ML19256A874

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New Fuel Storage Array-Criticality Safety Analysis. Describes Criticality Safety Analysis of Proposed Reduction in Capacity of Array, & Demonstrates Compliance W/ Suggested Limits
ML19256A874
Person / Time
Site: Palisades Entergy icon.png
Issue date: 07/14/1975
From: Hoffman D
CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:
Shared Package
ML18043A404 List:
References
TASK-09-01, TASK-9-1, TASK-RR XN-309, NUDOCS 7901160260
Download: ML19256A874 (26)
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PALISADES NEW FUEL STORAGE ARRAY CRITICALITY SAFETY ANALYSIS JULY,1975 I

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XN-309 4

PALISADES NEW FUEL STORAGE ARRAY CRITICALITY SAFETY ANALYSIS July, 1975 EXXON NUCLEAR COMPANY, INC.

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XN-309 PALISADES NEW FUEL STORAGE ARRAY CRITICALITY SAFETY ANALYSIS I

Prepared by:

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L. E. Hahsen, Specialist Date Criticality Safety Approved by:

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MI W. E. Niemuth, Manager Date Contract Performance lY V

R. Nilson, Abnager Date Quality Assurance and Licensing

XN-309 TABLE OF CO.VTENTS Page No.

LNTRODUCTION 1

SUhf4\\RY 1

FUEL ASSBBLY DESCRIFFION 2

STORAGE ARRAY DESCRIFFIGN 5

CALCULATIONAL ben 0DS 7

RESULTS 11 CONCLUSIONS 16 APPENDIX A (Second-Party Review) 18

XN-309 INTRODUCTION American Nuclear Society Standard ANS-N18.2 stipulates that new fuel storage arrays be designed to limit the reactivity of such arrays to a value of < 0.98 for all credible conditions of moderation. Further, it is suggested that such conditions of moderation include full flooding or the envelopment of the array in a unifom low density aqueous foam as could credibly exist as a result of fire fighting.

As originally designed, the Palisades new fuel storage array location and design was felt to preclude the addition of moderation. The 3 x 24 array of fuel assemblies, therefore, was not designed to remain sub-critical in the event of water flooding or the addition of low density hydrogeneous materials within and between the stored fuel assemblies.

To comply with suggested limits established in ANS-N18.2; the Consumers Power Company has proposed a reduction in capacity of the array from 72 to 36 fuel assemblies located in a checkerboard array with alternate locations occupied by stntctural steel box beams. This document de-I scribes the criticality safety analysis of. that proposed storage array and demonstrates compliance with the suggested limits established in ANS-N18.2.

SlMtARY Criticality safety analyses of the proposed new fuel storage array con-taining steel box beams in alternate storage locations demonstrate the safety of the array for all credible conditions of moderation. Array 1-

XN-309 reactivities were computed using the KENO-2 bbnte Carlo code assuming fully flooded conditions and the presence of low density aqueous mater-ials unifomly distributed both within and between the stored fuel assemblies. Calculated reactivities are well below the suggested limit established in ANS-N18.2 for new fuel storage arrays. 'Ihe highest reactivity (keff = 0.8961 007) occurs when the array is fully flooded by water.

FUEL ASSBBLY DESCRIPTION A typical Palisades fuel assembly design (Reload E(1)) is depicted in Figure 1.

As indicated, this arrangement includes a single zirconium instnanent guide tube located in the center of the assembly and eight zirconium guide bars positioned on the exterior of the assembly. U02 235 fuel rods within the assembly contain uranium of three different U

enrichments.

Actual " Reload E" fuel assembly specifications and two different assumed conditions evaluated as part of this analysis are given in Table I.

These conditions include existing and bundle averaged cell parameters.

The bundle averaged cell parameters were calculated by including the zirconium associated with the guide bars and instrument guide tube in the :irconium clad thickness of each rod. Water within the assembly was included by increasing the unit cell dimensions (lattice pitch). Such assumptions pemit an estimation of the effect of the extra :irconium and water within the fuel assembly.

It should be noted that the analysis discussed herein assumed a single S

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XN-309 TABLE I PALISADES FUEL ASSDBLY PARAMETERS Bundle Lattice Averaged Actual (1)

Cell Cell (Nominal)

Parameters Parameters lattice Pitch 0.5500" 0.5500" 0.5613" Clad OD 0.4150" 0.4150" 9.4241" Clad Thickness 0.0285" 0.0285" 0.0331" U0 Pellet Diameter 0.3505" 0.3505" 0.3505" 2

Pellet Density (%TD) 94+1.5 96%

96%

Percent Dish 1.0 0%

0%

Avg. Enrichment (wt 's U-235) 3.04 3.20*

3.20*

Rod Array 15x15 15x15 15x15 Design base enrichment specified by Consumers Power Cuarpany( ).

XN-309 bundle averaged enrichment, however, is 3.04 wt % 235U.

Consequently, for the water to fuel volume ratio of these assemblies, this analysis is valid for bundle averaged enrichments of s 3.2 wt % 235U.

STORAGE ARRAY DESCRIPTION The Palisades new fuel storage rack has been measured to detemine actual "as built" dimensions. Figure 2 is an arrangement drawing giving those measured dimensions. This information was supplied by Consumers Power Company (2)

Subsequent infomation(3) established that the dimen-sions indicated are actually measured center-to-center distances between adjacent top bands. Such dimensions, therefore, represent nominal center-to-center spacings between adjacent storage locations.

It should be noted that the nominal center-to-center separation between assemblies was designed to be 9.5 inches. Measurements of the installed rack, however, show that a maximum negative tolerance of 1/8" exists on the design value (i.e., a minimum nominal center-to-center separation of 9-3/8" was measured).

It should also be noted that the plates con-necting adjacent angle irons are 5" wide and 3/16" thick. These plates establish a minimum edge-to-edge separation between adjacent storage locations (i.e., 3/16").

Concrete walls are adjacent

  • to three sides of the storage array and are separated from the fuel by 0.5 to 1.5 inches. For the purpose of this analysis, a 16" thick concrete reflector were assumed to be touching three edges of the storage rack. The fourth side of the array was assumed to be reflected by 4" of water, which is effectively an in-finitely thick water reflector.

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UEL STORAGE I

STALLED RACK

XN-309 The proposed arrangement for storing Palisades fuel assemblies is in-dicated in Figure 3.

Calculations for this arrangement result in the assumption of a 3 x 25 array of storage locations. This " checkerboard" array contains fuel bundles with altemate positions occupied by 8"x8" structural steel box beams having a nominal wall thickness of 5/16"(2)

A minimum wall thickness of 0.25" was established for this analysis (#)

In addition to the nominally spaced array of fuel assemblies shown in Figure 3, it is quite possible that the fuel assemblies will not be centered in each storage location. As previously noted, however, each storage location is bounded by a 3/16" plate. Hence, a minimum edge-to-edge separation between adjacent storage locations is assured. A section of the array considered to evaluate the effect of abnomal arrangements of fuel assemblies is indicated in Figure 4.

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CALCULATIONAL MET 110DS The KESO-II Monte Carlo codeU) was utili:ed to calculate the reactivity of the Palisades new fuel storage array. hbltigroup cross section data (18 energy groups) utilized in these calculations were averaged using the CCELL(6), BRT-1( ), and GAVTEC-II(8) codes.

Specifically, the cross section data for various regions within the storage array were averaged as follows:

CCELL - Utilized to obtain cell averaged multigroup cross section data for fuel rod-water lattices. Such calculations included both the bundle averaged cell parameters and the actual lattice cell parameters (See Table I).

7-

PALISADES NEh* luil. STOlW;E ARRAY (NOMINAL SPACING)

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-9

XN-309 BRT Themal group (< 0.683 ev) cross section data for the stmetural steel box beams were averaged using the Battelle Revised THER50S code. Such data were averaged assuming a 0.25" thick region of iron separated from the rod-water lattice by the thickness of interspersed water (at varying densities) that would exist in the nominally spaced storage array (0.4375"). Epithemal multigroup data were averaged over a slowing down neutron energy distribu-i tion in water.

i GAMrEC-II - Multigroup cross section data for water and con-crete were averaged over neutron energy spectra character-istic of infinite media of these respective materials. The concrete was assumed to have constituents as specified by the Consumers Fvwer Company (9)

(Due to the fact that cross section data for sulfur were not readily available, and i

since the atom density of sulfur represents only 0.06% of the total constituents, the effects of sulfur in the con-crete were ignored.)

In addition to the codes identified above, the XMC bbnte Carlo code was utilized to verify the accuracy of CCELL calculations for undemoderated e

rod-water lattices. The XMC code is a puesdo-point energy bbnte Carlo code (s 9'0' energy groups) which permits the discrete representation of the entire fuel assembly.

XN-309 RESULTS Values of k= were computed using the CCELL code for 3.2 wt % 235U rods as a function of water density within the unit cell. Both lattice cell and bundle averaged cell parameters (See Table I) were assumed to gair, an insight into the reactivity effects of the zirconium guide bars and the water associated with guide bars and the instn2 ment guide tube. The results of those calculations are given in Table II.

Comparison of the calculated values of k= indicates that the uncertainty associated with the :irconium and water extraneous to the actual lattice cells is quite small.

Indeed a maximum effect of 5 mk was calculated assuming these materials were uniformly distributed throughcut the bundle.

i The CCELL code has been used extensively at Exxon Nuclear for averaging multigroup constants in rod-water lattice configurations. Theory-experiment correlations (10) indicate that the reactivity of measured critical U0 rod water lattices can be calculated to within 18 mk.

2 Biases associated with such calculations appear to be consenative in nature.

No experimental data are known to exist at water densities consistent with those assumed in this analysis. As a consequence, in an effort to detemine the validity of the CCELL code for water densities of < 0.6 g/cm, the XMC Monte Carlo code was used to calculate k= of the " Reload E" bundle with all rods assumed to contain uranium enriched to 3.2 wt %

235U.

Water interspersed between the rods was assumed to have a density XN-309 TABLE II Water Infinite Media bbltiplication Factors Densijy Lattice Cell Bundle Averaged (g/cm )

Parameters Cell Parameters i

1.00 1.401 1.402 0.75 1.369 1.371 0.50 1.299 1.302 l

0.30 1.187 1.190 0.20 1.090 1.093 0.15 1.023 1.023 0.10 0.936 0.934 0.05 0.816 0.811 b

e XN-309 3

of 0.2 g/cm. The value of k= calculated for this case was 1.079 ;

.005.

This va> ie compares favorably to that calculated by the CGLL code using bundle averaged cell parameters (1.093) or lattice cell parameters (1.090). These calculations indicate that the CGIJ. code calculates, with reasonable accuracy, the spatial and energy distri-bution of neutrons within the lattice cells at water densities far below the values where experimental data have been utilized to confim the validity of this calculational method.

It appears, therefore, thtt the calculational biases will result in the calculation of array reactivi-ties which are conservative.

For the nominally spaced " checkerboard" array shown in Figure 3, re-activities were computed as a function of water density with such water unifomly distributed both within and between the fuel assembli6s. The results of those calculations are tabulated in Table III and shown graphically in Figure 5.

These reactivities were calculated assuming bundle averaged cell parameters. Figure 5 also includes bundle-averaged values of k= calculated using the CCELL code and,. for comparison, in-dicates the single check calculation performed using the XMC Monte Carlo code.

The highest array reactivity occurs in the fully flooded condition. For all water densities, however, the array reactivity is well below the limiting value of 0.98 established in ANS-N18.2. The highest calculated reactivity (cited at the 95% confidence level) is 0.910. All available checks of the validity of this calculated value indicate a probable conservative bias of 1 to 2% k.

XN-309 TABLE III Fractional Water KENO-II Calculated Density Array Reactivity 1.00 0.896 +.007 0.75 0.825 j;.007 0.50 0.783 +.007 0.30 1.792 +;.006 0.20 J.786 +;.006 0.15 0.789 j;.006 0.10 0.755 j;.006 0.05 0.669 +.006 t

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XN-309 In addition to the nominally spaced array, the array of fuel assemblies spaced as indicated in Figure 4 was also evaluated. For this abnormal arrangement, which assumes the minimum possible spacing between bundles, 3

water densities of 1.0 and 0.2 g/cm were assumed and the resultant a

calculated reactivities for such conditions were 0.914 +.007 and 0.812

+.005, respectively. Hence, even for abnomal conditions, the re-activity of the array remains well below established liraits.

CONCLUSIONS This analysis conservatively demonstrates that the proposed Palisades new fuel storage array will remain subcritical under all conditions of moderation as could credibly exist through the unifom addition of water at any density both within and between the stored fuel assemblies.

It i

is important to note, however, that the array might not be critically safe if moderators are added to individual stored fuel assemblies with-E out the addition of such materials between the assemblies.

It is believed valid to assume that this is not a credible moderator arrange-i ment which could result from any accidental addition of water.

t The analytical efforts described herein were reviewed by an independent second party knowledgeable in the performance of criticality safety evaluations. This independent assessment of the adequacy of this analysis is discussed in Appendix A.

XN-309 REFERENCES 1)

F. D. Lang, G. R. Correll, and K. P. Galbraith, " Final Design Report for Palisades Fuel," XN-74-32, Eu.on Nuclear Company, Inc.,

October, 1974.

2)

Letter, W. J. Beckius to W. E. Niemuth, hhy 13, 1975.

3)

W. J. Beckius, Personal Communication, Consumers Power Company, 7

June 10, 1975.

i 4)

B. Webb, Personal Communication, Consumers Power Company, June 10, 1975.

5)

G. E. hhitesides and N. F. Cross, " Keno - A bbltigroup bbnte Carlo Criticality Program," CTC-5, Union Carbide Corporation hbclear Division, September,1969.

6)

W. W. Porath, "CCELL Users Guide," Bhli/JN-86, Pacific Northwest Laboratories, February, 1977.

7)

C. L. Bennett and W. L. Purcell, "BRT-1: Battelle Revised THER50S,"

BhTfL-1434, Pacific Northwest Laboratories, June, 1970.

I I

8)

L. L. Carter, C. R. Richey, and L. E. Hushey, "GAMTEC-II: A Code for Generating Consistent bbltigroup Constants Utilized in Dif-fusion and Transport Theory Calculations," BNWL-35, Pacific North-west Laboratories,hbrch, 1965.

9)

Letter, W. J. Beckius to W. E. Niemuth, June 3, 1975.

10)

U. P. Jenquin and D. R. Oden, " Verification of Neutronic Design Methods Using Lattice Criticals as Benchmarks," BNW/JN-118, Pacific Northwest laboratories, December,1972.

XN-309 0

6 i

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f APPENDIX A i

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CBare!!e Paoln Northwest Laboratories ove July 11, 1975 h

L. E. Hansen

[hh f rom D. R. Oden Q.A. Review of Criticality Safety Calculations For The Palisades Plant New Fuel Storage Rack i

We have completed the Q.A. review of the criticality safety l

calculations for the Palisades Plant New Fuel Storage Rack, and have found no significant errors or omissions. We concur with your con-clusion that k of the loaded rack, under the co ditions specified, j

eff is less than 0.98 at a 95% confidence level for all degrees of neutron moderation.

Infomation provided by Exxon Nuclear and reviewed by BNW is l

summartzed in Attachments I and II. This consists of letters of cor-respondence, a copy of a drawing of the storage rack, and several computer printouts of Exxon Nuclear calculations. All of this informa-j tion is being returned to you at this time.

j Items checked in the Q.A. review included the following:

  • Code input parameters including nuclei densities for all materials
  • Computer code options selected
  • Assumptions made in problem set up and analysis versus problem definition in correspondence
  • Lattice and array geometry in CCELL and KEN 0
  • Method of homogenization to obtain bundle average parameters
  • Region / material correspondence in CCELL and KEN 0
  • Results of analysis
  • Adequacy of computer codes used.

Concerning this last item of code validation, there is a degree of uncertainty. The CCELL code has gained acceptance as a reliable method of generating soss sections for UO -water lattice through its 2

L. E. Hansen Page 2 July 11, 1975 use at Exxon Nuclear for criticality safety and fuel design work. The KEN 0 code is used throughout the industry for the calculation of arrays of fissile material in criticality safety applications. However, as you are aware, the lack of experimental data on lattices with low density water moderation makes it impossible to completely validate methods of calculation in this area. With no evidence to the contrary, one must thus take the position that if a model is validated for the full water i

density UO lattice criticals it will adequately handle the low water 2

density cases, which may not be tenable.

Although it is not a validation of the array k calculation, the gff fact that the XMC calculated bundle average k= agrees well with that from CCELL for the 20% water density case, lends a degree of credibility to the CCELL cross section generation technique.

In the absence of experimental data this is about all that can be done short of setting up the whole array calculation in XMC. The cost of this approach must however be weighed relative to the benefit resulting from having yet another (although more sophisticated) calculated result.

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i i

ATTACHtENT I REFERENCE LIST OF CORRESPONDENCE AND ORAWINGS PROVIDED BY EXXON NUCLEAR FOR PALISADES STORAGE RACK 0.A. REVIEW l.

Letter, W. J. Beckius to Wayne Niemuth, S/13/75.

2.

Letter, G. R. Correll to L. E. Hansen, " Criticality Analysis of Palisades New Fuel Storage Racks", 5/19/75.

i 3.

Letter, Wayne Niemuth to W. J. Beckius, 5/27/75.

I 4

Lotter, W. J. Beckius to W. E. Niemuth, 6/3/75.

5.

Figure 2.4.2, Reload E Assembly.

6.

Drawing, Palisades New Fuel Storage - Measurements of Installed Rack.

7.

Sketch, Palisades New Fuel Storage Array (Nominal Spacing) 8.

Sketch, Abnormal Array Arrangement.

9.

Tables, Two Pages of Exxon Calculated Results.

10. Figure, Array Reactivity and Bundle Averaged k= Versus Fractional Water Density

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ATTACHMENT II LIST OF COMPUTER PRINT 0UTS PROVIDED BY EXXON NUCLEAR FOR PALISADES STORAGE RACK 0.A. REVIEW Reference i

Number Code Date Title 1

GAMTEC 6/9/75 Palisades Concrete (No Sulfur) 2 THERMOS 6/24/75 Fe thermal data 0% void 25% void 50% void 70% void 85% void 90% void 95% void 3

THERMOS 6/23/75 Fe thermal data 80% void 4 (5 parts) XMC 6/13/75 Bundle k=, 80% void 5

CCELL 6/9/75 Lattice Avg. Cell 0% void 6

CCELL 6/9/75 Bundle Avg. Cell 0% void i

7 CCELL 6/12/75 Bundle Avg. Cell 70% void 50% void 25% vcid 8

CCELL 6/9/75 Bundle Avg. Cell 80% void 85% void 90% void 95% void 9

CCELL 6/9/75 Lattice Avg. Cell 80% void 85% void 90% void 95% void 10 KEN 0 9/25/75 Worst Spacing, 80% void 11 KENO 9/25/75 Worst Spacing, 0% void 12 KENO 9/25/75 0% Void Het. Fe 13 XEN0 9/25/75 25% void Het. Fe 14 KEN 0 9/25/75 50% void Het. Fe 15 KEN 0 9/25/75 70% void Het. Fe 16 KEN 0 9/25/75 80% void Het. Fe 17 KEN 0 9/25/75 85% void Het. Fe la KEN 0 9/25/75 90; void Het. Fe 19 XEN0 9/25/75 95% void Het. Fe

e O

ATTACICE?r2 3

EXAMPLE CAICUIATIO:T OF U235 LI!?AR DEIISITY Based on January, 1979, Exxon Data Active Fuel Length 131.8 + 25 in Weight of UO2 per rod 2117.h8 h

1 2.35 grams Average enrichment (Batch H) 3.27 1 05 High enrichment 3.43 t.05 I.

Using Averare Enrichment (3 27 1 05 = 3.32%)

Weight of Uranium only (U235 + U238)

= (2117.h8 + 42 35)g ((.0332)(235)+(.9E8M238)

.0332)(235)+(.9teo)(23o)+(2)(16)

= (2159.83)g(.8814)

= 1903.67 g Weight of U 35 only 2

= (1903.67)(.0332)g

= 63 2g Lineardensity(g/cm)

" (63.2g of U235/ rod)(208 rods)

(131.0 -.25 in)(2 54/cm/in)

=393hg/cmU235 per Bundle II. Using Hich Enrichment (3.h3 +.05 = 3.48)

Weight of Uranium only (U235 + u238)

= (2117.h8 + k2*35)g (. 3h8)(235)+(.9652)(238)

(.0346)(235)+(.9o52)(236)+(2)(16)

= (2159.83)g(.8814)

= 1903.67g Weight of U235 only

= (1903.67g)(.03h8)

= 66.25g Linear Density g/cm

,(66.25 8U 35)(208 rods) 2 red (131.o.25 in)(2.54 cm/in)

= hl.2k g/cm U 35 per bundle 2

i;cte: Values for active fuel length and weight of CO2 per red was cbtained by phene frem E=en ?;uclear on 1/8/79 These are to be the final design ptrameters.

2-CAICJIL ION OF U235 LINEAR DE"SITY Based on October,1978, Enon Data Active Fuel Length 131.8 i.25 in Weight of UO2 per rod 2101.2 g=s Average Enrichment 3.27 +.05 High Enrichment 3.43 1 05 I.

For Average Enrichment Weight of Uranium only (U235 + U238)

= (2101*2g) (( 0332)(235)+(.9668)(238) 0332)(235)+(.9666)(230)+(2)(16)

= (2101.2g) (.8814)

= 1852. cog Weight of Uranium - 235 only

= (1851 99g)(.0332)

= 61.h9g Linear Density (g/cm) 61.h9 g U235

" ((131.0

/ red)(208 rods)

.25 in)(2.54 cm/in)

= 38.28 g/cm U 35 per bundle 2

II.

For High Enrich =ent Weight of Uranium only (U235 + U 38) 2

= (2101.2g) (.03h8)(235)+(.9652)(238)

(.03kc)(235)+(9c52)(230)+(2)(16)

= (2101.2g) (.8814) 1852.00 g

=

Weight of U 35 only 2

= (1852.0g) (.03h8)

= 64.h5g Linear Density dem

" (6h.h5g-U 35/rodl208 reds) 2 (131.6.25 in)(2.5h c=/in)

= ho.12 g/c= U 35 b udle 2

r 3-CALCULATION OF U235 LUTEAR DEISITY Ensed on Batch G Fu-l Data Active Fuel Length (in) 131.8 +.25 Weight of UO2 (c=s) 2104.0 + 2%

Average Enrich =ent (Batch G) 3.00 t.05 Eigh Enrichment 3 20 +.05 I.

Weight of Uranius Only (U235 + u238)

= (1*02)(210h*0) ((.0325)(235)+(.9675)(238)

.0325)(235)+(.9675)(25c)+(2)(16)

= (21k6.c8)(.8814)

= IS91.56g Weight of U 35 only 2

= (1891 56g)(.0325) 61.h8g

=

Linear Density (g/cs)

= (61.h8g U 35/ rod)(2c8 rods) 2 (131.0 -.25 in)(2.54 c=/in)

= 38'.27 g/c= U 35 per bundle 2

Our present Technical Specificetions (for Batch G Fuel) assume the highest enrichment and limit us to a maximum linear density of 38.3 g/cm U235,

9 4

ATTACICCIT 4

~,

53

ruCLEAR s m : SUPPLY SYSTri (::sss) (Contd) 5.3.2 Reacter Core and Centrol The reactor core shall approximate a right circular cylinder with a.

an equivalent dia=eter of about 136 inches and an active height of about 132 inches.

b.

The reactor core shall consist of approximately 43,000 Zircaloy-4 clad fuel rods containing slightly enriched uranium in the forn of sintered UO2 pellets. The fuel rods shall be grouped into 204 assemblies.

A core plug or plugs may be used to replace cne or more fuel assemblies subject to the analysis of the resulting power distribution.

The fully loaded core shall contain approximately 211,000 pounds UO c.

2 and approminately 56,000 pounds of Zircaloy h.

Poisen =sy be placed in the fuel bundles for long-term reactivity control.

d.

The core excess reactivity shall be controlled by a cc bination of boric acid che=ical shim, cruciform control reds, and techanically fixed boron rods where required. Forty-five control rods shall be distributed throughout the core as shown in Figure 3-5 of the FSAR.

Four of these centrol rods may consist of part-length absorbers.

5 3.3 Energency Core Coolin: System An emergency core cooling syste= shall be installed consisting of various subsyste=s each with internal redundancy. These subsystems shall include four safety injection tanks, three high-pressure and two low-pressure safety injection pumps, a safety injection and refueling vater storage tank, and interconnecting piping as shown in Section 6 of the FSAR.

5.4 FUEL STORAGE 5.k.1 :Tev Fuel Storage Unirradiated fuel may be stcred in the new fuel stcrage rack which a.

is designed to ensure an effective =ultiplication factor of less than 0 98 under the vorst credible cenditiens for fuel enriched to 3.30 veight percent U-235 5-3

p G

5.4 FUEL STORAGE (Centd) b.

New fuel =ay be stored in shipping containers.

c.

New fuel enriched to 3.27 veight percent U-235 =ay be stored in the poi-soned high capacity racks which are designed to ensure an effective =ulti-plication factor of less than 0 95 when flooded with unborated water.

d.

The new fuel storage racks are designed as a Class I structure.

5.h.2 Spent Fuel Storaae a.

Irradiated fuel bundles vill be stored, prior to off-site shipment in the stainless steel-lined spent fuel pool.

b.

The spent fuel racks are designed to maintain fuel in a gecretry which insures an effective =ultiplication facter of 0.95 or less with new fuel flooded with unborated water.

c.

The spent fuel pool water boron concentration shall be verified at least once =onthly to be equal to or greater than 1720 pp=.

d.

The spent fuel racks are designed as a Class I structure.

The fuel placed in the spent fuel pool and stored in the poiscned high e.

capacity storage racks shall not centain more than kl.2h grams of U-235 per axial centimeter of active fuel asse=bly subject to a maxirem as-sembly average loading of 3 27 veight percent U-235 The fuel placed in the spent fuel pool and stored in the unpoisoned lever capacity racks shall not contain more than 38.3 grs=s of U-235 per axial centimeter of active fuel asse=bly, subject to a maxi =u= asse=bly average loading of 3.05 veight percent U-235 f.

Spent fuel shipping casks shall not be moved in the fuel stcrage building until such time as the NRC has reviewed and approved the spent fuel cask drop evaluation. ~

g.

Fuel stored in the higher capacity storage racks as described in the SER supporting A=end=ent No. 26, shall have decayed for a mini =u= of 12 =enths if the storage racks are not supported by sir.ilarly designed, adjacent rac3s or the cask anti-tipping device.\\l)and the spent fuel pool vall References (1)Until needed for fuel storage, two A-type racks in the northeast corner of the spent fuel pool vill be removed and replaced with the cask anti-tipping device to provide necessary seis=ic restraint.

FSAR, Appendix A.

FSAR, Appendix 3.

5L

NOIrrHI!AST trrII.rFIIES

]

no.uw.'ca.vw i o" **

P.O. Box 270 n4 a w < nn oa mn m a o ww HARTFORD, CONNECTICUT 06101 INC.$lC[UN[

(203) 666-6911 L

L J %::2T3li"O'!%~

January 8,1979 Docket No. 50-336 Director of Nuclear Reactor Regulation Attn:

Mr. R. Reid, Chief Operating Reactors Branch #4 U. S. Nuclear Regulatory Comission Washington, D. C.

20555 Gentlemen:

Millstone Nuclear Power Station, Unit No. 2 Proposed Revisions to Technical Specifications Pursuant to 10CFR50.90, Northeast Nuclear Energy Company (NNECO) hereby proposes to amend its operating license, DPR-65, by incorporating the following proposed revisions into the Millstone Unit No. 2 Technical Specifications:

Revise Section 3/4.7.8, Hydraulic Snubbers, and Table 3.7-1 as shown in Attachment 1.

This change proposes to delete the generic applicability statement which requires the hydraulic snubbers listed in Table 3.7-1 to be operable while the plant is in Modes 1 through 4.

As an alternative, a new column has been added to Table 3.7-1 which provides the specific modes of operation for which snubber operability is actually required. Similarly, the single ACTION statement, which requires the plant to be in HOT STANDBY within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> in the event of an inoperable snubber, has been deleted and replaced with the six more appropriate ACTION statements listed on Page 3/4 7-29.

The deletion of these overly restrictive generic requirements will allow greater operating flexibility during the repair and/or maintenance of safety-related hydraulic snubbers, without compromising plant safety. The philosophy employed in developing this proposed change was based extensively on a review of the modes for which the plant system involved is required to be operable by Technical Specifications. This ce.a:ept is explaired below.

The safety-related snubbers can be grouped into the following eleven (11) categories:

(1) High Pressure Safety Injection (HPSI)

(2) Low Pressure Safety Injection (LPSI)

(3) Containment Spray (CS)

(4)

Shutdown Cooling (SDC)

(5) Safety Injection Tanks (SIT)

(6) Spent Fuel Pool (SFP) 7 90116 0 2M P

(7) Reactor Building Closed Cooling Water (RBCCW)

(8)

Service Water (SW)

(9)

Steam Generators (SG)

(10) Feedwater System (FEED)

(11) Main Steam (MS)

Justification for the applicable modes and action statements are provided below for each category.

(1) High Pressure Safety Injection No change. These snubbers are required to be operable in Modes 1 through 4 as under the present requirements.

In the event of an inoperable snubber, ACTION statement 1 is required which, in this case, is also the same as the present specifications.

(2) Low Pressure Safety Injection LPSI system snubbers are required to be operable in Modes 1, 2, 3(+), and 6.

The LPSI system as an ECCS subsystem is required to be operable in Modes 1, 2, and 3 (with the pressurizer pressure > 1750 psia designated by (+)) by Limiting Condition for Operation (LCO) 3.5.2.

Also, some snubbers designated as part of the LPSI system are common to the Shutdown Cooling system and are, therefore, required to be operable in Mode 6.

ACTION statement 4 identifies the appropriate action to be taken in the event of an inoperable snubber depending on the plant's operating mode.

(3) containment Spray LCO 3.6.2.1 requires two operable containment spray systems in Modes 1, 2, and 3(+).

The applicable modes for snubber operability have been revised accordingly.

ACTION statement 1 requires that the plant ultimately be in a mode not requiring snubber operability if the snubber cannot be restored to CPERABLE status.

(4) Shutdown Cooling LCO 3.9.8 requires that a SDC loop.Le operable in Mode 6.

The operability of SDC snubbers listed in Table 3.7-1 has been changed to reflect the required operability of the Shutdown Cooling System itself. The ACTION statement for inoperable SDC snubbers (No. 3) is similar to the action required in the event of an inoperable Shutdown Cooling System as in LCO 3.9.7.

(5)

Safety Injection Tanks The operability requirements of SlT snubbers have been revised to conform to the requirements of the SIT system itself, specifically, Modes 1, 2, and 3(+).

(See LCO 3.5.1). ACTION statement 1 requires that the plant ultimately be in a mode not requiring snubber operability if the snubber cannot be restored to OPERABLE status.

(6)

Spent Fuel Pool Spent fuel pool cooling is not required unless irradiated f uel assemblies are in the storage pool. ACTION statements 5 and 6 have been written to specifically respond to the inoperability of the three SFP snubbers listed in Table 3.7-1.

~

. (7) Reactor Building Closed Cooling Water No change.

(8) Service Water No change.

(9) Steam Generators LCO 3.4.5 requires each steam generator to be operable in Modes 1, 2, and 3.

The applicable modes for the steam generator snubbers have been revised accordingly. ACTION statement 1 requires that the plant ultimately be in a mode not requiring snubber operability.

(10) Feedwater System The snubbers on feedwater lines, in general, are required tu be operable in Modes 1, 2, and 3, unless the line containing the hanger is isolated from the af fected steam generator. The ACTION statement for these snubbers (No. 2) allows the isolation of the line containing the hanger from the affected steam generator or, the plant must be in a MODE not requiring operability of the snubber.

For certain feedwater snubbers, it has been determined that the snubbers are re-quired to be operable regardless of whether the line is isolated from the steam generator; the inoperability of these snubbers could af fect other lines. There-fore, these snubbers are required to be operable in Modes 1 through 3 and ACTION statement 1 is applicable.

(11) Main Steam As described above under "Feedwater System", the snubbers on main steam lines are similarly required to be operable in Modes 1, 2, and 3, unless the line containing the hanger is isolated from the affected steam generator. Again, certain snubbers are required to be operable regardless of whethr.r the line is isolated from the steam generator. The ACTION statements are the same as for the Feedwater System snubbers.

It is emphasized that although in most cases, snubber operability requirements are proposed to be relaxed, plant safety is not compromised. This is most readily demonstrated by the fact that the mode applicability determination and action statement wording is based upon the LCO for the plant system or component under consideration. A hydraulic snubber on a plant system is clearly not as vital as the system itself.

In addition to the above described revisions to the applicability modes and action statements, this review has resulted in several non-safety related snubbers being identified and, therefore, deleted from Table 3.7-1.

The snubbers which Fave been deleted and the justification for the deletion are provided below.

. I.

HANGER NO.

SYSTEM SNUBBER INSTALLED ON, LOCATION, AND ELEVATION 411010 FEED-E/22/+41 413021 MS-D/16/+43 413036 MS-E/18/+48 413039 MS-E/21/+49 490001(2)

MS-D/19/+28 490002(2)

MS-D/19/+25 490003(2)

MS-C/19/+27 490004(2)

MS-D/19/+25 490005(2)

MS-C/19/+33 490006 MS-D/19/+33 490007 MS-D/19/+32 490008 MS-C/19/+32 490018(2)

MS-D/17/+33 490019 MS-D/17/+33 490031 MS-B/17.1/+46 511001 FEED-D/22/+37 51119-R27(2)

MS-D/17/+45 51119-R28(2)

MS-B/17/+43 51119-C29 MS-D/17/+45 The above snubbers are installed on Main Steam and Feedwater Lines which are both designed to the requirements of seismic category II piping. For these lines in these areas, the only safety concern is that the postulated failure of these lines during a seismic event not affect the performance of any safety-related equipment. In each of these areas, this protection is provided by pipe whip restraints which provide far greater assurance of the integrity of these lines as they-affect saf ety related systems during a seismic event.

II.

HANGER NO.

SYSTEM SNUBBER INSTALLED ON, LOCATIONS, AND ELEVATION 310022 HPSI-46S/21E/-7 427115(2)

SW-L.5/15.9/-14 These snubbers are mechanical snubbers, which have been inadvertently included in the specifications for hydraulic snubbers.

The above proposed changes have been reviewed pursuant to 10CFR50.59 and have not been found to constitute an unreviewed safety question.

The Millstone Unit No. 2 Nuclear Review Board has reviewed and approved the above proposed changes, and concurred in the above determination.

NNECO has reviewed the above proposed license amendment pursuant to the require-ments of 10CFR170, and has determined that the proposal constitutes a Class 3 amendment. Accordingly, enclosed herewith is payment in the amount of $4,000

. (Four Thousand Dollars). The basis for this determination is that the proposal involves a single safety issue which does not involve a significant hazards consideration.

Very truly yours, NORTHEAST NUCLEAR ENERGY COMPANY

'? {

?VNAA W. G. Counsil Vice President Enclosure

STATE OF CONNECTICUT )

)

ss. Berlin

u. [ /9.79 COUNTY OF HARTFORD

)

Then personally appeared before me W. G. Counsil, who being duly sworn, did state that he is Vice President of Northeast Nuclear Energy Company, a Licensee herein, that he is authorized to execute and file the foregoing information in the name and on behalf of the Licensees herein and that the statements contained in said information are true and correct to the best of his knowledge and belief.

h. AM Notary Public My Commissi:n Expires March 31,1931

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