ML20207S002
| ML20207S002 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 03/09/1987 |
| From: | DUQUESNE LIGHT CO. |
| To: | |
| Shared Package | |
| ML20207R864 | List: |
| References | |
| NUDOCS 8703180406 | |
| Download: ML20207S002 (21) | |
Text
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ATTACHMENT A
Revise the technical specifications as follows:
Remove Pages Insert Pages 3/4 9-14 3/4 9-15 4
B 3/4 9-3 B 3/4 9-3 5-4 5-4 5-5 5-5 i
- l r
i B703180406 870309 i
PDR ADOCK 05000334 l.
p PDR
3/4.9.14 FUEL STORAGE - SPENT FUEL STORAGE POOL s 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:
1) 4.5 w/o stored in a 2 of 4 checkerboard configuration; or 2) 4.0 w/o stored in a 3 of 4 checkerboard configuration; j
and 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 the i
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 boron concentration is less than 1050 ppm, until such time as
[
the boron concentration is increased to 1050 ppm or j
greater.
c.
The provisions of Specifications 3.0.3 and 3.0.4 are j
not applicable.
I SURVEILLANCE REQUIREMENTS i
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 > 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.
BEAVER VALLEY - UNIT 1 3/4 9-14 PROPOSED WORDING
9 Table 3.9-1 g
BEAVER VALLEY FUEL ASSEMBLY MINIMUM BURNUP VS. INITIAL U235 ENRICHMENT FOR STORAGE IN REGION 2 SPENT-FUEL RACKS Initial U235 Assembly Discharge Enrichment Burnup (GWD/MTU) 3.1 0
3.3 1.6 3.5 3.0 3.7 4.4 3.9 5.8 4.1 7.2 4.3 8.5 4.5 9.7 NOTE:
Linear interpolation yields conservative results.
l l
l f
l l
BEAVER VALLEY - UNIT 1 3/4 9-15 l
PROPOSED WORDING t
l i
l
REFUELING OPERATIONS _
gBASES 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 analysis.
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 absorber 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.
3/4.9.14 FUEL STORAGE - SPENT FUEL 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,dencegg 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 baron concentration is maintained at 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 k gg is
<0.95; and (2) the boron and it was determined that e
concentration meets the 1050 ppm limit.
i BEAVER VALLEY - UNIT 1 B 3/4 9-3 PROPOSED WORDING
~
e DESIGN FEATURES DESIGN PRESSURE AND TEMPERATURE 5.2.2 The reactor' containment building is designed and shall be maintained for a
maximum internal pressure of 45 psig and a
temperature of 280*F.
PENETRATIONS 5.2.3 Penetrations through the reactor containment building are designed and shall be maintained in accordance with the original design provisions contained in Section 5.2.4 of the FSAR with allowance for normal degradation pursuant to the applicable Surveillance Requirements.
5.3 REACTOR CORE FUEL ASSEMBLIES 5.'.1 The reactor core shall contain 157 fuel assemblies with each fuel assembly containing 264 fuel rods clad with zircaloy-4, except for fuel assemblies which may be reconstituted to replace fuel rods with non-fueled rods (e.g., zircaloy or stainless steel).
Each fuel rod shall have a
nominal active fuel length of 144 inches.
Reload fuel shall be similar in physical design to the initial core loading and shall have a maximum enrichment of 4.5 weight percent U-235.
l CONTROL ROD 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 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 rods shall be clad with stainless steel tubing.
BEAVER VALLEY - UNIT 1 5-4 PROPOSED WORDING
e DESIGN FEATURES 4
_..v 5.4 REACTOR COOLANT SYSTEM y
?'
DESIGN PRESSURE AND TEMPERATURE c
c f
5.4.1 The reactor coolant system is designed and sha'll be maintained:
a.
In accordance with the code requirements specified in Section-4.2 of.
the
- FSAR, with allowance for normal
/1 degradation pursuant to the applicable-Surveillance Requirements, b.
For a pressure of 2485 psig, and c.
For a
temperature of 650*F, except for,ethe pressurizer which is 680*F.
z.
VOLUME 5.4.2 The total water and steam volume of the reactor coolant system is 9370 cubic feet at a nominal T of 525*F.
avg 5.5 EMERGENCY CORE COOLING SYSTEMS ri 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.,
5.6 FUEL STORAGE i
CRITICALITY 5.6.1 The spent fuel storage racks are designed with a minimum of l
12.0625 inch center-to-center distance between fuel assemblies pl.ced,
a I
in the storage racks.
The fuel will be stored in accordance with the provisions described in UFSAR Sections 3.3 and 9.12 to ensure a kefg equivalent to
<0.95 with the storage pool filled with unborated water.
DRAINAGE i
5.6.2 The spent fuel storage pool is designed and shall be i
maintained to prevent inadvertent draining of the pool below l
elevation 750' - 10".
l BEAVER VALLEY - UNIT 1 5-5 PROPOSED WORDING l
. lN'O ATTACHMENT B
'l; JL No Significant Hazard Consideration s
?'
l Pbboosed Clhange Request No.
132 mends the Beaver Valley Power i
Station, iunit' No.
1, Technical Specifications to limit the maximum fuel enrichsent to 4.5 w/o.
I t:
Description of amendment request:
Change Request No. 132 revises p'
Design Fcature Section 5.3.1, 5.6.1 and incorporates section 3.9.14 and associated Bases to setforth fuel assembly U-235 enrichment limitations; on placement of fuel in the spent fuel pool.
These changes rbaie ba' ed on an evaluation performed by Westinghouse s
4
" Criticality Analysis of Beaver Valley 1
Fresh and Spent Fuel i
Racks".
The results of this evaluation provide justification for:
1.
,New fuel storage rack enrichment limit of 4.5-w/o, 2.
Two spent fuel storage rack enrichment limits, where Region 1
is limited to 4.5 w/o for a
two out of four cell checkerboard ' storage pattern, or 4.0 w/o for a three out of four cell checkerboard storage pattern and Region 2 is limited %o3.1w/ousingeverystoragecell.
g J
Fuel. assembly criticality in fuel storage racks is prevented by
,4--design,.to limit fuel assembly interaction by fixing the minimum 7'sseparation between assemblies.
The design basis for preventing criticality outside the reactor is that, including uncertainties, there is a/ a 95%
probability at a
95%
confidence level that the effective multiplication factor (Keff) of the fuel assembly array will be less than 0.95 as recommended in ANSI 57.2-1983 and ANSI 57_.3-1983.
s F'or accident ' conditions where reactivity is postulated to a-
- increase, the double contingency principle of ANSI N16.1-1975 is applied.
.This states that it is not required to assume two unlikely, independent,'
concurrent events to ensure protection against a
,V - criticality-accident.
Therefore, the accident analysis assumes the presence of soluble boron in the storage pool water as a realistic
' initial condition since not assuming its presence would be a second 4 ll 1unlikely event.
- The 1050 ppm soluble baron in the pool water will provide the negative reactivity required for postulated accidents
~'
1 affecting a
reactivity
- increase, to maintain K gg less than or e
equal,to 0.95.
Kg including uncertainties at the 95/95 The maximum probabilityb:onfidence kevel is presented for the limiting cases:
e
,3 Case Keff 6
"1. t Spent Fuel Rack Region 1, 4.5 w/o
.9049 2 of 4 cell storage.
'N 2.'. Spent Fuel Rack Region 1, 4.0 w/o
.9466 3 of 4 cell storage, e
1 3.-
Spent Fuel Rack Region 2, 3.1 w/o
.9462 all cell storage.
\\
AT'TACHMENT B
y No Significant Hazard Consideration t
Proposed ' Change Request _ No.
132' mends the Beaver Valley Power
- Station, Unit No.
1 Technical. Specifications to limit the maximum fuel enrichment to 4.5 w/o.
Description of amendment request:
Change Request No. 132 revises Design Feature Section 5.3.1, 5.6.1 and incorporates Section 3.9.14
'and associated' Bases Eto setforth fuel assembly U-235 enrichment limitations on placement of fuel in the spent fuel pool.
These changes are based on an evaluation performed by Westinghouse
" Criticality Analysis of Beaver Valley 1
Fresh and Spent Fuel Racks"..The results of this evaluation provi,de justification for:
1.
New fuel storage rack enrichment limit of 4.5 w/o, 2.
Two spent fuel storage rack enrichment limits, where Region 1
is limited to 4.5 w/o for a
two out of four cell checkerboard storage pattern, or 4.0 w/o for a three out of four cell checkerboard storage pattern and Region 2 is limited to 3.1 w/o using every storage cell.
Fuel assembly criticality in fuel storage racks is prevented by
(
design to limit fuel assembly interaction by fixing the minimum
(
separation ~ between assemblies.
The design basis for preventing
~
criticality outside the reactor is that,1 including uncertainties, there is a
95%
probability at.a 95%' confidence level that the (K gg) of the fuel assembly array effective multiplication factor e
will-be less than 0.95 as recommended in ANSI 57.2-1983 and ANSI 57.3-1983.
For accident conditions where reactivity is postulated to
- increase, the double contingency principle of ANSI N16.1-1975 is applied.
This states that it is not required to assume two unlikely, independent,
. concurrent events to ensure protection against a
criticality-accident.
Therefore, the accident analysis assumes the presence of soluble baron in the storage pool water as a realistic
. initial condition since not assuming its' presence would be a second unlikely event.
The 1050 ppm soluble boron in the pool water will provide the negative reactivity required for postulated accidents affecting a
reactivity
- increase, to maintain K gg less than or e
equal to 0.95.
K gg including uncertainties at the 95/95 The maximum e
probability / confidence level is presented for the limiting cases:
K gg Case e
1.
Spent Fuel Rack Region 1, 4.5 w/o
.9049 2 of 4 cell storage.
2.
Spent Fuel Rack Region 1, 4.0 w/o
.9466 3 of 4 cell storage.
3.
Spent Fuel Rack Region 2, 3.1 w/o
.9462 all cell storage.
__--i-
,Atttchment B Paga. 2 Case K gg e
4.
Fresh Fuel Racks, 4.5 w/o
.9160 moderatiog - full density 1.0 gm/cm 5.
Fresh Fuel Racks, 4.5 w/o
.9407 moderation - optimum low density 3
0.045 gm/cm,
K gg for each of the above limiting cases is less Since the e
than 0.95 including uncertainties at the 95/95 probability / confidence
- level, the acceptance criteria for criticality is met under all conditions.
Spent fuel rack region 1
with 4.5 w/o U-235 in 2 of 4 cell storage and 4.0 w/o U-235 in 3
of 4
cell storage will be the technical specification limiting restrictions.
Spent fuel rack region 2 with the burnup dependent restrictions provided in Table 3.9-1 and all cell storage is the limiting criteria for region 2, therefore, this will also be a technical specification restriction.
Since the maximum enrichment restriction for the spent fuel pool is 4.5 w/o and the new fuel racks are also analyzed for 4.5 w/o, no technical specification restrictions are required on the new fuel racks.
Design Feature section 5.6.1 was revised to reference the applicable UFSAR sections for fuel storage.
These UFSAR sections are being revised to reflect the new criticality analysis which includes a
description of the uncertainties applied, therefore, the sentence describing the uncertainties is not required and has been deleted.
Based on the criteria for defining no significant hazards consideration setforth in 10 CFR 50.92(c),
plant operation in accordance with the proposed amendment would not:
(1) involve a
significant increase in the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated because:
The criticality analysis acceptance criteria (Keff 0.95) is consistent with that stated in UFSAR Section 9.12 Fuel Handling System and UFSAR Section 3.3.1.5 Shutdown Margin.
Attachment C provides a
revision to UFSAR Section 9.12 to describe the segregation of the spent fuel pool into regions 1 and 2 and how the region 1
administrative constraints ensure that for 4.5 w/o fuel the 2
of 4 cell array and for 4.0 w/o fuel the 3 of 4 cell array is maintained.
In addition to the administrative constraints available to maintain the required checkerboard arrays in Region 1, the minimum boron concentration will provide an additional safety margin to ensure criticality will not be achieved.
Even if new fuel assemblies were not stored in the specified checkerboard
- arrays, the minimum boron concentration would preclude criticality.
Attachment D provides a revision to
Attechment B PIga 3 UFSAR' Section 3.3.2.7 to incorporate the new criticality analysis.
This is-provided as background information for this technical specification change and will be included in the'next UFSAR update.
4 Fuel assembly decay heat production is a function of core power
- level, and since the core power level remains. unchanged, the
~ decay heat load on the spent fuel pool cooling system will not be significantly impacted by the proposed enrichment limits.
The proposed changes will not have a significant impact on safety or on the intended operation of the spent fuel storage pool.
The Table 3.9-1 criteria provides assurance that fuel assemblies are qualified for storage in region 2 to ensure Kefg will be < 0.95 at the 95/95 confidence level.
- Thus, no adverse safety considerations are introduced by this proposed change to the technical specifications.
(2) Create the probability for an accident or malfunction of a
.different type than previously evaluated because:
The proposed changes are bounded by UFSAR Section 14.2.1 Fuel Handling Accident and the activities in the fuel rod gap presented in Appendix 14B which uses a
conservative value of 650 days at a full power value of 2766 MWt to determine fission product inventories and calculate resultant doses.
It is not required to assume two
- unlikely, independent, concurrent events to ensure protection against a criticality accident in accordance with the double contingency principle of ANSI N16.1-1975.
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 that criticality would be precluded.
The capability of the reactor core to handle up'to 4.5 w/o reload fuel will be demonstrated in the cycle-specific reload safety evaluations (RSE) which are performed prior.to fuel loading (the RSE considers 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) Involve a
significant reduction in the margin of safety because:
Specification 3.9.14 and associated bases provide for the administrative controls required to assure that fuel assemblies with the potential to form a critical array are segregated such will be less
- that, the effective multiplication
- factor, K9ff than 0.95.
Criticality will be prevented in region 1 by limiting
..Attcchment B Pcge.4 fuel assembly interaction and maintaining the minimum boron concentration.
Fuel assembly placement in region 1 will be administratively controlled by storage of fuel with an enrichment between 4.0 and 4.5 w/o in a 2 of 4 cell array and fuel with an enrichment between 3.1 and 4.0 w/o in a 3 of 4 cell array.
Where region 1
is adjacent to region 2 the arrangement of the 2 of 4
cell array (See Figure 1 example) and the 3 of 4 cell array (See Figure 2 example) will be maintained to limit fuel assembly interaction.
This satisfies the design basis for preventing
+
criticality outside the reactor where, including uncertainties, there is a
95%
probability at a 95% confidence level that K gg e
of-the fuel assembly array will be less than 0.95 in accordance with ANSI 57.2 1983 and ANSI 57.3 - 1983.
Therefore, since.
K be maintained less than 0.95 and this is consistent wIkb will the current design basis, the margin of safety will not be reduced.
CONCLUSION The design and use of the spent fuel pool will be changed to reflect the
" Criticality Analysis of Beaver Valley 1 Fresh and Spent
-Fuel Racks".
The criticality analysis supports the storage of fuel enriched up to 4.5 w/o U-235.
This will facilitate longer fuel
- cycles, higher nuclear capacity factors and lower plant power-4 generation costs.
Spent fuel pool region 1 will provide for storage of fuel with enrichments up to 4.5 w/o in an administratively controlled 2 of 4 cell array and up to 4.0 w/o in an administratively controlled 3 of 4 cell array.
Region 2
will provide for storage of fuel assemblies with the burnup dependent enrichment limitations provided in Table 3.9-1.
The proposed 3.1 w/o limit is more restrictive than the current limit of 3.3 w/o due to the difference in analysis methodologies used.
The new criticality methodology is based on a three dimensional Monte Carlo theory program KENO IV which is designed for reactivity calculations.
Keff will be maintained less than 0.95 consistent with the current UFS M design basis and with the region 1
checkboard
- array, the segregation of fuel assemblies into regions 1
and 2 and the proposed technical specification changes no adverse safety considerations are introduced.
Therefore, the proposed changes will not increase the likelihood of a malfunction of safety related equipment, increase the consequences of an accident previously
- analyzed, nor create the possibility of a malfunction different than previously analyzed.
Based on the above, it is proposed to characterize the change as involving no significant hazard.
4
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YYkYV 0 BVPS-1-UPDATED FSAR Rev. 0 (1/82) 9.12 FUEL HANDLING SYSTEM The fuel handling system provides a safe, effective means of transporting and handling fuel from the time it reaches the station in an unirradiated condition until it leaves the station after post-irradiation cooling.
The system is designed to minimize the possibility of mishandling or of malfunctions that could cause fuel damage and potential fission product release.
The fuel handling system consists basically of:
1.
The refueling cavity and fuel transfer canal which are flooded only during station shutdown for refueling i
2.
The spent fuel pool, which is kept full of water and is always accessible to operating personnel 3.
The transfer tube, which connects the above two areas, fuel transfer
- system, RCC changing
- fixture, i
manipulator
- crane, new fuel elevator and other equipment used to manipulate the fuel assemblies.
9.12.1.
Design Bases 9.12.1.1 Prevention of Fuel Storage Criticality The new and spent fuel storage racks are designed so that it is impossible to insert assemblies in other than the prescribed locations within the racks.
The new fuel storage rack accommodates one-third of a core plus 17 spara assemblies.
The spent fuel storage pool accommodates five cores plus 48 spare assemblies, plus the required spent fuel shipping cask area.
Borated water is used to fill the spent fuel pool at a
concentration to match that used in the refueling cavity and fuel transfer canal during refueling operations.
The spent fuel is stored in a vertical array with sufficient center-to-center distance between assemblies to ensure k
even if unborated water is used to fill the pool. In,*e,I < 0.95 4A T
9.12.1.2 Fuel Storage Decay Heat The fuel pool cooling system which removes the spent fuel decay heat is described in detail in Section 9.5.
i 9.12.1.3 Fuel Building Radiation Shielding l
The shielding provided in the fuel building and the resulting maximum doses outside of this building are discussed in Section 11.3.
9.12-1
BVPS-1-UPDATED FSAR Rev. 0 (1/82) 1 Space is provided in the pool for the spent fuel storage rack, one shipping cask, the new fuel elevator, and fuel transfer and upending equipment.
The spent fuel pool is divided'into three areas:
1.
The fuel transfer mechanism area 2.
The spent fuel storage area 3.
The spent fuel cask laydown area.
The fuel transfer mechanism area can be dewatered without emptying the entire pool.
A concrete wall separates the fuel transfer mechanism area from the spent fuel storage area.
The arrangement of the wall within the pool can be seen in Figure 1.2-10.
The top of the wall is level with the top of the pool.
By insertion of a gate dam at the open end of the transfer mechanism area, this area can be isolated from the spent fuel storage area.
4 A concrete wall also separates the spent fuel cask laydown area from the spent fuel storage area.
The only penetration of the wall is a 24 inch slot to allow for underwater passage of spent fuel elements into the laydown area.
The slot is entirely above the storage level of the spent fuel elements in the fuel pool.
The stainless steel storage racks, bolted to the floor and braced to the walls of the storage area, are sized to hold 785 spent fuel assemblies plus 48 spares.
The spent fuel assemblies are placed in vertical cells within the rack, continuously grouped in parallel rows about.12 inches on centers, in both directions.
The rack is so arranged that the spacing between fuel elements cannot be less than that prescribed.
The necessary spacing between assemblies is ensured to prevent criticality (Kag= 0.95) even if the pool were filled with unborated water.
TKd control rod clusters and burnable poison rods are stored in the fuel l
element assemblies.
>Inser-/ 0 A movable platform with two 10-ton electric hoists runs over the spent fuel pool.
The fuel assemblies are moved with the long fuel handling tool suspended from one hoist.
Within the spent i
fuel pool, the control rods are not handled separately from the fuel assemblies.
A 125-ton overhead crane and trolley is provided for moving the spent fuel shipping cask into or out of the pool.
This crane passes only over the shipping cask loading area which is isolated from the spent fuel pool.
The fuel pool cooling and purification systems are described in Section 9.5.
New Fuel Storage New fuel assemblies are stored dry in a steel and concrete structure within the fuel building.
Both the elevation and type 9.12-6
~
4
' INSERT A Limitations on the placement of fuel in the spent fuel pool are discussed in Sections 3.3.2.7 and 9.12.2.2.
INSERT B Fuel stored in the spent fuel pool is segregated into two areas (Region 1
and Region 2).
Spent fuel pool Region 1 will provide for storage of fuel with enrichment between 4.0 and 4.5 w/o in a 2 of 4 cell array and fuel with enrichment between 3.1 and 4.0 w/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 assemblies with burnup dependent enrichment limitations provided in the technical specifications.
The soluble baron in the pool water provides-available negative reactivity to maintain K gg less than or equal e
to 0.95 for postulated accidents that would affect an increase in reactivity.
These limitations satisfy the design basis for preventing criticality outside the reactor
- where, including uncertainties, there is a 95% probability at a 95% confidence level that the K gg of the fuel assembly array will be less than 0.95.
e
A7FM/MBW h BVPS-1-UPDATED FSAR Rav. 2 (1/84)
(and thus reactivity importance) is assumed to be skewed to the bottom of the core.
The result of these calculations is shown on Figure 3.3-39.
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 subcritical at hot shutdown if all Rod Cluster Assemblies are tripped, but assuming that the highest worth assembly remains fully withdrawn and no changes in xenon or boron concentration 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.
The values given in Table 3.3-3 show that the available reactivity in withdrawn Rod Cluster Control Assemblies provides the design bases minimum shutdown margin allowing for the highest worth cluster to be at its fully withdrawn position.
An allowance for uncertainty in the calculated worth of N-1 rods is made before determination of the shutdcwn margin.
3.3.2.7 Criticality of Fuel Assemblies
$6C Q/[d ded I6 h as.r.r.za
-iticality of fuel assemblies outside of the reactor pr luded by adequate design of fuel transfer and fuel sto ge faci ' ties and by administrative control procedures.
This s
section identifies those criteria important to criticalit', safety
)
analyses.
New fuel is ~ nerally stored in fuel storage fac4 '. ties with nc water present b
- which are designed so as to pnvent accidental criticality even l' unborated water is present In the analysis for t.
storage faciliti the fuel assemblies are assumed to be in th most reactiv ccndition, namely fresh or undepleted and with control _ods or removable neutron i
absorbers present.
Assemb11 s can not be closer together than the design separation provided % the storage facility except in special cases such as in fuel oing containers where analyses are carried out to establic. the 'cceptability of the design.
The mechanical integrity o the fuel assembly is assumed and no credit it taken for neu on absorption roperties of the storage facility unless speci".cally included in the design.
Fcr full flooding with unbor ed water, the fuel as 'mbly spacing of the facility provides ssentially full nuclear iso' tion and K,,,
for the array is n greater than K for the sin, e =cs: reactive fuel assembly The criterion for*,,t:ll ficoding is,,, < 0.90.
T The fuel ssembly (17 x 17 fuel rods) of standard desi.n and 3.5 w/o en".ched uranium cxide, without a control rod or - urnable pois rods, fully ficoded and reflected with cold clean
-ter, ha a K of about 0.85.
Two such fuel a s se:,blie s spaced ^ne j
..ch ap8 E with parallel axes 9.5 inches apart have a Keff 3.3-31
BVPS-1-UPDATED FSAR Rev. 2 (1/84)
Mm+
0.99.
Three such fuel assemblies spaced one ih
]
with parallel am.;cM d be m ercritical.
J ary fuel assemblies of this dEE-1gm-would 3.3.2.8 Stability 3.3.2.8.1 Introduction The stability of the PNR cores against xenon-induced spatial oscillations and the control of such transients are discussed extensively in References 4,
8, 9 and 10.
A summary of these reports is given in the following discussion and the design bases are given in Section 4.3.1.6.
In a large reactor core, xenon-induced oscillations can take place with no corresponding change in the total power of the core.
The oscillation may be caused by a power shift in the core which occurs rapidly by comparison with the xenon-iodine time constants.
Such a power shift occurs in the axial direction when a plant load change is made by control rod motion and results in a
change in the moderator density and fuel temperature distributions.
Such a power shift could occur in the diametral plane of the core as a result of abnormal control action.
For a discussion of the methods for detecting azimuthal or diame-tral induced power oscillations, see the discussions of measure-ments in the X-Y plane in Section 3.3.2.8.4 and the discussion of instrumentation applications in Section 3.4.5.
Cperation w:.th two locps dces~ not have an effect en the xenon distribution since there is a plenum at the core inlet to mix the flew from all operating loops and distribute the ficw across the core inlet.
Since the core is open (uncanned assemblies) and provides for free crcss ficw, no gross asymmetry in flow, which could affect the xenon distribution, can exist for 2 or 3 loop operation.
Due to the negative power coefficient of reactivity, PWR cores are. inherently stable to oscillations in total pcwer.
Prctection against total power instabilities is provided by the Control and Prctection System as described in Section 7.7.
- Hence, the discussion on the core stability will be limited here to xenen-induced spatial oscillations.
3.3.2.8.2 Stability Index Pcwer distributions, either in the axial direction or in the X-Y plane, can undergo oscillaticns due to perturbations intrcduced in the equilibrium distributions without changing the total core power.
The xenon-induced oscillations are essentially limited tc the first flux cvertones in the current PWR's, and the stability of the core against xenon-induced oscillations can be determined l
l l
3.3-32 L
REPLACEMENT PAGES
=
Replace BV-1 UFSAR Section 3.3.2.7 with the following:
Criticality of fuel assemblies outside the reactor is precluded by adequate
. design' of fuel transfer, shipping and storage facilities, 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.
The design basis for preventing criticality outside the reactor is that, considering possible variations, there is a 95 percent probability at a 95 percent confidence level that the effective multiplication factor (k,ff) of the fuel assembly array will be less than 0.95 as recommended in ANSI 57.2-1983, ANSI 57.3-1983 and Reference 33.
The following are the conditions that are assumed in meeting this design basis:
r 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.
2.
For flooded conditions, the moderator is pure water at the temperature within the design limits which yields the largest reactivity.
3.
The array is either infinite in lateral extent or is surrounded by a conservatively chosen reflector, whichever is appropriate for the design.
I 4.
Mechanical uncertainties are treated by either using " worst case" conditions or by performing sensitivity studies and obtaining appropriate uncertainties.
5.
No credit is taken for any spacer grids or spacer sleeves.
1
REPLACEMENT PAGES (cont.)
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.
The ' design method which ensures the criticality safety of fuel assemblies outside the reactor uses the AMPX system of codes (Reference 34, 35) for cross-section generation and KEN 0 IV (Reference 36) for reactivity determination.
The 227 energy group cross-section library (Reference 34), that is the common starting point for all cross-sections used for the benchmarks and the storage racks has been generated from ENDF/B-V data. The NITAWL program (Reference 35) includes in this library the self-shielded resonance cross-sections that are appropriate for each particular geometry. The Nordheim Integral Treatment is used. Energy and spatial weighting of cross-sections is performed by the XSDRNPM program (Reference 35), which is a one-dimensional S transport theory N
code. These multi group cross-section sets are then used as input to KENO IV (Reference 37), which is a three-dimensional Monte Carlo theory program designed for reactivity calculations.
A set of 33 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 (Reference 37) to dry harder spectrum uranium metal cylinder arrays with various interspersed materials (Reference 38) that demonstrate the wide range of applicability of the method.
Table 3.3-12 summarizes these experiments.
The average k of the benchmarks is 0.992.
The standard deviation of the bias gff value is 0.0008Ak. The 95/95 one sided tolerance limit factor for 33 values is 2.19.
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 0.0018ak.
REPLACEMENT PAGES (cont.)
The following equation is used to develop the maximum k,7f.
eff
- worst method + [(ks)2
- ( S) method 3
+ O
~
worst Where:
k
= w rst case KEN 0 k that includes material tolerances, and -
worst eff mechanical tolerances which can result in spacings between assemblies less than nominal.
B
= method bias determined from benchmark critical comparisons.
method ks
= 95/95 uncertainty in the worst case KENO k worst eff' ksmethod = 95/95 uncertainty in the method bias.
I The criticality acceptance criteria is met when the effective multiplication factor (kgff) including uncertainties at a 95/95 probability / confidence level is less than 0.95.
These methods conform with ANSI N18.2-1973, Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants, Section 5.7, Fuel Handling System; ANSI 57.2-1983, Design Objectives for LWR Spent Fuel Storage Facilities at Nuclear Power Stations, Section 6.4.2; ANSI N16.9-1975, Validation of Calculational Methods for Nuclear Criticality Safety; NRC Standard Review Plan, Section 9.1.2, Spent Fuel Storage; and the NRC guidance, NRC Position for Review and Acceptance of Spent Fuel Storage and Handling Applications; ANSI 57.3-1983, Design Requirements for New Fuel Storage Facilities at Light Water Reactor Plants.
4 4
r 4
-~----.-,--r--.--,,,r.,,.mm------.--.---m n,ew,---~,w--,,,
m-,_.m,,-4w~~,
.,.---.v.
- ~
BVFS-1-UFDATED FSAR Rev. 3 (1/85) a References for Section 3.3 (Cont'd) 24 Deleted'by Revision 0.
25.
Deleted by Revision O.
26.
Deleted by Revision 0.
27.
J.
M.
- Hellman, (Ed.),
" Fuel Densification Experimental Results and Model for Reactor Application," WCAP-8219, Westinghouse Electric Corporation (October, 1973).
28.
J. M. Hellman and J.
W.
Yang, " Effects of Fuel Dentificatict.
Power' Spikes on Clad Thermal Transients," WCAP-8359, West-inghouse Electric Corporation (July, 1974).
29.
T.
Morita, et al.,
"Tcpic.11 Peport, Power Distributien Control and Load Following Procedures," WCAP-8385, Westing-house Electric Corporation (September. 1974).
30.
C.
Eiche1dinger,- "Westincheuse Letter to D.
O.
Vassate,"
HS-CE-687, Westinghcuse Electric Ccrpet3tien (July, l9*!).
31.
Camden, T.
M.,
et. al., "l'ALADCN - Wastinghouse ' led a l Computer Code",
WCAP 9485A (Proprietaryl and WCAP.94861.
(tlon-Preprietary), Decamber, lo'A.
I 32.
- Ankney, R.
D.,
"PALACON-Westinqheusa Medal Ccmputaar Ccde.
Supplement 1",
WCAP 0485 supplement. September. 1941.
33.
Nuclear Regulatory Ccmmission, Letter to All Power Reactor Licensees, from
- 8. K. Grimes April 14,1978. *0T Position for Review and Acceptance of f
Spent Fuel Storage and Handling Applications."
$/,
W. E. Ford III, et al., 'CSRL-V: Processed ENDF/B-V 227-Neutron-Group and v_
Pointwise Cross-Section Libraries for Criticality Safety Reactor and
]
Shielding Studies.* ORNL/CSD/TM-160 (June 1982).
4 JJ~
N. M. Greene, et al., 'AMPX: A Modelar Code System for Generating Coupled Multigroup Neutron-Gama Libraries from ENDF/B," ORNL/fM-3705 (March 1976).
Jg.
L. M. Petrie and N. F. Cross
- KENO IV--An Improved Monte Carlo Criticality Program." ORNL-4938 (November 1975).
I 37, M. N. Baldwin, et al., " Critical Experiments Supporting Clese Pronimity i
Water Storage of Power Reacter Fuel " BAW-1484-1, (July 1979).
II, J. T. Thomas, ' Critical Three Dimensional Arrays of U (93.2) -- Metal
.ylinders
- Nuclear Science and Engineering, Volume 52, pages 350-359 (1973).
1 l
3.3-44 I
.. -,,. -.. - - - _ _ -..-._-_- - - - - - - - - - - - -..._-.