To ensure that Boral's neutron absorbing capability is maintained, the licensee is planning to establish a Boral surveillance program which will be implemented if the tests carried out by other utilities indicate that signNnt boron carbide loss from Boralis occurring. The licensee proposed program will have 20 Boral coupons suspended from the mountings (" trees") located in two spent fuel rack cells. For conservatism, during the first four fuel offloads these coupons will be exposed to higher than average radiation fields by surrounding them with the cells carrying freshly-discharged fuel assemblies. The coupons will be removed at predetermined times and subjected to examinations. The examinations will consist of visual observation, neutron attenuation dimensional measurements and determination of weight and specific gravity. The results from these examinations will provide information on the degree of Boral degradation in ample time to take such corrective action as deemed to be necessary.

2.5.3 Material Enaineerina Conclusions Based on its evaluation, the staff finds that the materials in the spent fuel racks manufactured by Hollec Intemational are compatible with the environment of Waterford Generating Station's spent fuel pool and the racks will not undergo material degradation which could affect theirlability to safely store spent and new fuel. A vented design of the Boral sheathings will prevent the corrosion generated hydrogen from building up pressures that could cause distortion of the fuel cells and the program of Boral surveying will ensure that timely corrective actions could be taken should its degradation be detected.

2.6 Radiation Protection 2.6.1 Occupational Radiation Exposure The staff has reviewed the licensee's plan for the modification of the Waterford spent fuel racks with respect to occupational radiation exposure. As stated before, for this modification the licensee plans to remove sixteen rack modules from the SFP and replace them with sixteen high-density rack modules. The licensee will also install four new high-density rack modules in the Cask Storage Pit during the upcoming modification. The sixteen rack modules removed from the SFP will be washed down and decontaminated in preparation for packing and shipment. A number of facilities have performed similar operations in the past. On the basis of the lessons leamed from these operations, the licensee estimates that the proposed reracking can be performed for between 6 and 12 person-rem.

All of the operations involved in the re-racking will utilize detailed procedures prepared with full consideration of as low as is reasonably achievable (ALARA) principles. The Radiation Protection department will prepare Radiation Work Permits (RWPs) for the varlaus jobs associated with the SFP reracking. These RWPs willinstruct the project personnelin the areas of protective clothing, general dose rates, contamination levels, arid dosimetry requirements.

Each member of the project team will be properly trained in each of the reracking evolutions and each team memberwill be required to attend daily pre-job briefings on the scope of the work to byerformed. Personnel will wear protective clothing and will be required to wear personnel monitoring equipment.

. Each diver used to perform work in the SFP will be equipped with five remote readout radiation detectors, which will be continuously monitored by Radiation Protection personnel. The divers will be in continuous communication with the Dive Controller / Radiation Protection personnel.

l The licensee will conduct a radiation survey of the diving area prior to each diving operation and will verify the new location of fuel assemblies whenever fuel movement occurs. In order to minimize diver dose, divers will be instructed to stay at least ten feet from any spent fuel assemblies in the SFP. In order to ensure that divers maintain a safe distance from irradiated sources, divers' movements may be restricted by the use of an umbilical. The diving area will be I

well illuminated and at least one underwater camera will be trained on the diver at all times.

l Upon removal of the old fuel rack modules from the SFP, the racks will be washed with demineralized water and then held over the SFP until significant dripping has stopped. The racks will then be put into anti-contamination bags before they have a chance to dry out, thereby minimizing the chance of any airbome contamination from the suspended rack modules. The licensee does not expect the concentrations of airbome radioactivity in the vicinity of the SFP to increase due to the expanded SFP storage capacity. However, the licensee will operate a continuous air monitor in the SFP area during the en; ire reracking operation to detect any potential airbome radioactivity. In addition, the plant effluent radiation monitoring system will monitor any gaseous releases.

The licensee will monitor and control personnel traffic and equipment movement in the SFP area to minimize contamination cad to assure that exposures are maintained ALARA. The licensee has developed remote tooling,'such as lift fixtures, pneumatic grippers, a support leveling device, and a lift rod disengagement device to perform numerous activities, remotely, from the pool surface. Prior to removing any existing structure from the spent fuel pool, the licensee will use a pressure washer or other acceptable mechanism to reduce general contamination levels and to

]

remove any hot particles.

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During raracking operations, there is the potential for an increase in radioactivity concentrations J

'in the SFP due to spalling of crud from spent fuel assemblies during movement. However, past refuelings at Waterford have not resulted in a major increase in radionuclides concentrations in the SFP water.

The licensee plans to use an underwater vacuum to clean the bottom of the SFP during reracking to minimize any potential radiological effects of spalling from the fuel elements. In addition, the licensee will use the existing SFP filtration system during reracking to maintain water clarity in the SFP.

The fuel assemblies in the new storage racks will be stored in a more tightly packed arrangement and the fuel racks themselves will be located in closer proximity to the SFP walls. The licensee performed extensive calculations to predict dose rates in the areas below and adjacent to the sides of the SFP as a result of the increased fuel storage in the new SFP rack modules. The resulting calculations showed that the dose levels in most areas adjacent to the fuel storage areas will remain within acceptable levels.- However, the licensee will implement administrative controls (designating certain storage cell locations where freshly discharged Icel cannot be stored) and bridge-movement interlocks to ensure that area dose rates during fuel storage and movement do not exceed the maximum dose rate levels for which the areas are zoned. The storage of spent fuelin the Cask Storage Pit will necessitate the rezoning of the i

lower e8 vations of the Cask Storage Pit from Zone IV (less than 100 mrem /hr) to Zone V l

9

- (greater than or equal to 100 mrem /hr). The area at the top of the Cask Storage Pit will be rezoned from Zone 11 (less than 2.5 mrem /hr) to Zone 11-111 (less than 15 mrem /hr). These changes will make the zoning for the Cask Storage Pit at the various elevations the same as the current zoning for the SFP and Refueling Canal. The staff finds these changes to be acceptable.

i

. The calculated dose rate at the surface of the SFP with the increased fuel storage is estimated to be 0.25 mrom/hr. Dose rates on the fuel pool level are primarily due to radionuclides in the pool water. During the SFP reracking operation, the licensee expects dose rates in this a.ea to be between 2.5 and 5.0 mrom/hr, except during pool vacuuming operations, when the area dose rates are estimated to be between 4.0 and 8.0 mrem /hr. The staff finds then dose rates to be acceptable and in accordance with SFP dose rates at other plants.

On the basis of our review of the Waterford 3 proposal, the staff concludes that the Waterford 3 SFP rack modification can be performed in a manner that will ensure that doses to the workers will be maintained ALARA. The staff finds the projected dose for the project of 6 to 12 person-rom to be in the range of doses for similar SFP modifications at other plants and therefore acceptable.

2.6.2 Solid Radioactive Weste Spent resins are generated by the processing of SFP water through the SFP purification system.

In order to maintain the SFP water as clean as possible, and thereby minimize the generation of spent resins, the licensee will vacuum the floor of the SFP to remove any radioactive crud and other debris before the new fuel rack modules are installed. The licensee does not expect that the additional fuel storage made possible by the increased storage capacity will result in a significant change in the generation of solid radwaste.

Prior to removal from the SFP, each fuel rack module will be hydrolased with demineralized water and then scrubbed with a wire brush or equivalent abrasive tool, if necessary, to remove any loose crud and hot particles. Once removed from the SFP, the fuel rack modules, as well as portions of sparger pipe and other piping removed during the SFP reracking operation will be washed with demineralized water and then put into anti-contamination bags. These bags will then be put into special DOT approved shipping containers for shipment offsite to a facility l

licensed for the processing of low-level radioactive waste.

1 2.6.3 Desian Basis Accidents in its application, the licensee evaluated the possible consequences of a fuel handling accident to determine the thyroid ar'd whole-body doses at the Exclusion Area Boundary (EAB), Low-Population Zone (LPZ), and Control Room. The proposed reracking of the Waterford SFP will not affect any of the assumptions or inputs used in evaluating the dose consequences of the fuel handling accident.

The staff reviewed the licensee's analysis and performed confirmatory calculations to check the acceptability of the licensee's doses. In performing these calculations, the staff used the assumptions of RG 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. For the fuel handling accident, the staff assumed that the cladding of all of the fuel rods (236 rods) in a single fuel assembly would be perforated if the fuel assembly were dropped during handling. The damaged fuel assembly is assumed to contain freshly off-loaded fuel with a minimum of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> of decay. The parameters which the staff utilized in its assessment are presented in Table 1.

The staff's calculations confirmed that the thyroid doses at the EAB, LPZ, and Control Room from a fuel handling accident meet the acceptance criteria and that the licensee's calculations are acceptable. The results of the staff's calculations are presented in Table 2. For a fuel handling accident, the staff calculated a dose of 15.1 rem thyroid at the EAB and 0.20 rem

l l thyroid at the LPZ. The acceptance criterion at the EAB and LPZ for these accidents is l

contained in SRP Section 15.7.4 of NUREG-0800 and is 75 rem thyroid dose (25 percent of 10 CFR Part 100 guidelines of 300 rem). For a fuel handling accident, the staff calculated a dose of 0.19 rem thyroid to the control room operator. The acceptance criterion for the control room operator is 30 rem thyroid (SRP Section 6.4 of NUREG-0800). The staff, therefore, finds the 3

l proposed reracking at Waterford to be acceptable with respect to potential radiological consequences as a result of a hypothetical fuel handling accident.

3.0 STATE CONSULTATION

in accordance with the Commission's regulations, the Louisiana State official was notified of the l

proposed issuance of the amendment. The State official had no comments.

4.0 ENVIRONMENTAL CONSIDERATION

Pursuant to 10 CFR 51.21,51.32, and 51.35, an environmental assessment and finding of no significant impact was published in the Federal Register on July 7,1998 (63 FR 36722). In this finding, the Commission determined that issuance of this amendment would not have a significant effect on the quality of the human erivironment.

5.0 CONCLUSION

The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

Principal Contributors: L. Kopp D.Shum B. Thomas C. Hinson i

Y. Kim K. Parczewski Date: July 10,1998 i

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I i TABLE 1 ASSUMPTIONS USED FOR CALCULATING RADIOLOGICAL CONSEQUENCES OF A FUEL HANDLING ACCIDENT Parameters Power Level, Mwt 3844.3 Number of Fuel Rods Damaged (Single Assembly) 236 Total Number of Rods in Core 51,212 Shutdown Time, hours 72 Power Peaking Factor 1.8 Fission-Product Release Fractions (%)*

lodine (corrected for extended bumup) 12 Noble Gases 30 Pool Decontamination Factors *

- lodine -

100 Noble Gases i

lodine Forms (%)*

Elemental 75 Organic _

25 Filter Efficiencies for Control Room (%)

Elemental 99 Organic -

99 i

Control Room Flow Rates (ft*/ min)

Recirculation flow 3800 Pressurization flow 200 Unfiltered inleakage Atmospheric Dispersion Factors, X/Q (sec/m ).

13 8

d Exclusion Area Boundary (0-2 hours) 5.1 x 10 4

Low Population Zone (0-8 hours) 6.9 x 10 8

Control Room (0-8 hours) 1.66 x 10 Core Fission Product inventories per TID-14844

• Regulatory Guide 1.25 L____---__.____--_--_____

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. TABLE 2 THYROID DOSES FROM FUEL HANDLING ACCIDENT AT WATERFORD NALUES CALCULATED BY NRC STAFF)

DOSE (REM)

FUEL HANDLING ACCIDENT EAB*

0.19 LPZ*

0.20 Control Room **

2.5

• Acceptance Criterion = 75 rem thyroid

• • Acceptance Criterion = 30 rem thyroid l

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