ML20207L236

Safety Evaluation Supporting Amends 104 & 91 to Licenses NPF-76 & NPF-80,respectively

ML20207L236
Person / Time
Site:	South Texas
Issue date:	03/03/1999
From:	NRC (Affiliation Not Assigned)
To:
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ML20207L223	List: ML20207L220 ML20207L236
References
NUDOCS 9903180046
Download: ML20207L236 (11)

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UNITED STATES y

NUCLEAR REGULATORY COMMISSION i

WASHINeToN, D.C. 30008 80M i

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION i

RELATED TO AMENDMENT NOS.104 AND 91 TO FACILITY OPERATING LICENSE NOS. NPF-76 AND NPF-80 STP NUCLEAR OPERATING COMPANY i

DOCKET NOS 50-498 AND 50-499 l

SOUTH TEXAS PROJECT. UNITS 1 AND 2 I

1.0 INTRODUCTION

By application dated July 7,1998, as supplemented by letters dated October 15 and October 26,1998, STP Nuclear Operating Company, et al., (STPNOC, the licensee) requested changes to the Technical Specifications (TSs) (Appendix A to Facility Operating License Nos.

NPF-76 and NPF-80) for the South Texas Project, Units 1 and 2 (STP). The proposed changes would revise the spent fuel pool criticality analysis and rack utilization schemes by allowing credit for spent fuel pool soluble boron.

The October 15 and October 26,1998, and February 16,1999, supplements provide clarifying information and corrected administrative errors, and did not change the initial no significant hazards consideration determination.

2.0 CRITICALITY ANALYSIS 2.1 Discuss 6on in a letter of July 7,1998 (Ref.1), supplemented by letter of October 15,1998 (Ref. 2),

STPNOC requested changes to the STP TSs to allow the use of credit for soluble boron in the 4

spent fuel pool criticality analyses. These criticality analyses were performed using the methodology developed by the Westinghouse Owners Group (WOG) and described in WCAP-14416-NP-A, " Westinghouse Spent Fuel Rack Criticality Analysis Methodology,"

(Ref. 3).

The staff's evaluation of the criticality aspects of the proposed TS changes follows.

2.2 Evaluation The STP spent fuel storege racks were analyzed using the Westinghouse methodology, which has been reviewed and approved by the Nuclear Regulatory Commission (NRC) (Ref. 3). This methodology takes partial credit for soluble boron in the fuel storage pool criticality analyses and requires conformance with the following NRC acceptance criteria for preventing criticality outside the reactor:

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1) k,, shall be less than 1.0 if fully flooded with unborated water, which includes an allowance for uncertainties at a 95% probability, 95%

t confidence (95/95) level as described in WCAP-14416-NP-A; and 4

2) k., shall be less than or equal to 0.95 if fully flooded with borated water, which includes an allowance for uncertainties at a 95/95 level as described in WCAP-14416-NP-A.

l The analysis of the reactivity effects of fuel storage in the STP spent fuel racks was performed j

with the three-dimensional Monte Carlo code, KENO-Va, with neutron cross sections generated with the NITAWL-Il and XSDRNPM-S codes using the 227 group ENDF/B-V cross-section data.

Since the KENO-Va code package does not have bumup capability, depletion analyses and the j

determination of small reactivity increments due to manufacturing tolerances were made with the two-dimensional transport theory code, PHOENIX-P, which uses a 42 energy group nuclear i

data library from ENDF/B-V data. The analytical methods and models used in the reactivity analysis have been benchmarked against experimental data for fuel assernblies similar to those I

for which the STP racks are designed and have been found to adequately reproduce the critical values. This experimental data is sufficiently diverse to establish that the method bias and uncertainty will apply to rack conditions which include close proximity storage and strong neutron absorbers. The staff concludes that the analysis methods used are acceptable and capable of predicting the reactivity of the STP storage racks with a high degree of confidence.

The STP spent fuel pool contains two types of storage racks. The Region 1 racks were designed to use Boraflex panels in a removable stainless steel box to absorb neutrons. The Region 2 racks were fabricated by trapping Boraflex panels between the cell walls. The STP spent fuel storage racks have previously been qualified for storage of various Westinghouse 17x17XL fuel assembly types with maximum enrichments up to 5.0 weight percent (w/o) U-235.

The maximum enrichment is based on a nominal value of 4.95 w/o U-235 plus a manufacturing tolerance of 0.05. The spent fuel rack Boraflex absorber panels were considered in this previous analysis. Because of concems with the Boraflex deterioration that has been observed in many spent fuel pools, the STP spent fuel storage racks in Regions 1 and 2 have been reanalyzed neglecting the presence of Boraflex to allow storage of all 17x17XL fuel assemblies with nominal enrichments up to 4.95 w/o U-235 using credit for checkerboarding, bumup, bumable absorbers, and soluble boron. Also, because of concems with the spent fuel pool silica levels resulting from Boraflex degradation, STPNOC has stated that they may decide to physically remove the Boraflex, and the stainless steel water box inserts upon which the Boraflex panels are mounted, from the Region 1 racks. Therefore, the criticality analysis has been performed for the Region 1 racks both with and without the steel water box insert. Since removal of the water box insert would decrease the amount of neutron capture in this area and therefore increase the reactivity, the results from these cases bound the results from the cases with the water boxes included and were used in the TS revisions, thereby allowing either configuration.

The moderator was assumed to be pure water at a temperature of 68'F and a density of 1.0 gm/cc and the array was assumed to be infinite in lateral (x and y) extent. Uncertainties due to tolerances in fuel enrichment and density, storage cell inn 3r diameter, storage cell pitch, stainless steel thickness, assembly position, calculational uncertainty, and methodology bias uncertainty were accounted for. These uncertainties were appropriately determined at the 95/95 probability / confidence level. A methodology bias (determined from benchmark calculations) as well as a reactivity bias to account for the effect of the normal range of spent fuel pool water temperatures (50*F to 160*F) were included. These biases and uncertainties meet the previously stated NRC requirements and are, therefore, acceptable.

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For Region 1, the nominal enrichment required to maintain k,less than 1.0 with all cells filled with Westinghouse 17x17XL fuel assemblies and no soluble boron in the pool water was found

- to be 2.50 w/o U-235 (Category 4 fuel as defined by TS 5.6.1.2). This resulted in a nominal k, of 0.97070. The 95/95 L was then determined by adding the temperature and methodology biases and the statistical sum of independent tolerances and uncertainties to the nominal k, values, as described in Reference 2. This resulted in a 95/95 k, of 0.99660. Since this value is less than 1.0 and was determined at a 95/95 probability / confidence level, it meets the NRC criterion for precluding criticality with no credit for soluble boron and is acceptable.

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Soluble boron credit is used to provide safety margin by maintaining k,less than or equal to 0.95 including 95/95 uncertainties. The soluble boron credit calculations assumed the all cell storage configuration moderated by water borated to 200 ppm. As previously described, the individual tolerances and uncertainties, and the temperature and methodology biases, were added to the calculated nominal 4 to obtain a 95/95 value. The resulting 95/95 k,was 0.94579 for fuel enriched to 2.50 w/o U-235. Since k,is less than 0.95 with 200 ppm of boron and uncertainties at a 95/95 probability / confidence level, the NRC acceptance criterion for precluding criticality is satisfied. The required amount of soluble boron is well below the minimum spent fuel pool boron concentration value of 2500 ppm required by TS 3.9.13 and is, therefore, acceptable.

The concept of reactivity equivalencing due to fuel bumup was used to achieve the storage of fuel assemblies with enrichments higher than 2.50 w/o U-235 for the all cell storage configuration. The NRC has previously accepted the use of reactivity equivalencing predicated upon the reactivity decrease associated with fuel depletion. To determine the amount of soluble boron required to maintain k,s0.95 for storage of fuel assemblies with maximum enrichments up to 5.0 w/o U-235, a series of reactivity calculations were performed to generate a set of enrichment versus fuel assembly discharge burnup ordered pairs which all yield an equivalent k, when stored in the STP spent fuel storage racks. These are shown in TS Figure 5.6-4 and represent combinations of fuel enrichment and discharge burnup which yield the same rack k, as the rack loaded with 2.50 w/o fuel (at zero bumup). Uncertainties associated with burnup credit include a reactivity uncertainty of 0.01 Ak at 30,000 MWD /MTU applied linearly to the bumup credit requirement to account for calculational and depletion uncertainties and 5% on the calculated bumup to account for bumup measurement uncertainty. The NRC staff concludes that these uncertainties conservatively reflect the uncertainties associated with burnup calculations and are acceptable. The amount of additional soluble boron, above the 200 ppm value required above, that is needed to account for these uncertainties is 300 ppm.

This results in a total soluble boron credit of 500 ppm for the all cell configuration. This is well below the minimum spent fuel pool boron concentration value of 2500 ppm required by TS j

3.9.13 and is, therefore, acceptable.

Storage of assemblies with enrichments higher than 2.50 w/o U-235 in the all cell storage configuration was also determined by crediting the reactivity decrease associated with the addition of integral fuel bumable absorbers (IFBAs). IFBAs consist of neutron absorbing material applied as a thin ZrB, coating on the outside of the UO pellet. As with bumup credit, 2

for IFBA credit reactivity equivalencing, a series of reactivity calculations are performed to generate a set of IFBA rod number versus initial enrichment ordered pairs which all yield the equivalent k, when the fuel is stored in the all cell configuration analyzed for the STP spent fuel racks as shown in TS Figure 5.6-5. Uncertainties associated with IFBA credit include a 5%

manufacturing tolerance and a 10% calculational uncertainty on the B-10 loading of the IFBA rods. The staff finds these uncertainties adequately conservative and acceptable. The amount j

of additional soluble boron needed to account for these uncertainties is the same as the 300 ppm required for bumup credit.

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t Therefore, with the above reactivity equivalencing, fuel assemblies with maximum enrichments up to 5.0 w/o U-235 can be stored in all Region 1 cell locations by taking credit for a total additional amount of soluble boron of 300 ppm. When added to the 200 ppm required without reactivity equivalencing, this results in a total boron requirement of 500 ppm. This is well below the minimum spent fuel pool boron concentration value of 2500 ppm required by TS 3.9.13 and is, therefore, acceptable.

In a similar fashion, criticality analyses were performed for a two separate Region 1 checkerboard storage configurations. The first configuration (Checkerboard #1) consisted of 17x17XL fuel assemblies in a 2x2 checkerboard arrangement containing fuel assemblies in two diagonally adjacent cells with an initial, nominal enrichment no greater than 1.70 w/o U-235 (Category 6) and fuel assemblies in the two remaining cells with a nominal enrichment no greater than 3.55 w/o U-235 (Category 3). Fuel assemblies with enrichments greater than these values may be stored in the Region 1 Checkerboard #1 arrangement if they satisfy the minimum bumup requirements given in TS Figures 5.6-7 and 5.6-2 for Category 6 and Category 3 fuel, respectively, or if the Category 3 fuel contains a minimum number of IFBAs as given in TS Figure 5.6-3. The soluble boron concentration that results in k,,s0.95 was calculated to be 300 ppm.

The second Region 1 checkerboard configuration (Checkerboard #2) consisted of 17x17XL fuel assemblies in a 2x2 checkerboard arrangement containing assemblies in two diagonally adjacent cells with an initial, nominal enrichment no greater than 1.40 w/o U-235 (Category 10) and 1.70 w/o U-235 (Category 6) respectively, an assembly in one remaining cell with an initial, nominal enrichment no greater than 2.50 w/o U-235 (Category 4) and an assembly in the remaining cell with a nominal enrichment no greater than 4.95 w/o U-235 (Category 1). Fuel assemblies with enrichments greater than these values may be stored in the Region 1 Checkerboard #2 configuration if they satisfy the minimum bumup requirements given in TS Figures 5.6-11, 5.6-7, and 5.6 4 for Category 10, 6, or 4 fuel, respectively, or if the Category 4 fuel contains the minimum number of IFBAs given in TS Figure 5.6-5. The soluble boron concentration that results in k,,s0.95 was calculated to be 400 ppm.

The criticality analysis for Region 2 fuel storage showed that 17x17XL assemblies with initial nominal enrichments no greater than 1.20 w/o U 235 (Category 11) can be stored in all cell locations. Fuel assemblies with enrichments greater than this may be stored in all cells if they satisfy the minimum bumup requirements of TS Figure 5.6-12. The Figure also credits the time

- an assembly has been discharged from the core. Decay time credit is an extension of the bumup credit process and results from the radioactive decay of isotopes in the spent fuel to daughter isotopes, which results in reduced reactivity. Although decay of the fission products has the effect of further reducing the reactivity of the spent fuel, in this amendment request, credit is taken only for the decay of actinides. Decay time credit has been previously approved by the NRC (Ref. 4). Calculations were also performed to assess the impact of loading a slightly higher enriched assembly in the peripheral cells (next to the pool wall or separated from Region 1 fuel by an empty row). For fuel assemblies on the periphery of the Region 2 racks, storage of 17x17XL assemblies with an initial nominal enrichment no greater than 1.40 w/o U-

. 235 (Category 2) or which meet the bumup requirements of TS Figure 5.6-9 was found to be acceptable. The soluble boron concentration that results in k,,s0.95 was calculated to be 700 ppm.

A 3-out-of-4 checkerboard storage arrangement with an empty cell was analyzed for the Region 2 racks. Three assemblies with initial nominal enrichment no greater than 1.70 w/o

l U-235 (Category 5) can occupy any 2x2 matrix with the fourth cell vacant. Higher enriched l

assemblies must meet the burnup requirements given in TS Figure 5.6-6. The soluble boron concentration that results in k,,s0.95 was calculated to be 550 ppm.

A 2-out-of-4 checkerboard storage arrangement with empty cells was analyzed for the Region 2 racks. Two assemblies with initial nominal enrichment no greater than 4.85 w/o U-235 (Category 2) can be stored corner adjacent in a 2x2 matrix with the other two cells vacant. The soluble boron concentration that results in k,,s0.95 was calculated to be 300 ppm.

Two additional storage configurations were analyzed for Region 2 taking credit for silver-indium-cadmium (Ag-in-Cd) or hafnium (Hf) rod cluster control assembly (RCCA). STP Units 1 and 2 have used both Ag-In-Cd and Hf RCCA absorber material. Since the Hf RCCAs provide slightly less reactivity holddown than the Ag-In-Cd RCCAs, the Hf RCCAs were used in the criticality analysis for conservatism. In addition, the staff concludes that a conservative allowance for the i

reactivity worth of the RCCA absorber material was assumed by depleting the full length of the RCCA for 60,000 MWD /MTU exposure. Credit for added absorber (rods, plates, or other con /igurations) has been allowed by the NRC provided it can be clearly demonstrated that design features prevent such absorbers from being removed, either inadvertently or intentionally, without unusual effort such as the necessity for special equipment maintained l

under positive administrative control. In response to a staff request for additionalinformation, l

STPNOC stated that special equipment, i.e., the RCCA Change Tool, is required in order to l

move RCCAs in the spent fuel pool (Ref. 2). Operation of this tool requires a unique electrical power cord and an instrument air line with regulator. Authorization to use this equipment is controlled by the Core Loading Supervisor during refueling outages, or by the Shift Supervisor l

at all other times. Based on this, the staff concludes that special equipment, which is maintained under positive administrative control, is necessary in order to remove any inserted i

RCCAs. Therefore, the use of RCCAs for reactivity holddown in these fuel assemblies is acceptable.

Storage of 17x17XL fuel assemblies with an initial nominal enrichment of no greater than 1.40 w/o U-235 (Category 9) was analyzed for a 2x2 checkerboard where one of the four assemblies contains a Ag-In-Cd or Hf RCCA (RCCA #1 Checkerboard). Higher enriched assemblies must I

meet the bumup requirements of TS Figure 5.6-10. The soluble boron concentration that results in k,,s0.95 was calculated to be 650 ppm.

The final Region 2 analyzed storage configuration was for 17x17XL assemblies with initial nominal enrichment no greater than 1.65 w/o U-235 (Category 7) in a 2x2 checkerboard where two of the diagonally adjacent assemblies contain a Ag-in-Cd or Hf RCCA (RCCA #2 Checkerboard). Higher enriched assemblies must satisfy the bumup requirements of TS Figure 5.6-8. The soluble boron concentration that results in k,,s0.95 was calculated to be 700 ppm.

Based on the above analyzed storags configurations, the maximum required total soluble boron (700 ppm) occurs for both the Region 2 all cell storage and the Region 2 RCCA #2 Checkerboard pattems.

Although most accidents will not result in a reactivity increase, three accidents can be postulated for each storage configuration which would increase reactivity beyond the analyzed i

conditions. The first would be a loss of fuel pool cooling system and a rise in pool water temperature from 160*F to 240'F. The second would be dropping an assembly into an already loaded cell. The third accident would be a mistoad of an assembly into a cell for which the restrictions on location, enrichment, or burnup are not satisfied.

O Calculations have shown that the misload assembly accident for a 2-out-of-4 checkerboard i

configuration results in the highest reactivity increase. The reactivity increase requires an l

additional 1800 ppm of soluble boron to maintain k,,,50.95. However, for such events, the double contingency principle can be applied. This states that the assumption of two unlikely, independent, concurrent events is not reqaired to ensure protection against a criticality L

accident. Therefore, the minimum amount of boron required by TS 3.9.13 (2500 ppm) is more than sufficient to cover any accident and the presence of the additional boron above the i

concentration required for normal conditions and reactivity equivalencing (700 ppm maximum) can be assumed as a realistic initial condition since not assuming its presence would be a second unlikely event.

In order to prevent an undesirable increase in reactivity, the boundaries between the different storage configurations were analyzed. The boundary between checkerboard zones and the boundary between a checkerboard zone and an all cell storage region must be controlled to i

prevent an undesirable increase in reactivity. The fuel storage patterns used within Region 1 must comply with the interface requirements shown in TS Figures 5.6-15 and 5.6-16. Fuel storage patterns used within Region 2 must comply with the interface requirements shown in TS Figures 5.6-17 through 5.6-1g. One row of empty water cells must be maintained at the interface between Region 1 and Region 2, and can be positioned in either Region. Non-fissile items can be stored in these empty water cells per the provisions of TS 5.6.1.3.

l The TS changes proposed as a result of the revised criticality analysis are consistent with the NRC-approved methodology given in Westinghouse topical report, WCAP-14416-NP-A, Rev.1, (Ref. 3). Based on this consistency with the approved methodology and on the above i

evaluation, the staff finds these TS changes acceptable. The proposed associated Bases changes adequately describe these TS changes and are also acceptable.

2.3 Summarv l

Based on the review described above, the staff finds the criticality aspects of the proposed STP l

license amendment request are acceptable and meet the requirements of General Design l

Criterion 62 for the prevention of criticality in fuel storage and handling. The analysis assumed l

credit for soluble boron, as allowed by WCAP-14416-NP-A, but no credit for the Boraflex l

neutron absorber panels. The required amount of soluble boron for each analyzed storage l

configuration is shown in attached Table 1. The criticality analysis conformed to the NRC guidance on the regulatory requirements for criticality analysis of fuel storage at light-water reactor power plants (Ref. 5).

The following storage configurations and U 235 e'nrichment limits for Westinghouse 17x17XL l

fuel assemblies were determined to be acceptable:

l Reaion 1 l

All Cell Storace Ass 6mblies with initial nominal enrichments no greater than 2.50 w/o U-235 can be stored in any cell location. Fuel assemblies with initial nominal enrichments greater than this and up to 4.95 w/o U-235 must satisfy the minimum bumup requirements shown in TS Fig. 5.6-4 or contain a minimum number of IFBAs as shown in TS Fig. 5.6-5.

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Checkerboard #1 Storaos Assemblies can be stored in a 2x2 checkerboard arrangement consisting of two diagonally ad,acent fuel assemblies with initial nominal enrichment no greater than 1.70 w/o U 235 and fuel assemblies in the two remaining cells with initial nominal enrichment no greater than 3.55 w/o U-235. Fuel assemblies with initial enrichments greater than this and up to 4.95 w/o U-255 must satisfy the minimum bumup requirement shown in TS Fig. 5.6-7 (1.70 w/o assemblies) or Fig. 5.6-2 (3.55 w/o assemblies) or must satisfy a minimum IFBA requirement as shown in TS Fig. 5.6-3 (3.55 w/o assemblies).

Checkerboard #2 Storaos 1

Assemblies can be stored in a 2x2 checkerboard arrangement consisting of two diagonally adjacent fuel assemblies with initial nominal enrichment no greater than 1.40 w/o and 1.70 w/o U-235, an assembly in one remaining cell with a l

nominal enrichment no greater than 2.50 w/o U-235, and a fuel assembly in the j

remaining cell with a nominal enrichment no greater than 4.9E w/o U-235. Fuel assemblies with initial nominal enrichments greater than these and up to 4.95 w/o U-235 must satisfy the minimum burnup requirement shown in TS Figs. 5.6-11 (1.40 w/o assemblies), Fig. 5.6 7 (1.70 w/o assemblies), or i

Fig. 5.6-4 (2.50 w/o assemblies) or must satisfy a minimum IFBA requirement as shown in TS Fig. 5.6-5 (2.50 w/o assemblies).

Realon 2 i

All Cell Storace Assemblies with initial nominal enrichments no greater than 1.20 w/o U-235 can be stored in any cell location. Fuel assemblies with initial nominal enrichments greater than this and up to 4.95 w/o U-235 must satisfy the minimum bumup requirements shown in TS Fig. 5.6-12.

4 Perioherv Location Storaae Assemblies with initial norhinal enrichments no greater than 1.40 w/o U-235 can be stored on the periphery of the Region 2 rack modules. Fuel assemblies with initial nominal enrichments greater than this and up to 4.95 w/o U-235 must satisfy the minimum burnup requirements shown in TS Fig. 5.6-9.

3-out-of-4 Checkerboard Storaae Assemblies with initial nominal enrichments no greater than 1.70 w/o U 235 can be stored in a 3-out-of-4 checkerboard arrangement with empty cells. This means that no more than three fuel assemblies can occupy any 2x2 matrix of cells. Fuel assemblies with initial nominal enrichments greater than this and up to 4.95 w/o U 235 must satisfy the minimum burnup requirements shown in TS Fig. 5.6-6.

4 2-out-of-4 Checkerboard Storace Assemblies with initial nominal enrichments nn greater than 4.85 w/o U-235 can be stored in a 2-out-of-4 checkerboard arran ement with empty cells. This 1

means that no two assemblies may be storet..me adjacent but must be stored comer adjacent. Fuel assemblies with initial nominal enrichments greater than j

this and up to 4.95 w/o U-235 must satisfy the minimum burnup requirements 1

shown in TS Fig. 5.6-1.

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RCCA #1 Checkerboard Storaos l

Assemblies with initial nominal enrichments no greater than 1.40 w/o U-235 can l

be stored in a 2x2 checkerboard arrangement where one of the four assemblies contains a Ag-in-Cd or Hf RCCA. Fuel assemblies with initial enrichments l

greater than this and up to 4.95 w/o U 235 must satisfy the minimum bumup requirements shown in TS Fig. 5.6-10.

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RCCA #2 Checkerboard Storage l

Assemblies with initial nominal enrichments no greater than 1.65 w/o U-235 can be stored in a 2x2 checkerboard arrangement where two diagonally adjacent of the four assemblies contain a Ag-In-Cd or Hf RCCA. Fuel assemblies with initial l

enrichments greater than this and up to 4.95 w/o U-235 must satisfy the minimum bumup rsquirements shown in TS Fig. 5.6-8.

l 3.0 BORON DlLUTION ANALYSIS 3.1 Discussion in accordance with the NRC Safety Evaluation (Ref. 6) of the Westinghouse methodology i

described in WCAP-14416-A, the licensee performed a boron dilution analysh to ensure that l

sufficient time is available to detect and mitigate the dilution prior to exceeding the 0.95 k,

design basis. The licenae provided a boron dilution analysis on July 7,1998 (Ref.1), and i

supplemental information on October 26,1998 (Ref. 7). Potential events were quantifed to show that sufficient time is available to enable adequate detection and suppression of any dilution event.

3.2 Evaluation Deterministic dilution event calculations were performed for STP to define the dilution times and volumes necessary to dilute the spent fuel pool from the minimum TS boron concentration of 2500 ppm to a soluble boron concentration of 700 ppm. Because the Unit 1 and Unit 2 spent fuel pools are essentially identical, the analysis applies to both pools. Each spent fuel pool has a water inventory of 420,000 gallons. Assuming a well-mixed pool, the volume required to dilute the spent fuel pool from the TS limit of 2500 ppm to 700 ppm is 534,600 gallons. The various events that were considered included dilution from the boron recovery system, reactor makeup system, domineralized water system, fire protection system, and other events that may affect the boron concentration of the pool, such as seismic events, pipe break, and loss of offsite power.

There are three water storage sources that could provide the 534,600 gallons of water needed to dilute the spent fuel pool boron concentration to 700 ppm. The reactor makeup water tank has a volume of 153,050 gallons with an automatic makeup source, which would be sufficient with the makeup to dilute the spent fuel pool boron concentration to 700 ppm. The reactor j

makeup system is directly connected to the spent fuel pool and is isolated by one closed manual valve. However, the largest dilution rate would be 240 gpm, which would take over 37 hours4.282407e-4 days <br />0.0103 hours <br />6.117725e-5 weeks <br />1.40785e-5 months <br /> to dilute the spent fuel pool to 700 ppm. The licensee's domineralized water tank (961,000 gallons) also is sufficiernt to dilute the spent fuel pool boron concentration to 700 ppm.

However, the most rapid dilution would occur through the 2-inch makeup line. The dilution event would require over 46 hours5.324074e-4 days <br />0.0128 hours <br />7.60582e-5 weeks <br />1.7503e-5 months <br /> at 190 gpm to dilute the pool to 700 ppm. These events would be identified through the h3gh level pool alarm or by operator rounds, which are conducted every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

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e The licensee identified that a random pipe break in the fire protection standpipe and sprinkler manifold piping provi@.1 the possible largest flow rate of the dilution sources. Additionally, this pipe is evaluated because the water in the fire protection tanks (600,000 gallons) is sufficient to dilute the spent fuel pool boron concentration to 700 ppm withot.t replenishment. The licensee identified that a break in this 6-inch piping could have a flow rate of 4,000 gpm. If all the water was deposited directly into the spent fuel pool, a break in this fire protection standpipe would take approximately five minutes to fill the pool to the high level alarm actuation level. However, the 6-inch pipe is located below the spent fuel pool deck elevation and branches to smaller two-and three-inch lines prior to penetrating the deck elevation. Therefore, the licensee concluded that a break in the 6-inch line has no effect on the spent fuel pool inventory or boron concentration. The licensee evaluated the smaller lines and determined that it would take at least 35 hours4.050926e-4 days <br />0.00972 hours <br />5.787037e-5 weeks <br />1.33175e-5 months <br /> to dilute the pool boron concentration to 700 ppm. Since a dilution due to a break in one of the smaller lines would take longer than 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, it would be identified and terminated by plant personnel during rounds, if not earlier, price to reaching a boron concentration of 700 ppm in the spent feel pool.

Other evaluated dilution events take longer than twelve hours to reah the minimum boron concentration. These events would be detected by plant personnel during required rounds every twelve hours. To detect low flow, long term dilution events, the licensee samples its spent fuel pool every seven days. This frequency is consistent with the standard TSs for Westinghouse plants and is considered appropriste for this plant.

The licensee concluded an unplanned or inadvertent event that would dilute the spent fuel pool boron concentration from 2,500 ppm to 700 ppm is not credible for STP. The staff finds that the combination of the large volume of water required for a dilution event, TS-controlled spent fuel pool concentration and 7-day sampling requirement, and plant personnel rounds would adequately detect a dilution event prior to k, reaching 0.95 (700 ppm); therefore, the analysis j

and proposed TS controls are acceptable for the boron dilution aspects of the request.

Additionally, the criticality analysis for the spent fuel storage pool show that k, remains less than 1.0 at a 95/95 probability / confidence level even if the pool were completely filled with unborated water. Therefore, even if the spent fuel storage pool were diluted to zero ppm, the racks are expected to remain suberitical.

3.3 Summary Based on the review described above, the staff finds the boron dilution aspects of the proposed license amendment request acceptable. The TS boron concentration of 2500 ppm and 7-day surveillance requirements are acceptable to ensure that sufficient time is available to detect and mitigate a dilution event prior to exceeding the design basis k, of 0.95.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the Texas State official was notified of the proposed issuance of the amendments. The State official had no comments.

5.0 ENVIRONMENTAL CONSIDERATION

The amendments change a requirement with respect to installation or use of a facility compcnent located within the restneted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendments involve no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is l

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[ no significant increase in individual or cumulative occupational radiation exposure. The i

Commission has previously issued a proposed finding that the amendments involve no l

significant hazards consideration, and there has been no public comment on such finding (63 FR 45530). Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connectk.n with the issuance of the amendments.

6.0 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 amendr'ents will not be inimical to the common defense and security or to the health and safety of the puolic.

7.0 REFERENCES

1)

T. H. Cloninger, STPNOC, letter to U. S. Nuclear Regulatory Commission, " South Texas Project, Units 1 and 2, Docket Nos. STN 50-498, STN 50-499, Proposed Amendment to Technical Specifications for Spent Fuel Storage Pool Rcck, Criticality Analysis with Soluble Boron Credit," July 7,1998.

2)

D. A. Leazar, STPNOC, letter to U. S. Nuclear Regulatory Commission, " South Texas Project, Units 1 and 2, Docket Nos. STN 50-498, STN 50-499, Response to Staff Questions on Proposed Amendment to Technical Specifications for Spent Fuel Storage Pool Rack Criticality Analysis with Soluble Boron Credit," October 15,1998.

3)

W. D. Newmyer, " Westinghouse Spent Fuel Rock s - iticality Analysis Methodology,"

Westinghouse Electric Corporation, WCAP-14416-NP-A, Rev.1, November 1996.

4)

Prairie Island Nuclear Generating Plant, Unit Nos.1 and 2, Issuance of Amendment Re:

Credit for Soluble Boron in Spent Fuel Pool Criticality Analysis," June 12,1997.

5)

L. Kopp, Guidance on the Regulatory Requirements for Criticality Analysis of Fuel Storage at Light-Water Reactor Power Plants, NRC Memorandum to T. Collins, August 19,1998.

6)

NRC letter to Mr. T. Greene, Westinghouse Owners Group, dated October 25,1996,

Enclosure:

NRC Safety Evaluation of WCAP-14416-P.

7)

T.J. Jordan, STPNOC, letter to U.S. Nuclear Regulatory Commission,

• Response to NRC Staff Questions on Fire Protection System with regard to the Proposed Amendment to the Technical Specifications for Spent Fuel Storage Pool Rack Criticality Analysis with Soluble Boron Credit," dated October 26,1998.

PrincipalContributors: L.Kopp D. Jackson Date: March 3, 1999

Attachment:

Table 1

p-i L

4

\\

g TABLE 1 o

Summarv of Soluble Boron Credit Reauirements for South Texas Units 1 and 2 Total Soluble Soluble Boron Boren Credit Soluble Boron Required for Required l

Storage Required for Reactivity Without Configuration k, s 0.95 Equivalencing Accidents (ppm)

(ppm)

(ppm) l Reaion 1 All Cells 200 300 500 l

Checkerboard

1. 1 200 100 300 Checkerboard

1. 2 250 150 400 Recion 2 All Cells 200 500 700 3-out-of-4 Checkerboard 200 350 550 2-out-of-4 Checkerboard 250 50 300 RCCA #1 Checkerboard 200 450 650 RCCA #2 Checkerboard 250 450 700 ATTACHMENT

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