ML20203H651

Safety Evaluation Supporting Amends 99 & 77 to Licenses NPF-68 & NPF-81,respectively

ML20203H651
Person / Time
Site:	Vogtle
Issue date:	02/20/1998
From:	NRC (Affiliation Not Assigned)
To:
Shared Package
ML20203H648	List: ML20203H644 ML20203H651
References
NUDOCS 9803030281
Download: ML20203H651 (12)

Text

.

pn ang fl A

UNITED STATES g

NUCLEAR REGULAYORY COMMISSION WASHINGTON, D.C. 8004H001 8eee*

SAFETY GVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT No. 99 TO FACILITY OPERATING LICENSE NPF48 AND AMENDMENT NO. 77 TO FACILITY OPERATING LICENSE NPF-81 SOUTHERN NUC! PAR OPEP.ATING COMPANY. INC.. ET AL.

VOGTLE ELECTRIC GENERATING PLANT. UNITS 1 AND 2 DOCKET NC 0 50-424 AND 50-425 1.0 INTROE JCTION By letter dated August 8,1997, as supplemented Octobcr 10,1997, January 16,23, and 27, 1998, Southem Nuclear Operating Company, Inc., et al. (the licensee, SNC) proposed license amendments to change the Technical Specifications (TS) for Vogtle Electric Generating Plant (VEGP), Units 1 and 2. The proposed changes v ould revise VEGP TS 3.7.17, ' Fuel Storage Pool Boron Concmtration,' TS 3.7.18, " Fuel Assembly Storage in the Fuel Storage Pool,' and TS 4.3, ' Fuel Storage,' to credit soluble boron,n the spent fuel pool for maintenance of subcriticality associated with spent fue' c,orage. The supplements dated January 16,23, and 27,1998, prov:ded clarifying information that did not change the scope of the August 8, i

1997, application and the initial proposod no significant hazards determination.

2.0 EVALUATION The VEGP spent fuel storage facility utilizes spent fuel racks that incorporate a fixed neutron polson, which is referred to as "Boraflex." Boraflex, an elastomer that conta:ns boron, is manufactured in sheet form and contained in the sides c 'he spent fuel racks, and is credited for reduction of the reactivity associated with spent fuel. The VEGP, Units 1 and 2, spent fue; j

poole normally contain borated water, which has not previously been credited in the reduction of reactivity associated with spent fuel. The spent fuel racks are described in Section 9.1.2,

" Spent Fuel Stcrage" of the VEGP Fl.1al Safety Analysis Report.

The August 8,1997, application and supplements profie analyses that would support the crediting of the borated water in the spent fuel pool 6 ; duction of reactivity associated with i

the spent fuel. The licensee has provided a criticality analysis to demonstrate that the borated water in the spent fuv. pools provides criticality control that meets NRC staff requirements fo spent fuel storage without crediting the poison effects of the Bormflex in the spent fuel storage racks. In addition, since the borated water in the spent fuel storage pools is subject to dilution, an analyris of the effects of boron dilution of the spent fuel pools' water was also provided by

. the licensee. Finally, the licensee has proposed changes to the VE3P TS that assure te assumptions of the spent fuel pool storage analyses remain valid.

980?O30281 980220 PDR ADOCK 05000424 P

PUR l

2 2.1 Snant Fuel Criticality in its letter of August 8,1997 (Ref.1), SNC requested changes to the VEGP, Units 1 and 2, TS to allow the use of credit for soluble boron in the 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 Rock Criticality Analysis Methodology"(Ref. 2).

3 The VEGP spent fuel ste, rage rocks were analyzed using the Westinghouse methodology, which has been reviewr1 and approved by the 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:

(1)

% shall be less than 1.0 if the pool is fully flooded with unborated water, which includee an allowance for uncertainties at a 95% probability, g5% conMence (95/95) level as

. described in WCAP-14416-NP A; and-(2) 4 shall be less than or equal to 0.95 If the pool is fully flooded with borated water, which includes an allowance for uncertaitses at a 95/95 level as described in WCAP 14416-NP-A.

= The analysis of the reactivity effects of fuel storage in the VEGP opot fuel rocks wa.s performed with the three-dimensional Monte Carlo code, KENO-Va, with neutron creas sections generated with the NITAWL-ll and XSDRNPM-S codes using the 227 group ENDr /B-V cross-section library.. Since the KENO Va code package does not have bumup capability, depletion analyses and the determination of small reactivity increments due to manufacturing tolerances were made with the two dimensional transport theory code, PHOENIX P, which uses e 42 energy group nuclear data library. _ The analytical methods and models used in the reactivity analysis have been benchmarked against experimental data for fuel assemblies similar to those for wnich the VEGP racks are deigned and have been found to adequately reproduce the

- critical values. This exoerimental data is sufficiently ulverse 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 VEGP storage racks with a high degree of confidence.

The VEGP spent fuel storage racks have previously been qualified for storage of various Westinghouse 17 x 17 fuel assembly types with maximum enrichments up to 4.55 weight

. percent (w/o) U 235. The maximum enrichment is based on a noriinal value of 4.50 w/o U-235 plus a manufacturing tolerance of 0.05. The spent fuel rock Boreflex absorber panels were considered in this previous analysis. Because of the Boreflex deterioration that has been observed in many spent fuel pools, the VEGP spent fuel storage racks have been reanalyzed neglecting the presence of Boraflex to allow storage of all 17 x 17._ fuel assemblies with nominal enrichments up to 5.0 w/o U 235 (enrichment tolerance of *0.05 w/o U 235) using credit for checkerboarding, bumup, bumable absorbers, and soluble boron.

3-The moderator was assumed to be pure water at a tempmture of 88 'F and a density of 1.0 gm/cc and the array was as:,umed to be iranite in lates al extent. Uncertainties due to tolerances in fuel enrichment and density, storage cell inner diameter, storage cell peh, stainless steel thickness, assembly position, calculational uncertainty, and methodology bias uncertainty were accounted for. These uncertainties were appropriately determined at the l

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 185 'F) were includedi These biases and uncertainties meet the previously stated NRC requirements and are, therefore, scooptable.

l For Unit 1, an enrichment of 1.79 w/o U 235 was found to be adequate to maintain 4 less than 1.0 with all cells filled with Westinghouse 17 x 17 fuel assemblies and no soluto boron in the

- pool water. This resulted in a nominal 4 of 0.94250. The 95/95 4 was then determined by adding the temperature ano methodology biases and the statistical sum of independent tolerances and uncertainties to the nominal 4 values, as desenbod in Reference 2. This resulted in a 95/95 4 of 0.39784. For Unit 2, a maximum initial nominal enrichment of 1,77 w/o U 235 resulted in a 95/95 4 of 0.99851. Since these values are less than 1.0 and were determined at a 95/95 probability / confidence level, they meet the NRC criterion for precluding criticality with no credit for soluble boron and are acceptable.

Soluble boron credit is used to provide safety margin by maintaining the effective multiplication factor,4, 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 (Unit 1) and 150 ppm (Unit 2). 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 4 was 0.93457 for fuel enriched to 1.79 w/o U 235 in Unit 1 and 0.94998 for fuel en;iched to 1.77 w/o U 235 in Unit 2. Since 4 is less than 0.95 with 200 ppm (Unit 1) and 150 ppm (Unit 2) of boron and uncertaintiea at a 95/95 probability / confidence level, the NRC acceptance criterion for precluding critica!ity is satisfied.

These values are well belev the minimum spent fuel pool boron concentration velue of 2000 ppm required by proposed TS 3.7.17 and are, therefors, neceptable.

The concept of reactMty equivalencing due to fuel bumup was used to achieve the storage of fuel assemblies with enrichments higher than 1,79 w/o U 235 (Unit 1) or 1.77 w/o U 235 (Unit 2) 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 4 s0.95 fo storage of fuel assemblies wl'h 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 bumup ordered pairs, which all yloid an equivalent 4 when stored in the VEGP spent bel storage racks, These are shown in proposed TS Figure 3.7.18-1 for Unit 1 and proposed TS Figure 3.7.18-2 for Unit 2 and represent combinations of fuel enrichment and discharge bumup, which yield the same rack 4 as the rack loaded with fresh 1.79 w/o fuel (Unit 1) or 1.77 w/o fuel (Unit 2).

Uncertainties associated with bumup credit include a reactivity uncertainty of 0.01 Ak at 30,000 MWD /MTU applied linearly to tha bumap credit requirement to account for calculational and depletion uncertainties and 5% on the calculated bumup to account for bumup measurement uncertainty.

L

/""

4 The NRC statt concludes that these uncertainties conservatively reflect the uncertainties associated with bumup calculations and are acceptable. The amount of additional soluble -

boron, above the value required above, that is needed to sooount for these unoortainties is 250 ppm in UnN 1 and 200 ppm in Unit 2. This results in a total soluble boron credM for the all cell conflgun.

' 450 ppm (Unit 1) and 350 ppm (Unit 2). These values are well below the minimum og bel pool boron concentration value of 2000 ppm required by proposed l

TB 3.7.17 and are, therefore, acceptable.

The VEGP spent fuel pool was also analyzed assuming a bout-of.4 checkerboard storage configuration containing three initially enriched 2.45 w/o U 235 assemblies (Unit 1) and three l

Indially enriched 2.40 w/o U 235 assemblies (Unit 2) and an empty cell. This resulted in a 95/95 l

4 of 0.99578 for Unit 1 and 0.99464 for Unit 2 with no credit for soluble boron or Boreflex.

These values meet the NRC criterion of 4 less than 1.0 with no credit for boron. The same I

configurations were then analyzed to obtain the required 5% subcritical margin assuming i

200 ppm of soluble boron. The resulting 95/95 4 was 0.93777 (Unit 1) and 0.93716 (Unit 2).

Since these 4 values a*e less than 0.95, including soluble boron credit and uncertainties at a 95/95 probability / confidence level, the NRC acceptance criterion is met for the 3-out of-4 cells storage configuration in Units 1 and 2c Burnup reactivity equivalencing, as previously described, was also used to determine the allowed storage of fuel assemblies with enrichments higher than 2.45 w/o (Unit 1) and 2.40 w/o (Unit 2) but no greate than 5.0 w/o U 235 in the 3-out-of 4 configuration. The amvunt of soluble boron needed to account for the additional uncertainties associated with bumup credit in both Units was 150 ppm. This is additional boron above the 200 ppm required above, resulting in a total soluble boron requirement of 350 ppm. This is well below the minimum spent fuel pool boron concentration value of 2000 pom required by proposed TS 3.7.17 and is, therefore, acceptable.

A separate criticality analysis for a 2-out of.4 checkerooard storage configuration in unborated waterresulted in a 95/95 4 of 0.95741 for Unit 1 and 0.96067 for Unit 2. The soluble boron credit calculations ylalded a 95/95 4 of 0.93835 with the presence of 100 ppm of boron for Unit 1 and 0.94737 with the presence of 50 ppm of boron for Unit 2.

A final configuration was analyzed for Unit 2 which consisted of a 3x3 checkerboard arrangement of cells containing one 3.20 w/o assembly in the etnter surroundM by 1.48 w/o U-235 enriched assemblies. This configuration resulted in a 95/95 4 of 0.99911 in unborated water, thereby meeting the subcriticality acceptance criterion of less than 1.0 with no credit for boron. The amount of soluble boron required to maintain 4 s0.95 was 200 ppm, which resulted in a 95/95 4 of 0.94047.

Storage of assemblies with enric5ments higher than 1.48 w/o U-235 in the peripheral colla of the 3x3 checkerboard configuration was determined using bumup reactivity equivalencing.

Combinations of initial fuel enrichment and discharge bumup, which yield the same storage

-- rack 4 as the rock containing 1.48 w/o U 235 asse.mblies at zero bumup (proposed TS Figure 4.3.14) sequired an additional 300 ppm of boron to account for the uncertainties associated with bumup credit, c

l i

4 I

Storage of assembl!as with e.41chments higher than 3.20 w/o U 235 in the conter cell of tie 3x3 g

checkerboard configuration in the VEGP Unit 2 storage rocks was determined by crediting the reactivity decrease associated with the addition of integral fuel bumable absorbers (IFBAs).

The IFBAs consist of neutron absorbing material applied as a thin Zr8, costing on the outside of the UO, pellet. The fuel assembly it modeled at its most reactive point in life. This includes 1

any time in life when the IFRA has depleted and the fuel assembly becomes more reactive. As i

with bumup credit, for IFBA credit reactivity equivalencing, a series of reactivity calculatior.s are l'

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 3x3 checkerboard configuration analyzed for the VEGP spent fuel racks in Unit 2. 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 uncerbinties adequately conservative and acceptable.

The amount of additl9nal soluble boron needed to account for these uncertainties is bounded by the 300 ppm required for bumup credit in the 3x3 checkerboard configuration.

t Therefore, Wth the above reactivity equivalencing, fuel assemblies with nominal enr;chments up to 5.0 w/o U 235 can be stored in the center of a 3x3 checkerboard configuration by taking credit for a total additlenal amount of soluble boron of 300 ppm. When added to the 200 ppm

.equired without reactivity equivalencing, this results in a total boron requirement of 500 ppm, which is more than the amount required for any of the other storage configurations. However, this is well below the minimum spent fuel pool boron concentration value of 2000 ppm required by proposed TS 3.7.17 and is, therefore, acceptable.

l As an altemative method for determining the acceptability of fuel assembly storage based on -

IFBA loading, the infinite rnultiplication factor, k., was used as a reference reactivity value.

When k. is used as a reference reactivity point, the need to specify an acceptable enrichment versus number of IFBA rods correlation is eliminated. Fuel assemblies with a reference k of 1.410 in the VEGP core geometry at 68 'F have been shown to result in a maximum k,, s0.95 -

when stored in the VEGP spent fuel storage rocks. Therefore, the conter assembly in the 3x3 checkerboard configuration must have an initial nominal enrichment less than or equal to 3.20 w/o U 235,- or satitfy a minimum IFBA roouirement for higher initial enrichments to -

maintah the reference fuel assembly k. less than or equal to 1.410 at 68 'F in the VEGP core geom 64.

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 conditionsi. The first would be a loss of fuel pool cooling system and a rise in pool water temperature from 135 Y to 240 T. The second accidein involves a misloading of an assembly into a cell for which the restrictions on location, enrichment, or bumup are not satisfied.

Calculations have shown that the'misloaded assembly accident in the Unit 2 2 out-of 4 checkerboard results in the highest reactivity increase. The' reactivity increase requires an additional 1200 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 required to ensure protection against a criticality accident. Therefore, the minimum amount of boron required by proposed TS 3.7.17 i

1 h

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

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

6-(2000 ppm)is more than sufficient to cover any accident and the presence of the additional' bornn above the concentration required for normal conditions and reactivity equivalencing (500 ppm maximum) can be assumed as a realistic initial condition since not assuming its presence would be a second unlikely event.

Based on the review previously clescribed, the staff finds the criticality aspects of the proposed VEGP license amendment request are acceptable and meet the requirements of General l

Design Criterion 62 for the prevention of criticality in fuel storage and handling. The analysis a

assumed credit for soluble boron, as allowed by WCAP-14416-NP-A, but no credit for the Boraflex neutron absorber panels. The required amount of soluble boron for osch analyzed storage configuration is shown in Table 1 of this safety evaluation.

2.2 Prana =ad TS AaMa**d with CriHe=Htv Armbia The TS changes proposed as a result of the revised criticality analysis are consistent with the changes stated in the NRC Safety Evaluation (SE) for WCAP-14416-P (Ref. 3). Westinghouse submitted a revised topical report, WCAP-14416-NP A, Rev.1, which incorporated the changes stated in the NRC SE. Also, since the staff disagreed with the propriotsy finding of the original WCAP 14416-P, Westinghout s's revised topical report was submitted as a nonproprietary version.

Proposed TS 3.7.17. " Fuel Storage Pool Boron Concentration," requires that a minimum boron concentration of 2000 ppm be maintained in the spent fuel storage pool. The 2000 ppm concentration is consistent with the criticality analysis and is acceptable. Similarly, the licensee has proposed a limit of 4 <1.0, when the spent fuel racks are flooded with unborated water (in accordance with proposed TS 4.3.1.1b) and a 4 5.95 when flooded with water borated to 450 ppm (for Unit 1) or 500 ppm (for Unit 2) in accordance with proposed TS 4.3.1.1c. These limits on 4 are consistent with the criticality analysis and the proposed TS are acceptable.

Proposed TS 3.7.18, " Fuel Assembly Storage in the Fuel Storage Pool," and proposed TS 4.3,

" Fuel Storage," describe allowable spent fuel storage configurations. The following storage configurations and U 235 enrichment limits for Westinghouse 17 x 17 fuel assemblies were determined to be acceptable:

For VEGP Unit 1:

Assemblies with initial nominal enrichments no greater than 1,79 w/o U-235 can be stored in any cell location. Fuel assemblies with ir.itial nominal enricaments greater than 1.79 w/o U 235 and up to 5.0 w/o U 235 must satisfy a minimum burnup requirement as shown in proposed TS Figure 3.7.18-1, "Vogtle Unit 1 Bumup Credit Requir ements for All Cell Storage."

Assemblies with initial nominal enrichments no greater than 2.45 w/o U 235 can be stored in a 3-out-of 4 checkerboard errangement. Fuel assemblies with initial nominal enrichments greater than 2.45 w/o U 235 and up to 5.0 w/o U-235 must satisfy a minimum bumup requirement as shown in proposed TS Figure 4.3.1-1, "Vogtle Unit i Bumup Credit Requirerr.ents for 3 out-of 4 Storage."

i

l '

7 Assemblies with initial nominal enrichments no greater than 5.0 w/o U 235 can be stored in a 2-cvi of 4 checkerboard arrangement as shown in proposed TS Figure 4.3.14, "Vogtle Units 1 and 2 Empty Cell Checkerbod Storage Configuration."

For VEGP Unit 2:

Assemblies with initial nominal enrichments no greater than 1,77 w/o U 235 can be stored in any cell location. Fuel assemblies with initial nominal enrichments greater than 1.77 w/o U 235 -

and up to 5.0 w/o U 235 must satisfy a minimum bumup requirement as shown in proposed TS Figure 3.7.18-2, "Vogtle Unit 2 Bumup Credit Requirements for All Cell Storage."

Assemblies with initial nominal enrichments no greater than 2.40 w/o U-235 can be stored in a 3-out-of 4 checkerboard arrangement as shown in proposed TS Figure 4.3.1-4. Fuel assemblies with initial nominal enrichments grea'.or than 2.40 w/o U 235 and up 'o 5.0 w/o U 235 must satisfy a minimum bumup requirement as shown in proposed TS Figure 4.3.12, "Vogtle Unit 2 Bumup Credit Requirements for 3-out-of-4 Storage."

Assemtdies with initial nominal enrichments no greater than 5.0 w/o U 235 can be stored in a 2-out of.4 checkerbesrd arrangement as shown in proposed TS Figure 4.3.14.

Assemblies can be stored in a 3x3 checkerboard arrangement consisting of a conter assembly with an initial nominsi enrichment no greater than 3.20 w/o U 235 surrounded by assemblies with initial nominal ennchments no greater than 1,48 w/o U 235, as shown in propose TS Figure 4.3.15, "Vogtle Unit 2 3x3 Checkerboard Storage Configuration." Fuel assemblies with initial enrichments greater than 1.48 w/o U-235 and up to 5.0 w/o U 235 must satisfy a minimum bumup requirement as rhown in proposed TS Figure 4.3.14 or must satisfy a -

minimum IFBA requirement that maintains a maximum reference fuel assembly k. less than or equal to 1.410 at 68 'F.

in order to prevent an undesirable increase in reactivity, the bour daries between the different storage configurst.ons were snelyzed. The interface requirements are shown in proposed TS Figures as follows: TS Figure 4.3.16, "Vogtle Units 1 ano 2 Interface Requirements (All Cell to Checkerboard Storage)," TS Figure 4.3.1-7, "Vogtle Units 1 and 2 Inte ' ace Requirements f, Checkerboard Storage Interface)," TS Figure 4.3.1-8, "Vogtle Unit 2 Interface Requirements (3x3 Checkerboard to All Cell Storage)," and TS Figure 4.3.1-g, 'Vogtle Unit 2 Interface

- Requirements (3x3 to Empty Cell Checkerboard Storage).* These interface requirements are consistent with the criticality analysis and are acceptable.

Based on this consistency with the approved methodology and on the proceding evaluation, the staff fin 64 the proposed TS changes, associated with the criticality analysis, acceptable. The proposed associated bases changes adequately describe these TS changes and are hiso acceptable.

_i.i.,_.._.,...:.m..m.::

.a b.

w i

g

e 5

TACLE 1 Summary of Soluble Boron Credit Requirements for Vogtle Units 1 and 2 Total Soluble Soluble Boron Boron Credit Soluble Boron Required for Required Storage Required for Reactivity Without Configuration k, s 0.95 Equivalencing Accidents (ppm)

(ppm)-

(ppm)

Unit 1 All Cells 200 250 450 3-out-of-4 Checkerboard 200 150 350 2-out-of-4 Checkerboard 100 N/A 100 2-out-of-4 C:.eckerboard 50 N/A 50 3x3 Checkerboard 200 000 500

.g.

lL4 namn rmelan taalvala in accordance with the NRC 3E (Ref. 3) of the Westinghouse methodology described in WCAP.

14416-NP A (Ref. 2), the licensee performed a boron dilution analysis to ensure that sufficient time is available to detect and mitigate the dilution prior to exceeding the 0.95 k, design basis.

The licensee provided boron dilution analyses by letters dated August 8,1997 (Ref.1), and January 16,1998 (Ref. 4). Potential events ware quantified to show that sufficient time is availabe to enable adequate detection and suppression of any dilution event.

l L

Deterministic dilution event calculations were performed for VEGP to define the dilution times and volumes necessary to dilute the spent fuel pool from the minimum TS boron concentration of 2000 ppm to a soluble boron concentration of 600 ppm. This concentration of 600 ppm is consowative with respect to the criticality analysis, which indicated that a soluble boron credit of 500 ppm is sufficient to maintain k,less than or equal to 0.fv5. Unit 1 and Unit 2 spent fuel-pools have a combirwd volume of 772,000 gallons and are normally connscted. However, no co-trols exist to ensure this configuration. Therefore, for the analysis, the spent fuel pools are l

assumed to be separated because it is a more limiting configuration for a boron dilution event.

The volume required to dilute 386,000 gallons in r - *nent fuel pool from the TS limit of -

2000 ppm to 600 ppm is 465,000 gallons. The vmous events that were considered included -

dilution from the utility water, chilled water, domineralized water system, fire protection system, i

- component cooling water system, and chemical volume and control system, and other events that may affect the boron concentration of the pool, such as seismic events, pipe break, and loss of offsite power.

The licensse's evaluation concluded that the most limiting event was a random pipe break of L

the 6-inch fire protection line. This fire protection line provides the largest flow rete of the '

possible dilution sources. Additionally, the water in the fire protection tanks, which contain 600,000 gallons, is sufficient to dilute the spent fuel pool to 600 ppm without replenishment. A break in the 6-inch fire protection line would take approximately 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> at the pump's design flow rate of 2500 gpm to dilute the spent fuel pool to 600 ppm. The fire prctoction piping in the spent fuel pool area is not seismically qualified, but is seismically supported. As such, in accordance with the mechanical engineering and plant systems branch technical positions (BTP) 3-1, it is not required to assume a full break for moderate energy, esismically supported lines. Therefore, the use of the pump t'esign flow rate is a conservative va'ue for this application and is accaptable, in addition to the spent fuel pool level alarms, the fire pump running and low fire protection tank level alarms would provide indication of the event for plant pefsonnel. it is expected that the addition of the large volume of water in the spent fuel pool -

area for this dilution event would be detected by alarms or plant personnel and terminated prior to reaching 600 ppm.

For VEGP, a seismic _ event is a concem for the dilution of the spent fuel pool. The licensee determined that a seismic event could dilute the spent fuel pool in approximately g hours.

This is a concem because the dilution could cecur in the time between personnel rounds (overy

- 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />). Nonseismic piping located in the spent fuel pool room includes the fire protection

=

lines, domineralized water line, chiller water line, and utility water line. Like the fire protection lines discussed herein, these lines are not seismically qualified, but are seismically supported.

i 10-The licensee postulated a through wall crack in the fire protection line with a flow of 168 gpm and a fun break in al' other nonseismic piping. The licensee followed mechanical engineering BTP 3-1 to determire the crack flow rate, it would take approximately 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> at a rate of 1413 gpm for 15 minutes and then A13 gpm thereafter to dilute the spent tuel pool to 600 ppm Also, the licensee will re'tise plant procedures to instruct personnel to specifically check the spent fuel pool room for pipire damage following a seismic event. It is expected that plant personnel would detect the addition of the water in the spent fuel pool a,es prior to reaching 600 ppm i

following a seismic event and the dilution would be, subsequently, terminated, Other evaluated dilution events take longer than 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> to reach the minimum boron concentration. These events would be detected by plant personnel during required rounds overy 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. To detect low flow, long-term dilution events, the licensee will sample its spent fuel pool every 7 days in accordance with proposed TS 3.7.17. This frequency is consistent with the standard TS for Westinghouse plants and is considered adequate for VEGP.

The licensee concluded that an event that would dilete the spent fuel pool boron concentration from 2000 ppm to 600 ppm is not credible. The staff finds that the combirntion of the large volume of water required for a dilution event, TS-controlled spent fuel pool concentration and l

7 day sampling requirement, spent fuel poel alarms and other alarms, pbnt personnel rounds, and otler administrative controls, such as procedured, should adequately detect a dilution event prior to k, reaching 0.95 (600 ppm) and, therefore, the analysis and proposed technical specif. cation controls are acceptable for the boron dilution aspects of the request.

Additionally, the criticality analysis for the spent fuel storage pool shows that k, would remain i

less than 1.0 at a 95/95 probability / confidence level over..!the pool wors completely filled with unborated water. Therefore, even if the spent fuel storage pool was diluted to zero ppm, the racks are expected to remain subcritical.

Based on the review previous ly described, the staff finds the boron dilution aspects of the proposed VEGP license amendment request to be acceptable.

2.4 Pronomad TS Ma*H with Baron DihdM The proposed TS 3.7.17 boron concentratbn of 2000 ppm and 7-day surveillance requirement, together with the proposed remedial action requirements, are acceptable to ensure that sufficient time is available to detect and mitigate the dilution of a VEGP spent fuel pool prior to exceeding the design basis k, of 0.95. Accordingly, proposed TS 3.7.17 is acceptable.

3.0 STATE CONSULTATION

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

( 4.0 EtMRQHMENI6LCQHSIDEB&ILQN 1he amendments cbange requirements with respect to installation or use of a faciMy component located wdhin the rectricted area as defined in 10 CFR Part 20 and change surveillance requirements. 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 no significant increase in h ulividual or cumulative

- occupational radiation exposure. The Commission has previously issued a proposed finding that the amendments involve no significant hazards consideration, and there has been no public commerd on such finding (62 FR 68135 dated December 31,1997). Accordingly, the

- amendments meet the eligibility critoria for categodcol exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b) no environmental impact statem '.t or environmental assessment need be prepared in connection with the issutnce of the amendments.

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 operatien in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public.

Principal Contributors: L. Kopp D. Jackson D. Jaffe D..e:

February 20, 1998

=

==

k s

REFERENCES 1.

C. K. McCoy, SNC, letter to U.S. Nuclear Regulatory Commission, "Vogtle Electric Generating Plant, Response to Request for Additional Information, Revised Request to j

i Revise Technical Specifications, Credit For Boron And Enrict.rnent increase For Fuel Storage," August 8,1997, 2.

W. D. Newmyer, " Westinghouse Spent Fuel Rack Criticality Analysis Methodology,"

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

3.

T. E. Collins, NRC, letter to T. Greene, WOG, " Acceptance for Referencing of Licensing Topical Report WCAP 14416 P, Westinghouse Spent Fuel Rock Criticality Analysis Methodology"(TAC No M93254), October 25,1996.

4.

C. K. McCoy, SNC, letter to U.S. Nuclear Regulatory Commission, "Vogtle Electric Generating Plant, Additional information Fuel Storage Pool Boron Dilution,' January 16, 1998.

l

+

-