ML20153G234
| ML20153G234 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 08/30/1988 |
| From: | DUQUESNE LIGHT CO. |
| To: | |
| Shared Package | |
| ML20153G205 | List: |
| References | |
| NUDOCS 8809080085 | |
| Download: ML20153G234 (35) | |
Text
.. _.
ATTACHMENT A
Revise the Beaver Valley Power Station,. Unit 2
Technical Specifications a follows:
Remove Pace Insert Pace 3/4 9-14 3/4 9-15 B 3/4 9-3 B 3/4 9-3 5-6 5-6 5-7 5-7
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3/4.9.14 FUEL STORAGE - SPENT FUEL STORAGE POOL
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LIMITING CONDITION FOR OPERATION 3.9.14 Fuel is to be stored in the spent fuel storage pool with:
a.
The boron concentration in the spent fuel pool maintained greater than or equal to 1050 ppm when moving fuel in the spent fuel pool; and b.
Fuel assembly storage in Region 1 restricted to fuel with an enrichment less than or equal, to 4/,K v/o 5/cru/ in a 74f c4ecl(e,deaw/ cc, by
- rehd.j ooscf i
c.
Fuel assembly storage in Region 2 restricted to fuel which has been qualified in accordance with Table 3.9-1 APPLICABILITY:
During storage of fuel in the spent fuel pool.
ACTION:
a.
Suspend all actions involving movement of fuel in thG spent fuel pool if it is determined a fuel assembly has been placed in the incorrect Region until such time as the correct storage location is determined.
Move the assembly to its correct location before resumption of any other fuel movement.
b.
Suspend all actions involving the movement of fuel in the spent fuel pool if it is determined the pool baron concentration is less than 1050. ppm, until such time as the boron. concentration is increased to 1050 ppm or greater.
c.
The provisions of Specifications 3.0.3 and 3.0.4 aro not applicable.
SURVEILLANCE REQUIREMENTS 4.9.14.1 Prior to placing fuel or moving fuel in the spent fuel pool, verify through fuel receipt records for new fuel or by burnup analysis and comparison with Table 3.9-1 that fuel assemblies to be placed into or moved in the spent fuel pool are within the above enrichment limits.
4.9.14.2 Verify the spent fuel pool boron concentration is 1 1050 ppm:
a.
Within 8
hours prior to and at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> during movement of fuel in the spent fuel pool, and b.
At least once per 31 days.
DEAVER VALLEY - UNIT A
3/4 9-l'4 FAG (G5ED LOOdDINV
~
1 Table 3.9-1 BEAVER VALLEY FUEL ASSEMBLY MINIMUM BURNUP VS. INITIAL U235 ENRICHMENT FOR STORAGE IN REGION 2 SPENT FUEL RACKS Assembly Dische.rge Initial U Enrichme235 t
Burnup (GWD/WrU) 7, 4 0
A. S
$0
//A'
.C3 1/.tf U?
HCfrE Linear interpolation yields conservative results.
l BEAVER VALLEY - UNIT 2.
3/4 9-15 PLotosEo wou!NV
REFUELING OPERATIONS BASES 3/4.9.10 and 3/4.9.11 WATER LEVEL - REACTOR VESSEL AND STORAGE POOL The restrictions on minimum water level ensure that sufficient water depth is available to remove 99% of the assumed 10% iodine gap activity released from the rupture of an irradiated fuel assembly.
The minimum water depth is consistent with the assumptions of the accident anarlysis.
3/4.9.12 and 3/4.9.13 FUEL BUILDING VENTILATION SYSTEM
.The limitations on the storage pool ventilation system ensure that all radioactive material released from an irradiated fuel assembly will be filtered through the HEPA filters and charcoal adsorber prior to discharge to the atmosphere. The OPERABILITY of this system and the resulting iodine removal capacity are consistent with the assumptions of the accident analyses.
The spent fuel pool area ventilation system is non-safety related and only recircu-lates air through the fuel building.
The fuel building. portion of the SLCRS is safety related and continuously filters the fuel building exhaust air.
This; maintains a negative pressure in the fuel building, s
3/4.9.14 FUEL STORAGE - SPENT FUZL STORAGE POOL The requirements for fuel storage in the spent fuel pool ensure that (1) the spent fuel pool will remain subcritical during fuel storage; and (2) a uniform boron concentration is maintained in the water volume in the spent fuel pool to provide negative reactivity for postulated accident conditions under the guidelines of ANSI 0.95 or less for k,dencefg which includes all 16.1-1975.
The value of probability /confi level is the uncertainties at the 95/95 acceptance criteria for fuel storage in the spent fuel pool.
The Action Statement applicable to fuel storage in the spent fuel pool ensures that:
(1) the spent fuel pool is protected from distortion in the fuel storage pattern that could result in a critical array during the movement of fuel; and (2) the boron concentration is maintained at y.
1050 ppm (this includes a 50 ppm conservative allowance for uncertainties) during all actions involving movement of fuel in the spent fuel pool.
The Surveillance Requirements applicable to fuel storage in the spent fuel pool ensure that (1) the fuel assemblies satisfy the analyzed U-235 enrichment limits or an analysis has been performed and it was determined that k
is
<0.953 and (2) the boron concentrationmeetsthe1050ppmIbg
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Beaver Valley - Unit 2 B 3/4 9-3
(% U'C5 & D W CR u w 6~
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DESIGN FEATURES DESIGN PRESSURE AND TEMPERATURE 5.2.2 The reactor containment building is designed and shall be maintained for maximum internal pressure of 45 psig and a temperature of 280.0'F.
PENETRATIONS 5.2.3 Penetrations through the reactor containment building are designed and shall be maintained in accordance with the original design pN visions contained in Section 6.2.4 of the FSAR with allowance for normal degradation pursuant to the applicable Surveillance Requirements.
5.3 REACTOR CORE i
FUEL ASSEMBLIES 5.3.1 The reactor core shall contain 157 fuel assemblies with each fuel assee-bly containing 264 fuel rods clad with zircaloy-4.
Each fuel red shall have a nominal active fuel length of 144 inches.
Reload fuel shall be similar in i
physical design to the initial core ladingandshallhaveamaximumenrichmentj l
ofgweightpercantU-235, 4
CONTROL R00 ASSEMBLIES 5.3.2 The reactor core shall contain 48 full length and no part length control rod assemblies.
The full length control rod assemblies shall contain a nominal i
142 inches of absorber material.
The nominal values of absorber material shall be 80 percent silver, 15 percent indium and 5 percent cadmium.
All control i
rods shall be clad with stainless steel tubing, j
5.4 REAC' TOR COOLANT SYSTEM DESIGN PRESSURE AND TEMPERATURE 5.4.1 The Reactor Coolant System is designed and shall be maintained:
a.
In accordance with the code requirements specified in Section 5.2 of the FSAR, with allowance for normal degradation pursuant to the applicable Surveillance Requirements, b.
For a pressure of 2485 psig, and c.
For a temperature of 650*F, except for the pressurizer which is 680'F.
VOLUME 5.4.2 The total water and steam volume of the Rescto' rcolant System is i
9370 cubic feet at a nominal T,yg of 576'F.
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BEAVER VALLEY - UNIT 2 5-6 i
Mokstb acatNc-
DESIGN FEATURES i
5.5 EMERGENCY CORE COOLING SYSTEMS 5.5.1 The emergency core cooling systems are designed and shall be maintained in accordance with the original design provisions contained in Section 6.3 of the FSAR with allowance for normal degradation pursuant to the applicable Surveillance Requirements, 1ke fuel WU/ be S}crel s' accen/a m ch$ W$e i
n 5.6 FUEL STORAGE g,, g,,
4 3.jf,j,; pg ge e,A9,,, ej, y,gg,j CRITICALITY 5.6.1 The spent fuel storag<
racks are designed and shall be maintained with a minimum of 10.4375 inch eq 1ter-to-center distance between fuel assemblies placed in the storage racks ho ensure a k,ff equivalent to 10.95 with the storage pool filled with unborated water.
TM h,77 ef $0.95 i=1 & : ::=:-":tiv:
i
-:11=:=: Of :t 1:::t 1.4% t.k/k fer.7,;;rteir. tie:.
DRAINAGE E
5.6.2 The spent fuel storage pool is designed and shall be maintained to prevent inadvertent draining of the pool below elevation 751'-3".
CAPACITY I
5.6.3 The fuel storage pool is designed and shall be maintained with a storage capacity limited to no more than 1088 fuel assemblies.
1 5.7 SEISMIC CLASSIFICATION 5.7.1 Th*ose structures, systems and components identified as Category I items in Section 3.7 of the FSAR shall be designed and maintained to the ori.ginal de-sign provisions with allowance for normal degradation pursuant to the applicant
' Surveillance Requirements.
5.8 METEOROLOGICAL TOWER LOCATION 5.8.1 The meteorological tower shall be located as shown on Figure 5.1-1.
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1 4
BEAVER VALLEY - UNIT 2 5-7 fkQf6GD ldGADIVV
ATTACHMENT B
Safety Analysis Beaver Valley Power Station, Unit No. 2 Proposed Technical Specification Change No. 15 Description of amendment request:
The proposed amendment would incorporate Section 3.9.14 and associated bases and revise Design Feature Sections 5.3.1 and 5.6.1 to set forth fuel assembly U-235 enrichment limitations on storage of fuel in the new and spent fuel storage racks.
These changes are based on an evaluation performed by Westinghouse, "Criticality Analysis of Beaver Valley 2 Fuel Racks."
The results of the evaluation provide justification for:
1.
New fuel storage rack enrichment limit of 4.85 w/o, 2.
Two spent fuel storage rack enrichment limits where Region 1 limits fuel enriched from 3.6 to 4.85 w/o to a three out of four cell checkerboard storage
- pattern, and fuel assemblies can be stored in all cells in Region 2 limited by the burnup dependent restrictions provided in Table 3.9-1.
Criticality of fuel assemblics in a
fuel storage rack is prevented by the design of the rack which limits fuel assembly interaction.
This is done by fixing the minimum separation between assemblies.
The design basis for preventing criticality outside the reactor is
- that, including uncertainties, there is a 95 percent probability at a
95 percent confidence level that the effective (K gg) of the fuel assembly array will be multiplication factor o
less thu 0.95 as recommended in ANSI 57.2-1983 and ANSI 57.3-1983.
For accident conditions where reactivity is postuldted to increase (i.e.,
misloading an assembly with a burnup and enrichment combination outside of the acceptable criteria provided in proposed Table 3.9-1, or dropping a fuel assembly between the rack and pool wall),
the double contingency principle of ANSI 16.1-1975 is applied.
This states that one is not required to assume two
)
- unlikely, independent, concurrent ovants to ensure protection against a
criticality accident.
Thus, for accident conditions, the presence of soluble boron in the storage pool water can be assumed as a realistic initial condition since not assuming its presence would be a
second unlikely event.
The presence of approximately 1000 ppm boron in the spent fuel pool will decrease reactivity by about 15 percent AK.
- Thus, for postulated accidents, should there be a
reactivity increase, Kogg would be less than or equal to 0.95 due to the effect of the dissolved boron.
i i
m ATTACHMENT B P2ga 2 K gg including uncertainties at the 95/95 The maximum e
probability / confidence level is presented for the limiting cases:
l Case Eeff 1.
Spent Fuel Rack Region 1, 4.85 w/o
.9417 3 of 4 cell storage i
2.
Spent Fuel Rack Region 2, 3.6 w/o
.9486 all cell storage 3.
Fresh Fuel Rack, 4.85 w's
.9264 moderation - full density 8
1.0 gm/cm 4.
Fresh Fuel Rack, 4.85 w/o
.9398 moderation - optimum low density 8
O.076 gm/cm K gg for each of the above limiting cases is less than 0.95 t
The e
including uncertainties at the 95/95 probability / confidence icvel, therefore, the acceptance criteria for criticality is met under all i
,l conditions.
In accordance with the criticality analysis results, technical specification limitations on maximum enrichment are applicable for d
the spent fuel racks.
Fuel assembly storage in Region 1 is limited to a
maximum enrichment of 4.85 w/o in a 3 of 4 cell array.
Fuel assemblics can be stored in all cells in Region 2, limited by the burnup dependent restrictions provided in Table 3.9-1.
Both the spent and new fuel racks are analyzed for an enrichment limit of 4.85 w/o.
Since technical specification limits have been placed on fuel assembly enrichment for storage in the spent fuel pool, no additional technical specification restrictions are required on the new fuel racks.
Design Feature section 5.3.1 has boon revised to reflect the new fuel assembly enrichment limit of 4.85 w/o, and section 5.6.1 was revised to reference the applicable FSAR sections which describe the provisions for fuel storage.
FSAR sections 4.3 and 9.1 are being revised to reflect the new criticality analysis which includes a j
description of the uncertainties applied.
Therefore, the sentence describing the uncertainties is not required and has been deleted.
Storage of fuel in the new and spent fuel racks will be changed to reflect the "criticality Analysis of Beaver Valley Unit 2 Fuel Racks".
The criticality analysis supports the storage of fuel anriched up to 4.85 w/o U-235.
This will facilitate longer fuel
- cycles, higher nuclear capacity factors and lower plant power generation costs.
}
ATTACHMENT B Pago 3 Spent fuel pool Region 1 will provide for storage of fuel with enrichments up to 4.85 w/o in an administratively controlled 3 cf 4 cell array.
Region 2
will provide for storage of fuel assemblics with the burnup dependent enrichment limitations provided in Table 3.9-1.
Keff will be maintained less than 0.96 consistent with the current FSM design basis.
With the Region 1 checkerboard array, the segregation of fuel assemblies into Regions 1 and 2 and the proposed technical specification changes, no adverse safety considerations are introduced.
The new criticality analysis satisfies the design basis for preventing criticality outside the reactor
- where, including uncertainties, there is a 95% probability at a 95% confidence level that Koff of the fuel assembly array will be less than 0.95 in accordance with ANSI 57.2
- 1983Property "ANSI code" (as page type) with input value "ANSI 57.2</br></br>- 1983" contains invalid characters or is incomplete and therefore can cause unexpected results during a query or annotation process. and ANSI 57.3 - 1983.
Therefore, the proposed changes will not reduce the safety of the plant and are consistent with the current regulatory basis.
4 1
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ATTACHMENT C
No Significant Hazards Evaluation Beaver Valley Power Station, Unit No. 2 Proposed Technical Specification change No. 15 Basis for proposed no significant hazards consideration determination:
The Commission has provided standards for determining whether a
significant hazards consideration exists in 10 CFR 50.92(c).
A proposed amendment to an operating license for a facility involves no significant hazards if operation of the facility in accordance with the proposed amendment would not (1) involve a significant increase in the probability or consequence of an accident previously evaluated, (2) create the possibility of a
new or different kind of accident from any accident previously evaluated, or (3) involve a
significant reduction in a
margin of safety.
The proposed changes do not invclve a significant hazards consideration because:
1.
The criticality analysis acceptance criteria (K gg < 0.95) e is consistent with that stated in FSAR Sections 9.1.1 New Fuel
- Storage, 9.1.2 Spent Fuel Storage and 4.3.2.6 Criticality of the Reactor During Refueling and Criticality of Fuel Assemblies.
Attachment D provides a revision to FSAR section 9.1.1 and 9.1.2 to describe the segregation of the spent fuel pool into regions 1 and 2 and how the Region 1 administrative controls ensure that the 4.85 w/o fuel and the 3 of 4 cell array is maintained.
In addition to the administrative controls available to maintain the required checkerboard array in Region 1,
the minimum boron concentration will provide an additional safety nargin to ensure criticality will not be achieved.
Even if new fuel assemblies were not stored in the specified checkerboard
- array, the dissolved boron would provide sufficient neutron absorption capability to preclude criticality.
Attachment E
provides a
revision to FSAR Section 4.3.2.6 to incorporato changes to reflect the now criticality analysis.
These FSAR changes are provided as background information for this technical specification change and will be included in a future FSAR update.
Fuel assembly decay heat production is a function of core power
- level, and since the authorized coro power level is not being
- changed, the decay heat load on the spent fuel pool cooling system will not be significantly impacted by the proposed onrichment limits.
The proposed changes will not have a significant impact on the safety of the plant or on the operation of the spent fuel storage pool.
The critoria setforth in Table 3.9-1 provide assurance that fuel assemblics are qualified for storage in Region 2
to ensure Kaff will be 1
0.95 at the 95/95 confidence lovel.
Therefore, the proposed changes will not introduce any adverse safety considerations or involve a
significant increase in the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated.
ATTACHMENT C Pzga 2 2.
The proposed changes are bounded by FSAR Section 15.7.4 Radiological Consequence of Fuel Handling Accidents and the activities in the fuel rod gap presented in Table 15.0-7 which use a
conservative value of 650 days at a full power value of 2766 MWt to determine fission product inventories and calculate resultant doses.
In accordance with the double contingency principle of ANSI N16.1-1975 it is not required to assume two unlikely, independent, concurrent events to ensure protection against a
criticality accident.
Therefore, the minimum boron concentration limits on the spent fuel pool ensure that even if new fuel assemblies were not spaced to maintain the checkerboard arrays or a
fuel assembly was dropped on top of the rack (the rack structure portinent for criticality is not excessively deformed and the dropped assembly has more than twelve inches of water separating it from the active fuel height of stored assemblies) that criticality would be precluded.
The analysis of reactor core operation with up to 4.85 w/o reload fuel will be provided in the cycle-specific reload safety evaluations which are performed for each reload cycle (the standard reload design methods described in WCAP-9272 and
- 9273, "Westinghouse Reload Safety Evaluation Methodology",
and/or other appropriate criteria to demonstrate that the core reload will not adversely affect the safety of the plant).
Criticality accidents during fuel handling are precluded by stringent administrative procedures which require the qualification of fuel assemblies in accordance with Table 3.9-1 for fuel assembly storage in Region 2.
Therefore, the probability for an accident or malfunction of a different type than previously evaluated will not be created.
3.
Technical Specification 3.9.14 and associated bases provide the administrative controls required to assure that fuel assemblics with the potential to form a critical array are segregated such that the offective multiplication
- factor, K gg, will be less than 0.95.
Criticality will be prevented o
in Region 1 by limiting fuel assembly interaction by physical design of the fuel racks and maintainjng a minimum solubic boron concentration in the pool wattr.
Fuel assembly placement in Region 1 will be administratively controlled by storing fuel with an enrichment betwoon 3.6 and 4.85 w/o in a 3
of 4
cell array.
Whoro Region 1 is adjacent to Region 2, the arrangement will be maintained to limit fuel assembly interaction.
This is consistent with the design basis critoria for proventing criticality outside the reactor where, including uncertainties, there is a 95% probability at a 95%
confidence level that K gg of the fuel assembly array will o
be less than 0.95 in accordance with ANSI 57.2 - 1983 and ANSI 57.3 K gg will be maintained less than 0.95 1983.
o including uncertaintics consistent with the current design basis.
Therefore, the proposed changes will not involvo a significant.cduction in the margin of safety.
Thorofore, based on considerations expressed
- above, it is proposed that this amendment application does not involvo a
significant hazards consideration.
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ATTACHMENT D
FSAR Changes (provided for information only) s
O BVPS-2 FSAR Ejected rod worths.are given in Section 15.4.8 for several different conditions.
Allowable deviations due to misaligned control rods are discussed in the Technical Specifications.
A representative calculation for two banks of control rods simultaneously withdrawn (rod withdrawal accident) is given in Figure 4.3-36.
Calculation of control rod reactivity worth versus time following reactor trip involves control rod velocity and differential reactivity worth..
The rod position versus time of travel after rod.
release normalized to "Distance to Top of Dashpot" and "Drop Time to Top *of Dashpot" is given on Figure 4.3-37.
For nulcear design purposes, the reactivity worth versus rod position is calculated by a series of steady-state calculations at various control rod positions, assuming all rods out of the core as the initial position in order to minimize the initial reactivity insertion rate.
- Also, to.be conservative, the rod of highest worth.is assumed stuck out of the-core, and the flux distribution (and thus reactivity importance) is assumed to be skewed to the bottom of the core. The result of these calculations is shown on igure 4.3-38.
The shutdown groups provide additional negative reactivity to assure an adequate shutdown margin.
Shutdown margin is defined as the amount by which the core would be suberitical at hot shutdown if all rod cluster control assemblies are tripped, but assuming that the highest worth assembly remains fully withdrawn and no changes in xenon or boren take place. The loss of control rod worth due to the material irradiation is negligible, since only bank D may be in the core under normal operating conditions (near full power). The values given in Table 4.3-3 show that the available reactivity in withdrawn rod cluster control assemblies provides tF - design bases minimum shutdown margin, allowing for the highest
- rth cluster to be at its fully withdrawn position. An allowance for the uncertainty in the calculated worth of N-1 rods is made before determination of the shutdown margin.
i 4.3.2.6 Criticality of the Reactor During Refueling and Criticality i
of Fuel Assemblies Criticality of fuel assemblies outside the reactor is precluded by adequate design of fuel transfer, shipping and storage facilities, l
and by administrative control procedures.
The two principal methods of preventing criticality are limiting the fuel assembly array size and limiting assembly interaction by fixing the minimum separation between assemblies and/or inserting neutron poisons between assemblies.
]
i The design basis for preventing criticality outside the reactor is that, considering possible variations, there is a
95 percent
.I 4.3-33 J
BVPS-2 FSAR fj)Sr C7,2-M93 AME [7.3-Mff3 ca) G yk lettet A N/ If /975' f
i probability at a 95 percent Iconfidence level that the effective multiplication factor (k gg) ofgthe fuel assembly array will be less e
than 0.95 as recommended in l::: N2:0 1070.
The following are the conditions that are assumed in meeting this design basis:
1.
The fuel assembly contains the highest enrichment authorized without any control rods or any noncontained burnable poison and is at its most reactive point in life.
":fer ::
-4*e44 - * ' '
(+r-fuel---prope r t ie s-u sed-i n--e ri t4ee M 6y-
- 1cul: tier.:.
of MN temperature W[ conditions, the moderator is pure water at-6he-a For flooded 2.
ithir. the d::ig. limit ich yi:1d:
th: -
/,0
,k,' i asd 1
ft consco*vefra Vo l.s e W;e;;tn:,;h'ivi'y,d w*s.t v.
f
-lk deM V
3.
The array is either infinite in lateral extent or is surrounded by a conservatively chosen reflector, whichever is appropriate for the design.
4.
Mechanical uncertainties are treated by either using "worst case" conditions or by performing sensitivity studies and obtaining appropriate uncertainties.
5.
Credit is taken for the neutron absorption in structural materials and in solid materials added specifically for neutron absorption.
6.
Where borated water is present, credit for the dissolved boron is not taken, except under postulated accident conditions where the double contingency principle of ANSI N16.1-1975 is applied.
This principle states that it shall require at least two unlikely, independent, and concurrent events to produce a criticality accident.
For fuel storage application, water is usually present. However, the design methodology also prevents accidental criticality when fuel assemblies are stored in the dry condition.
For this case, possible i
sources of moderation, such as those that could arise during fine l
fighting operations. are included in the analysis. 45.e design-basis-i -0.?! :s-r44emmended-in-ANGhHH4-4M4s tac e tao ~d m totk kef4 g6/u
$n bt IY 0MIb *4YVQM (CY!
- b ON'r $ # NW dNU $MU'l Y
C " ^
, j y/ w The design method which ensures the criti6ality safety of fue
- ($,, g assemblies ut:id: th: re::ter uses the AMPX system of codes (Ford 44
- 4,44 4 and Greene :: :1-1976) for cross section generation and M+ENOIV(PetrieandCross1975)forreactivitydetermination.
fgy 217
$72 The 444-energy group cross-section library (Ford $4-+N), that is 54//d-V Tbommon starting point for all cross-sections, has been the generated r/P:'l data.
The NITAW:. program (Greene :: :1 1976) includes in this library the self-shielded resonance cross-sections that are appropriate for particular geometry.
The Nordheim Integral Treatment is used. ' Energy and spatial weighting of cross-sections is cod 4.3-34 1
BVPS-2 FSAR performed by the XSDR!lPH program (Greene et :1 1976), which is a one-dimensional S transport theory code.
These multi-group cross-g section sets are then used as input to KENO IV (Petrie and Cross 1975), which is a three-dimensional Monte Carlo theory program designed for reactivity calculations.
31 A set of 47 critical experiments has been analyzed using the above method to demonstrate its applicability to criticality analysis and to establish the method bias and variability.
The experiments range from water moderated oxide fuel arrays separated by various materials that simulate LWR fuel shipping and storage conditions 4Sierman-et-41-(Gade. /179) 4W7-and-194F to dry harder spectrum uranium metal cylinder arrays with various interspersed materials (Thomas 1973) that demonstrate the vide range of applicablity of the method.
13 Some descriptive facts about each of the -M benchmark critical k gg of the 0,9q 6-exp_eriments are given in Table 4.3-4.
The average e
benchmarks is* 0.0003.whi:P d:::n:trate: that-there-i: virtually no-bi : :::::i:ted uith th: :th:d.
Th: :t nd d devte44en-e(-the--ke
'1' g' value: i:
- 0. 0%7,A The 95/95 one sided tolerance limit factor for 33Dvalues iiiNr2(n There is thus a 95 percent probability with a 95 percent confidence level that the uncertainty in reactivity due to the method is not greater than g ak.
--The total unecrtainty te b: :dd:d te :- cr4t-ioeM4y ::1:ulation-is+-
DMRT1 5
2 2
e,a 2 Tr -
(ke
. yg w;c r
= ch W re (k:)
!! 0 012 :: discu::ed-ebev:, (h:)
-eta tistioal suomnoouneeete4nty-essociated---with--the--partiN$er is-the--
v
---ME NO--
- c e l c ule44en-b+4n9--us ed--and-the (ke)pecn t07 ' O '*~*~*
- F i * *~' I'~
stetistical un;ertsinties esse ict;d wit., m:chanicci televen ::
- uch--
- thic k neeees-and :pecings.
If "ucret-eese%ssumpt-ions-ere-ttsed-
-f o r-t o le r a nc e sr-t h is-t e rm-will-be-s e ro.-
-TE : critic:lity d::ign cr4t+r4: :::.et %:.;-th: ::leulet+d-ef fect ive-
-entidies44en-feeter plu:-th: t:::1 un :rt+inty (!'.')
i:
1 :: : hen-0.95 er, in th: :peci:1 :::: defin;d :bov, 0.03.
t Su-l913 These methods conform with ANS fil.2 'l 73, Nuclear, Safety Criteria for the Design of Stationary Pre surized Water Reactor Plants, Section 5.7, Fuel Handling System: J S! "210-1975, Design Objectives 4*47 J WR Spent Fuel Storage Facilities at Nuclear Power Stations, Section as ANSI Nii.9-1975, validation of Calculational Methods for Nuclear Criticality Safety:
NRC Standard Review
- Plan, Section 9.1.2, Spent Fuel Storage; and the NRC guidance, eview and Acceptance of Spent Fuel Storage and Handling Applications'*
M S F 5"7.)- 195 2, 'yos h p ue w is Q f k h et t h y e j,Q pm'y Q al L kl vhlu ksek fledc.
nacilikes 4
i 4.3-35 i
I
s b' TGAY.L The following equatio'n is used to develop the maximum keff' i
9
)
K.ve =
K.., n + B wn.. + B.,, + (( (k s) 8 en + (k s) 8 m.ines where:
worst case KENO K.ee'that includes centered fuel
=
assembly positions, material tolerances, and j
mechanical tolerance which can result in spacing 4
betw len assemblies less than nominal method bias determined from benchmark critical i
=
comparisons
'blas to account for poison partical self-shielding j
=
95/95 uncertainty in the worst case KENO K.e4
=
ks in.e 95/95 uncertainty in the method blas
=
i The criticality acceptance; criteria is met when the effective multiplication l
f actor (k,ff),. including. unc'ertainties at : a)95/95.'prebability/ confidence
[
level'is less than O'.95.
i j
i i
i P
f
}
l 1
4 4
l l
BVPS-2 FSAR are obtained from the three dimensional TURTLE calculation from which constants are homogenized by flux-volume weighting.
Validation of the spatial codes for calculating power distributions involves the use of incore and excore detectors and is discussed in Section 4.3.2.2.7.
Based on comparison with measured data, it is estimated that the accuracy of current analytical methods is:
10.1 percent 60 for Doppler defect 12 x 10 5/*F for moderator coefficient 150 ppm for critical boron concentration with depletion 13 percent for power distributions 10.2 percent 60 for rod bank worth 14 pcm/ step for differential rod worth 10.5 pcm/ ppm for boron worth 10.1 percent 60 for moderator defect 4.3.4 Revisions The design methods for the criticality of fuel assemblies outside the reactor now use the AMPX/ KENO system of codes as described in Section 4.3.2.6.
The design methods for the nuclear analysis of the core now use both TURTLE (Barry and Altomare 1975) and PALADON (Camden et al 1978) for multi-dimensional analyses.
4.3.5 References for Section 4.3
- Barry, R.F.
1963.
LEOPARD A Spectrum Dependent Non-Spatial Depletion Code for the IBM-7094. WCAP-3269-26.
Barry, R.F. et al 1975. The PANDA Code. WCAP-7048-P-A (Proprietary) and WCAP-7757-A (Honproprietary).
- Barry, R.F.
and Altomare.
S.
1975.
The TURTLT 24.0 Diffusion Depletion Code.
WCAP-7213-P-A (Proprietary) and WCAP-7758-A (Nonproprietary).
Ot:r :,,
0,R.
21 !?"
_CriL1:01 50; rat!ct SO P e -- "'*:riti:21
-eauet:r; ef 2. M ut *, ' *" S:d: ir. " te r ith-Fi;;d Neutreet--Pessetter-
--BetteM -i;;ifi: ': the:: Ld:::teri:: ?"L-2%
Amendment 12 4.3-45 June 1986
O BVPS-2 FSAR
[ ace >$bf Oi rmer.,
0.R.
- t--el-1974.
Criti:1 S:ptr:tien 50te::n Sch riti::1 7
(,a s-Clu:::r: :f 5.2? et t
'2*L' ?:d: in " ter eith Fired M::tr:n P i::::.
- tt:lle P :ifi: N rthu::t L:h:r: tori:: PNL-261&r Camden, T.M. et al 1978. PALADON - Westinghouse Nodal Computer Code.
WCAP-9485A (Proprietary) and WCA1 v486A (Nonproprietary).
y Ce rmak, J.C.
et al 1968. Pressurized Water Reactor pH - Reactivity Effect Final Report. WCAP-3696-8 (EURAEC-2074).
- Drake, M.K.
(Ed) 1970.
Data Formats and Procedure for the ENDF/B Neutron Cross Section Library BNL-50274, ENDF-102 Vol. 1.
Eggleston, F.T.
1977.
Safety-Related Researen and Development for i
j Westinghouse Pressurized Water Reactors Program Summaries l
Winter 1976, WCAP-8768, Revision 1.
a A one-Point Depletion and Fission l
- England, T.R.
1962.
CINDER Product Program. 'WAPD-TM-334.
,Flatt, H P.
and Buller, D.C.
1960.
AIM-5, A Multigroup, one Dimensional Diffusien Equation Code. NAA-SR-4694.
gep/get v$d T rd, L'.:..
- :t :1 107..
A 210-Cr ;; M:;ts:n Cr :: S::ti n
'I d'T 4#0 A. Library in the 1"?v "::ter Inter!:::
Format fer -Crit &4444ty 5 fety-Etudi::. CP?i/CSD/TM-t.
- Greene, N.M. et 21 1??S. AMPX: A Modular Code System for Generating Coupled Multigroup Neutron-Gamma Libraries from ENDF/Bf CRNL/TM-3706, /fack M76
- Hellman, J.M.
and Yang, J.W. 1974. Effects of Fuel Densification Power Spikes on Clad Thermal Transsents. WCAP-8359.
Hellman, J.M. (Ed) 1975. Fuel Densification Experimental Results and Model for Reactor Application.
WCAP-8218-P-A (Proprietary) and l
WCAP-8219-A (Nonproprietary).
- Langford, F.L.
and Nath, R.J.
1971.
Evaluation of Nuclear Hot
]
Channel Factor Uncertainties.
WCAP-7308-L (Proprietary) and i
WCAP-7810 (Nonproprietary).
?
~
Leamer.
R.D.
et al 1967.
PUO -UO Fueled tritical Experiments.
2 g
WCAP-3726-1.
- Lee, J.C.
et al 1971. Axial Xenon Transient Tests at the Rochester i
Gas and Electric Reactor. WCAP-7964.
I W
1 Amenduent 12 4.3-46 June 1986
)
i l
l,
o BVPS-2 FSAR McFarlane, A.F.
1975.
Power Peaking Factors, WCAP-7912-P-A (Proprietary) and WCAP-7912-A (Nonproprietary).
- Meyer, C.E.
and Stover, R.L.
1975.
Incore Power Distribution Determination in Westinghouse Pressurized Water Reactors. WCAP-8498.
- Meyer, R.0.
1976. The Analysis of Fuel Densification.
Division of System Safety, USRNC, NUREG-0085.
- Moore, J.S.
1971.
Power Distrib9 tion Control of Westinghouse Pressurized Water Reactors. WCAP-7811.
- Moore, J.S. 1971a. Nuclear Design of Westinghouse Pressurized Water Reactors with Burnable Poison Rods. WCAP-7806.
- Morita, T.,
et al 1974. Power Distribution Control and Load Follow Procedures. WCAP-8385 (Proprietary) and WCAP-8403 (Nonproprietary).
- Nodvik, R.J. et al 1969. Supplementary Report on Evaluation of Mass Spectrometric and Radiochemical Analyses of Yankee Core I Spent Fuel,
,J Including Isotopes of Elements Throium Through Curium WCAP-60SC.
CWg pg[e,g rec
- 01hoeft, J.E.
1962. The Doppler Effect for Non Unifonn Temperatgre Distribution in Reactor Fuel Elements WCAP-2040.
j Petrie, L.M. and Cross. N.F. 1975. KENO IV - An Improved Monte Carlo Criticality frogram. ORNL-4938.
- Poncelet, C.G.
1966.
LASER A Depletion Program for Lattice Calculations Based on MUFT and THERMOS. WCAP-6073.
- Poncelet, C.G.
and Christie, A.M.
1958.
Xenon-Induced Spatial Instabilities in Large PWRs WCAP-3680-20 (EURAEC-1974).
- Skogen, F.B.
and McFarlane, A.F.
1969a.
Control Procedures for Xenon-Induced X-Y Instabilities in Large PWRs.
WCAP-3680-21 (EURAEC-2111).
Skogen.
F.B.
and McFarlane.
A.F.
1969b.
Xenon-Induced Spatial Instabilities in Three-Dimensions. WCAP-3680-22 (EUr.AEC-2116).
- Suich, J.E.
and Honeck, H.C. 1967. The RAMMER System Retergeneous Analysis by Multigroup Methods of Exponentials and Reactors.
DP-1064.
- Thomas, J.T.
1973.
Critical Three-Dimensional Arrays of U (93.1) -
Metal Cylinders.
Nuclear Science and Engineering, Volume 52, pp. 350-359.
Westinghouse 1974.
Westinghouse Anticipated Transients Without Reactor Trip Analysis. WCAP-8330.
Amendment 3 4.3-47 October 1983
BIBLIOGRAPHY 1.
Nuclear Regulatory Commission, Letter to All Power Reactor Licensees, from B. K. Grimes OT Position for Review and Acceptance of Spent fuel Storage and Handling Applications, April 14, 1978.
2.
W.
E.
Ford llt, CSRL-V:
Processed ENDFIB-V 227-Neutron-Group and Pointwise Cross-Section Libraries for Criticality Safety, Reactor and Shielding Studies, OMNL/CSDITM-160, June 1982.
3.
N.
M.
Greene, AA1PX: A Aiodular Code System for Generating Coucled A1ultigroup Neu ron-Gamma Libraries from FNDFIB, ORNLITM-3706, March 1976.
4.
L. M. Petrie and N. F. Cross, KENO IV--An Improved Afonte Carlo Criticality Program, ORNL-4938, November 1975.
5.
M. N. Baldwin, Critical Experiments Supporting Close Proximity Water Storage of Power Reactor fuel, B AW-1484-7, July 1979.
6.
J. T. Thomas, Critical Three-Dimensional Arrays of U(93.2) A1etal Cylinde's, Nuclear Science and Engineering, Volume 52, pages 350-359,1973.
7.
A. J. Harris, A Description of the Nuclear Design and Analysis Programs for Boiling Water Reactors, WCAP-10106, June 1982.
8.
Askew, J. R., Fayers, F. J.,
and Kemshell, P. B., A General Description of the Lattice Code WIAfS, Journal of British Nuclear Energy Society, 5, pp.
564-584, 1966.
9.
England, T.
R., ClNDER A One-Point Depletion and Fission Product j
Program, WAPD-TM-334, August 1962.
- 10. Melenan, J.
B.,
yankee Core Evaluation Program Final
- Report, WC AP-3017-6094, January 1971.
Bibliography 36
f h
TA8tE 4.3-4 BENCHMARK CRSTl CAL EXPERIMtNTS
^ s.
- - -. - - ~
Gene ra l Enrichment Sepa ra t ing Cha racte riz i
\\
Descript ion wt1 U235 Reflector Me teria l Seoarati el 1 i.
Red..ttice 2.35 W.t.r Wat.r j
.,2 4
2.
00d lattice 2.35 Water Water 8.39 3.
UOd Rod las ce 2.35 Water Water 6.39 4
UOd Roe lattice 2.35 Water Water 4.46 5.
UOd Rod lattice 2.35 Water stainless st 10.44 6.
UOd Rod lattice 2.35 Water Staint steet 11.47 7.
UOd Rod sattico 35 Water S
nless steel 7.76 8.
UOd Rod fattice 2.35 Water stainless steel 7.42 9.
00d Rod lattice 2.35 Water Dora l 6.34
- 10. UOd Rod Iattice 2.35 Wetor 80ra I 9.03
- 11. UOd Rod lattice 2.35 ter Bora l 5.05
\\
- 12. UOd Rod lattice 4.29 Wate Wa ter 10.64
- 13. UOd Rod Isttico 4.29 Water Stainless steel 9.76
)
- 14. UOd Rod latt!ce 4.29 Water Stainless steel 8.08
- 15. UOd Rod lattice Water ra l 6.72
- 16. U MetoI cyIenders 93.2 Sere Air 15.43
- 17. U Metal cyIInders 93.2 Pa ra rrin Air 23.84
- 18. U Metal cylinders 93.2 Ba re Air 19.97
- 19. U Metas cyli rs 93.2 Pa ra rrin Air 36.47
- 20. U Meta yl inders 93.2 Bare Air 13.74
- 21. U tal cwlinders 93.2 Pa rarrin Air 3.48 i
\\
. U Meta l cyl inde rs 93.2 Bare Plexiglass 15.
-~
~
1 of' 2 m
_ ~...... -. - -. -. ~.-_
.... ~ -.- - - _ __ _ _.--
SVPS-2 FSAft TABLE 4.3-4 (Cont)
~
/
General E n rictumen t sepa ra ting Cha ractog,
j
,/
300 w11 W3S fleflector
_ leteria l
__. von Icel p
/
- 23. U Peetal cyl irule rs 93.2 Pa ra f fin Plexi I 25.43 i
24 U poe ta l cyl inders 93.2 Se re Plexig la ss 21.74 j
)
- 25. U Metal cylinders 93.2 Pa ra ffin Plexiglass 27.94
- 26. U peetal cylinder 93.2 Se re Steel 14.74 tal cylinders 93.2 Sare Plexiglass, steel t
l l
l l
l l
l l
2 or 2 I
h
1 i
1 1
I Table 1.
Benchmark Critical Experiments (5,6) l l
1 i
l l
-f/INYY $ - -_
l B
/
/
c eeeal enelet-at seeseating Sole e N
/
eeste tot ten
_e/o U235.
netletter
_seteial -
secon rs=
a,,
N t.
UO ros let t ite 2.46 ealee estec 0
0.9957 *.002e 2.
to ecus lat t tte 2.46 estee eatee 1037 0.9906 T. Colt 3.
tC Porf latttte 2.46 s a t ee ester 764 0.9896 T.CCIS 4
UO rod latttte 2.46 e a t er B*C pine O
O.ggle T 0025 S.
LC cod lattice 2.46 e s t ee Bec pine O
O.9491 T. 0026 i
6.
VO ec=1 latttte 2.46 e at ee 54C pint O
O.9955 T.0020 3
7 UO rcus lat t tee 2.86 e at er 84C ping O
O.9999 T.CO26 8.
tC rod tatttte 2.44 satee 8*C pint O
O.9903 7.0025 9.
to red lattice 2.86 eatee eatee O
O 9936 7.0020 10.
to ros fatttte 2.86 eatee estee 143 0.992g 7,0025 it.
LC rod lattite 7.46 satee statntets steet Sl4 0 18967 7.0020 12.
UO rod latttte 2.44 eater stalntest steel 7t?
O 9943 7.cong 13.
UO tcut tat t Ice 2.86 estre terat W aluminum 15 0 9992 Y.0023 14 UO red tatttte 2.46 satte boested aluminum 92 0 9994 T,0023 15.
to tod tattIce
- 2. ( ;
ester boret M stuminum 395 0.9432 7.002 16.
VO etes lat t tte 2.s4 eater tuoreted aluminum 120 0.9440 f.CO24 i
17.
VO rot letttte 2.44 estee terated aluminum 89?
O.9995 T,cotc t
IG.
to ecus lat t lee 2.46 eater ter st ed s t un t eve t97 0.9405 T.Cott 19.
UO Ptwf letttte 2.46 estee toest H aluminum 634 0.9921 T,pg 20.
VO ros lattice 2.46 satee terated stuminum 320 0.9920 T,cnto f
1 29.
VO roi lattice 2.e4 ester tuorated aluminum 12 0 992g 7.002o 22.
U tal cyttmoere 93.2 bare ele O
O.990g T,ooto 23.
U owt al cyllneers 93.2 bare ele O
O.9973 T.copo 24 U metal tylinders 93.2 bare air O
O.994 7 7.CO29 25.
U owtel tyllao*et 93.2 bar e ate O
O.992g T.CC tt 24.
U me t a l ty l t artee t 93.2 bare ate O
O 9922 T,ootg 27 U owt el tyline*et 93.2 bare air O
O.9950 T.CO27 i
29.
U metal cylladers 93.2 bar e plestglegg O
O 994 T,0030 1
29.
U met al tyl tadert 93.2 paraffin pleatglatt O
O.9929 T.CCel 30.
U owtal cyttno.es g3.2 bar e pleelglass O
O.9968 T.CCil 34.
U ental tyllerteet 9.1. 2 paraftle steelgtese O
I,0042 T,00 3 32.
U ental tyltewteet 93.2 paraffin plestglegg C
O 9963 T,co3o 33.
IJ metal tylinders 93,2 paraH in pleetglegg O
O.'/919 I.0032 i
-s#
l I
16
i BVPS-3 FSAR 4
CHAPTER 9 AUXILIARY SYSTEMS 9.1 FUEL STORAGE AND HANDLING 9.1.1 New Fuel Storage i
i The new fuel storage area is located in the fuel area shown on Figures 9.1-1, 9.1-2, and 9.1-3 and is designed to provide a safe, effective means for dry storage of the new fuel assemblies.
9.1.1.1 Decign Bases The new fuel storage area is designed in accordance with the following criteria:
1.
General Design Criterien 2, as it relates to the ability of i
structures housing the facility components to withstand the j
effects of natural phenomena such as earthquakes, tornadoes, hurricanes, and floods.
f i
t 1
2.
General Design Criterion 5, as it relates to shared structures, s.ystems, and components important to safety i
being capable of performing required safety functions.
l l
j 3.
General Design Criterion 61, as it relates to the facility
]
design for fuel storage.
1 l
4.
General Design Criterion 62, as it relates to the prevention t
of criticality by physical systems or the process utilizing ll geometrically safe configurations.
l 5.
Regulatory Guide 1.29, as it relates to the seismic design classification of facility components.
.I 6.
The new fuel storage facility is designed to store i
i sufficient fuel for one refueling (one-third core) plus j
17 spares, for a total of 70 assemblies, and maintain the l
4 fuel suberitical (X,gg <0.95) when fully loaded and flooded j
with non-borated water.
With fuel of the highest i
anticipated enrichment, assuming optimum moderation, the j
effective multiplication factor (X,gg) vill not exceed M,
l 1
.W.
j 9.1.1.2 Facilities Description
{
lI 5
i The r.ew fuel storage area is shown on Figures 9.1-1. 9.1-2, and l
f 9.1-3.
New fuel storage is provided for one-third core (53 fuel assemblies) plus ;7 spare assemblies. New fuel assemblies are stored i
l dry in a steel and concrete structure within the fuel building.
The i
new fuel storage racks consist of a stainless steel support structure 1
I l
Amendment 3 9.1-1 October 1983 1
9
BVPS-2 FSAR i
i l
into which 70 stainless steel fuel guide assemblies are bolted in
{
14 parallel rows of five fuel guide assemblies each.
It is not possible to insert a fuel assembly in other than a prescribed location due to the d1 sign of the fuel guide supporting structure.
{
The spacing of the new fuel assemblies, located in the new fuel guide I
assemblies, is a minimum of 21 inches center-to center.
Fuel assemblies are loaded into the fuel guide assemblies through the top.
Ldequate guidance is provided in each fuel guide assembly by means of
(
a flared lead-in opening to preclude damage to the fuel assemblies during insertion or withdrawal.
The accumulation of liquid in the
.j new fuel stcrage area is prevented by a 4-inch floor drain located in j
the area.
9.1.1.3 Safey Evaluation i
The new fuel storage area is located in the seismic category I fuel l
building. Handling of new fuel is done by a separate 10-ton hoist on the motor driven platform erane (section 9.1.4).
j New fuel assemblies are -tored vertically, with a minimum center to f
center spacing of 21 inches.
This wi.11 maintain the fuel in a l
suberitical condition with the effective multiplication factor, K,fg
[
1ess than 0.95, "he-the net 'u:1 :::::g: :::: i: 'ully 1::d:d
- r-t th; :::.:g cr:: i: '1::d:d with r.:r. b re;;d ;;;;r "ith ' :1 ' th:
l 4., m
.o 4 + o. A.
4,.u.o
...u4-
-.4-u-m e -
de'N//9ediI]udeNhb'8/y /c h "illnet:r55:'d^^!. Ic " la bks ndenhe corulNs.
l
& loe de.an)y (0 0%y-/c cf The new fuel storage racks are designed to seismic category I requirements. A detailed analysis of the storage racks have been pe rforme d to verify the adequacy of the design to withstand the loadings encountered during normal operation, an operating basis t
earthquake (CBE), and the safe shutdown earthquake (SSE).
The motor-driven platform crane,*hich is used for transfer of fuel, i
is the only overhead crane whic!. een pass over the new fuel.
j i
Damage to the fuel assemb.ies and the new fuel racks by excessive f
uplift forces from the new fue' handling hoist are prevented by operating procedures and by a load cell attached to the crane.
In I
addition, the new fuel storage area is protected from the effect of drcpped heavy objects by interlocks on the fuel handling hoist, which limit the lifting capability of the crane to the weight of a fuel i
cell and its handling tool.
Heavier loads will be handled by an f
administrative procedure, which will define the area over which these loads may ba handled to prevent damage to the new fuel.
i i
i l
t I
Amendment 11 9.1-2 January 1986 r
i l
L
BVPS-2 F2AR 9.1.2 Spent Fuel Storage The spent fuel storage area, located in the fuel building shown on Figures 9.1-1 and 9.1-2, is designed to provide a safe and effective means of storing spent fuel.
9.1.2.1 Design Bases The spent fuel storage area is designed in eccordance with the following criteria:
1.
General Design Criterion 2, as it relates to structures housing the facility and the facility itself being capable of withstanding the effects of natural phenomena, such as earthquakes, tornadoes, hurricanes, and floods.
2.
General Design criterion 4, as it relates to structures housing the facility and the facility itself being capable of withstanding the effects of environmental conditions, 1
external missiles, internally generated missiles, pipe whip, I
and jet impingement forces associated with pipe breaks, such l
that safety functions will not be precluded.
3.
General Design Criterion 5, as it relates to shared
]
structures, systems, and components important to safety being capable of performing required safety functions.
l 4.
General Design criterion 61, as it relates to the facility design for fuel storage and handling of radioactive materials.
5.
General Design criterion 62, as it relates to the prevention of criticality by physical systems or processes utilizing geometrically safe configurations, j
6.
General Design criterion 63, as it relates to monitoring systems provided to detect conditions that could result in j
the loss of decay heat removal capabilities, to detect excessive radiation levels, and to initiate appropriate 1
safety actions.
1 7.
The spent fuel storage area is designed in accordance with the requirements of Regulatory Guides 1.13, 1.29, 1.115, and 1.117.
.1 9.1.2.2 Facilities Description i
The spent fuel storage area is divided into three areas separated by a stainless steel-lined concrete wall, with a removable gate provided between each area to allow movement of fuel elements between them.
]
Each gate is equipped with an inflatable. seal to prevent leakage from one area to another. The three areas are defined as the fuel cask 1
]
9.1-3 1
BVPS-2 FSAR
- area, the spent fuel pool, and the fuel transfer canal. Each area is lined with stainless steel and is normally filled with borated demineralized water.
The fuel cask area consists of two locations at different elevations, which allow for the safe movement of spent fuel into the shipping cask.
The lower elevation provides a sufficient height of water above the fuel being transferred to allow for adequate shielding, while the upper elevation limits the potential spent fuel cask drop height and allows for preliminary decontamination using a floating spray ring.
The spent fuel pool houses the spent fuel storage racks, which provide sufficient space to store spent fuel from a total of 17 refuelings. plus the storage of one full core in the event the reactor must be emptied of fuel at any time during BVPS-2 life.
The spent fuel racks consist of 17 rack assemblies, each having a storage capacity of 64 spent fuel eleoents. Total spent fuel pool storsge capability is 1.088 spent fuel elements.
The spent fuel racks consist of two parts, a subbase beam system and the 17 individual rack assemblies.
The system of interconnected base beams is provided to bridge the space between embedment pads so load transfer from the racks to the floor occurs only through the embedment pads. Each spent fuel rack, consisting of an 8x8 array of storage cells, is bolted to the base beams.
Because the entire complement of base beams and 8x8 racks form a single structural unit, relative sliding between racks is eliminated. However. the base beam system is free to slide since it is not connected to the embedment AT))D plates.
The storoge racks are positioned such that adequate clearances
- t. r e provided between the racks and pool walls to avoid Jgf[g3 im
> pacting during seismic events.
The fuel transfer canal houses the fuel transfer system which provides for transfer of new and spent fuel elements between the fuel building and reactor containment during refueling. Spent fuel is transported between the fuel transfer canal, spent fuel pool, and the fuel cask area by the fuel building motor-driven platform crane.
This platform incorporates separate 10-ton hoists for new fuel and spent fuel. A complete description of fuel handling and utilization of the movable platform with hoists is provided in Sections 9.1.4 and 9.1.5.
Handling of the spent fuel casks utilizes the spent fuel cask trolley and is described in Section 9.1.5.
Normal makeup water for the spent fuel pool is provided by the primary grade water system.
Berated makeup water may be supplied from the refueling water storage tank (RWST) through the fuel pool cleanup system, as described in Section 9.1.3.
Boron concentration is normally maintained at 2,000 ppm and monitored by samples taken periodically.
Amendment 6 9.1-4 April 1984
e INSERT 3 Fucl stored in the spent fuel pool is segregated into two arcas (Region 1
and Region 2).
Spent fuci pool Region 1 will provido for storage of fuel with enrichment between 3.6 and 4.Sfw/o in a 3 of 4
cell array administratively controlled.
The non-fueled cells will provide adequate spacing to prevent criticality..
Criticality in Region 2
is prevented by limiting storage to fuel assemblics with burnup dependent enrichment limitations provided in the technical specifications.
The soluble boron in the pool water provides availabic negative reactivity to maintain K gg less than or equal e
to 0.95 for postulated accidents that would atfoct an increase in reactivity.
These limitations satisfy the design basis for preventing criticality outside the reactor
- where, including uncertaintics, there is a 95% probability at a 95% confidence level that the Kogg of tho' fuel assembly array will be less.than 0.95.
8 D
9 9
9 9
0 9
l e
0 e
3 6
e
l l
BVPS-2 FSAR Decay heat from fuel elements is removed by the fuel pool cooling system, as described in Section 9.1.3.
[
ventilation in the ' fuel building is designed to maintain a negative pressure and is described in Section 9.4.2.
t t
9.1.2.3 Safety Evaluation In accordance with Regulatory Guide 1.13, the storage and handling of fuel in the fuel building is designed to protect the fuel,-limit
[
[
l r
f I
I i
?
i i
t k
r i
k i
i l
{
l i
Amendment 6 9.1-4a April 1964 l
BVPS-2 FSAR potential offsite exposures, and prevent loss of water from the fuel pool which may uncover the fuel.
The spent fuel pool, spent fuel pool liner, and all supporting structures are designed for $$E seismic loads as described in Sections 3.8.4 and 3.2.1.2.
The BVPS-2 spent fuel pool structure and the spent fuel racks are classified,
- designed, and constructed as Seismic Category I items. The spent fuel pool liner and refueling cavity liner are classified, designed, and constructed as Seismic Category II items. The effects of tornadoes, hurricanes, and floods are described in Sections 3.3.1, 3.3.2, and 3.4.1.
The capability of these components and structures to withstand the effects of external missiles, pipe whip, and jet impingement forces are described in l
Sections 3.5.1.1, 3.6.1, and 3.6.2.
The spent fuel pool is designed such that the water level in the pool cannot be decreased below the top of the fuel stored in the spent fuel racks.
The fuel transfer gates do not extend below the top of the spent fuel assemblies, and all piping and piping penetrations of the spent fuel pool terminate no lower than 10 feet above the top of the fuel stored in the racks.
The fuel pool is lined with stainless steel and is equipped with a leak chase system and tell-tale drain connections which drain to a l
tell-tale drain tank located in the fuel building.
In the event of a lass of fuel pool cooling and normal makeup water supply, a supply of water is provided from the seismic Category I service water system, as described in Section 9.1.3.
Radiation levels are kept at a minimum (Chapter 12) and optical clarity is maintained by the spent fuel pool cleanup system, as described in Section 9.1.*.
i l
The release of radioactive material is prevented by the design of the fuel building ventilation system which maintains a negative pressure on the building and by the supplementary leak collection, as described in Sections 9.4.2 and 6.5.1.
The ASME !!! portions of the fuel posl cooling system and the ASME !!! portions of other systems important to safety of the spent fuel stored in the spent fuel storage facility undergo periodic inservice inspection and testing, as described in Sections 3.9.6 and 6.6.
Spent fuel assemblies are stored vertically in 17 free-standing high density storage racks.
The racks utilize a neutron absorbing material (boron carbide in nonmetallic binders) in vented storage l
corepartments to prevent the buildup of gases, and have a minimum center-to-center spacing of 10 7/16 inches to maintain the spent fuel in a suberitical condition. With spent fuel of a maximum enrichttent of-h+-percent by weight UO,, the fuel pool filled with pure water at l
Y.I Amendment 10 9.1-5 May 1985 i
J
BVPS-2 FSAR d
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-H'F, the fuel stored in the worst feasible geometric configuratien, and with the worst case seismic deflection, the effective multiplication factor, K,gg,blywill be less than 0.95.
For the condition of a fuel assem dropping on a storage rack and/or betweentheracksbadthefuelpoolline&r,theK,ggayysrss ~ lely itNp% daico/vsl$rw*
is also verified as less than 0.95, e To be f reseece The continued presence of neutron absorbing material is ensured by a poison surveillance program. This program provides samples which are exposed simultaneously to pool water and gamma radiation. Samples are exposed inside sample holders, which can be moved after each refueling to allow irradiation by fresh spent fuel.
Detailed criticality analyses are performed to demonstrate that the spent fuel racks are substantially suberitical (K,gg <0.95) for all credible combinations of the normal and abnormal fuel assembly / rack i
configurations. The criticality analyses are performed using the Monte Carlo Code, Keno IV.
AU d*tyy The Monte Carlo code, Keno IV, is a multi-group neutron transport code utilizing group crosssections which calculates K
gg life-time andgeneration-timeleakagefluxes,andfissiondensitIes.,
Extenrive bench marking calculations are performed on criticality experiments involving storage of simulated, fresh pressurized water i
reactor fuel assemblies in poison storage cells.
klllllC6 he normal and abnormal configurations considered in the analys Q)Tll Ar*
N*'**
TilSE M 4 1.
Centra esitioning of fuel assemblies within storage cells
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of normal
.ensions at normal temperature i
2.
Eccentric positt ng of adjacent fuel assemblies within the 4
storage cells, 3.
Variations in cell wall ickness, storage cell center-to-l center pitch, and poison 3en tration and thickness, as l
permitted by fabricationAeleran
, and
/
4.
Variation in fue parameters includin enrichment and fuel rod pitch.
Abnormal 7
1.
Bulk, pool temperature variations from 32*F to 260' ith fujr fier reduction in water density to determine the effech
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Amendment 10 9.1-6 May 1985 i
.TM.1EQT y Cedjurafime ceaslde<</ i; fle a.sr /ys e,s a <<e ';
The maximum K.ve under normal conditions arises from consideration of me-chanical and material thickness tolerances resulting from the manufacturing procass in addition to asymmetric positioning of fuel assemblies within the storage cells. Studies of asymmetric positioning of fuel assemblies within tr.o storage cells has shown that symmetrically placed fuel assemblies yield con-servative results in rack K.ve. The sheet metal tolerances are considered along with construction tolerances related to the cell 1.D., and cell center-to-center spacing.
For the Region 1 racks this resulted in a riduction of the nominal 1.106" water gaps to their minimum values. Thus, the "worst case" KENO model of the Region 1 storage racks contains minimum water gaps of 1.007" with symmetrically placed ruel assemblies.
Most accident conditions will not result in an increaue in Keve of the rack. Ex-amples are tne loss of cooling systems (reactivity decreases with decreasir.g water density) and dropping a fuel assembly on top of the rack (the rack structure pertinent for criticality is not excessively deformed and the dropped assembly has more than twelve inches of water separating it from the active fuel helght of stored assemblies which precludes interaction).
However, accidents can be postulated which would increase reactivity (i.e., or dropping a fuel assembly between the rack and pool wall). For these accident conditions, the dcuble contingency principle of ANSI N16.1-1975 is applied. This states that one is not required to assume two unlikely, independent, concurrent events to ensure protection against a criticality accident. Thus, for accident conditions, the presence of soluble boron in the storage pool water can be assumed as a realistic initial condition since not assuming its presence would be a second unlikely event.
The presence of approximately 1000 ppm boren in the pool water will decrease reactivity by about 15 percent AK. Thus, for postulated accidents, should tnere te a reactivity increase, K.*e would be less than or equal to 0.95 due to the effect of the dissolved boron.
j
BVPS-2 FSAR torage cells at minimum center-to-center spacing resulC ic vibration / displacement, s
3.
Fuel ha AndliEiiftt$on in which a fuel assembly is placed p t to a fully loade 1
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Amen & ment 10 9,1-6a May 1985.
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BVPS-2 FSAR h
Fuel handling incident in which a fuel assembly is dro d
om a height of 2 feet above the top of the rac I
5.
Fuel ha ing incident in whic el assembly is' lying across the t of the rack All analyses are me suming the fuel stored to be non-i irradiated wi weight percen g earichment in a pure water enviro An additional analyst a performed to determine the cality effects of missing poison plat s
The high density spent fuel storage racks have been designed to meet the requirements for Seismic Category I structures.
Detailed t
structural and seismic analyses of the storage racks have been-performed to verify the adequacy of the design to withstand the j
loadings encountered during normal operation. OBE, and SSE.
t As described in sections 9.1.4 and 9.1.5. the moveable platform with hoists is the only crane operating over the spent fuel and is described in Section 9.1.4.
The spent fuel cask trolley is described in Section 9.1.5. along with a description of the paths of I
travel and interlocks to preclude the dropping of heavy objects on i
stored spent fuel.
cooling of spent fuel stored in the spent fuel storage racks is accomplished by the safety-related seismic category I fuel pool
['
ccoling system described in Section 9.1.3.
The adequacy of natu: al circulation flow to cool the spent fuel assemblies was established by a
thermal hydraulic
- analysis, which concluded that natural i
circulation in the spent fuel pool is adequate to prevent local boiling.
The design of the spent fuel racks is such that it is not possible to I
insert a spent fuel element in other than a design location, for example, between storage locations or between racks.
l Damage to the spent fuel assemblies and the spent fuel racks by excessive uplift forces exerted by the spent fuel hoist during fuel handling are prevented by the hoist's load cell.
l All materials used in construction are compatible with the spent fuel l
pool environment. All materials are corrosion resistant stainless steel, with the exceptions sf the neutron absorbing material, gate seals, and fuel pool lights, and will not contaminate the fuel l
assemblies or pool environment.
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Amendment 3 9.1-7 October 1983
ATTACHMENT E
Criticality Analysis (provided for background information) l r
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