ML20116G580
| ML20116G580 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 11/02/1992 |
| From: | DUQUESNE LIGHT CO. |
| To: | |
| Shared Package | |
| ML20116G556 | List: |
| References | |
| NUDOCS 9211120036 | |
| Download: ML20116G580 (29) | |
Text
l ATTACHMENT A Beaver Valley Power Station, Unit No. 1 Proposed Technical Specification Change No. 202 MARKED UP PAGES Revise the Technical Specification as follows:
Remove Pagg Insert Pagg 3/4 9-14 3/4 9-14 3/4 9-15 3/4 9-15 B 3/4 9-3 B 3/4 9-3 B 3/4 9-4 B 3/4 9-4 5-5 5-5 5-6 5-6 9211120036 9211r2 PDR ADOCK 05000334 P
DPR-66 REFUELING OPERATIONS
?/4.9.14 SPENT FUEL STORAGE POOL 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 restrfcted to fugl with enrichment less than or equal to)( 50 U/o I/ A3G ar/
at
-1) 4.5 v/ o-ebored-ir.
2 of--4-checkerboard conf-19uratica;
-ee-2; 4.0 v/
etered in : 2 ef 4 checkerbe:rd configuration;-
or o Ut $1balIf QMl srh y
c.
Fuel assembly storage in Region 2 restricted t 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 spent fuel pool if it is determined a fuel assembly has been placed.in the incorrect Region until such tims 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 incroased to 1050 ppm or greater.
c.
The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
pot a U g h balt aMly!Is SURVEILLANCE REQUIREMENTS 7
4.9.14.1 Prior to placing fuel or mov g fuel in the spent fuel
- pool, verify through fuel receipt record for new fuel,-ee 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 2 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 Amendment No.
Mofost b
,OpR-66 TABLE 3.9-1 BEAVER VALLEY FUEL ASSEMBLY MINIMUM BURNUP VS.-INITIAL U235 ENRICHMENT FOR STORAGE IN REGION 2 SPENT FUEL RACKS NW.9 Initial U235 Assembly ischarge Enrichmeat;_
Burnuo MTU) 2.1
- 1. 0
-O~
>[Y[
-h4-
- 2. 7
-1,4-93Tl
-b4-7,0
+re- /f 7PV
- .?
2af
+ r + p t 6 y.1
-h4-4. 0
-sve-2 7.2 40 M 4.l M 3.77/O
-4.4-
.C0
-s,4-yo ooo
-4,4-
-0.?
"OT";
Linear interpolet-len-yie de-conservative results.
No7E: The dafr a the ahore 1/r may be eith MupreM h aearly oc may be calculated by tk co,senahk ep oM, below. TA4 egnabb., prvhhs a lin ea r M fi A hyj,buenap ha h.
N NI butnuf3 l MWb/nTR = /k/M + $.1os oo wkere &
- E edcl med (6 d 5%)
BEAVER VALLEY - UNIT 1 3/4 9-15 Amendment No.
l fkOf0C[~D
'DPR-66 REF ELING 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 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 tilters 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 analysis.
The spent fuel poo'. area ventilation system is non-eafety related and only recirculates air through the fuel building.
The SLCRS portion of the ventilation system is safety-related and mainta' ins a negative pressure in the fuel building.
The SLCRS flow is normally exhausted to the atmosphere without filtering, however, the flow is diverted through the main filter banks by manual actuation or on a high radiation signal.
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 1Nds; 16.1-1975.
The value of 0.95 or less for keff which includes all uncertainties at the 95/95 probability / confidence level is the 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 cron l
concentration is maintained at 2 1050 ppm (this includes a 50 ppm conservative allowance for uncertainties) during all actions involving movement of fuel in the spent fuel pool.
6fo The Surveillance Requiremente 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 Keff is s
0.95; and (2) the boron concentration meets the 1050 ppm limit.
feph I crF The enrichment limitations for storage of fuel in :. Of array ir the spent fuel pool.is based on a nominal region average enrichment with individual fuel assembly tolerance of + or - 0.05 w/o {-235.of STO J/o l 3
BEAVER VALLEY - UNIT 1 B 3/4 9-3 Amendment No.
PROPOSEb
\\
DPR-66 REFUELING OPERATIONS BASES EUEL STORAGE - SPENT FUEL STORAGE POOL (Continued) results of the spent fuel pool criticality analysis (Au e
1986 Westinghouse STD/ Vantage 5H and OFA/ Vantage el in three of storage locations show that thure is e than 0.3%
7jggg7 j margin to the limit of 0.95 with all (3r inties included.
Based on the sensit study comple with this analysis, an increase in the maximum ed ent for fuel stored in the spent fuel storage racks f
to 4.05 w/o will increase the maximum rack keff by 1 an 0.002.
efer' with Westinghouse 17 x
17 STD/ Van SH and OFA/ Vantage 5 fu c.;& riched at 4.05 w/o stored in spent fuel racks in three of fouh' age locations and w
all of the assumptions and conservatisms pres Q in the lity analysis, the maximum rack keff will be less than O K. ~
3/4.9.15* CONTROL ROOM EMERGENCY HABITABILITY SYSTEMS The OPLRABILITY of the control room emergency habitability system ensures that the control room will remain habitable for operations personnel during and following all credible accident conditions.
The ambient air temperature is controlled to prevent exceeding the allowable equipment qualification temperature for the equipment and instrumentation in the control room.
The OPERABILITY of this system in conjunction with control room design provisions is based on limiting the radiation exposure to personnel occupying the control room to 5
rem or less whole
- body, or its equivalent.
This limitation is consistent with the requirements of General Design Criteria 19 of Appendix "A",
l i
l l
BEAVER VALLEY - UNIT 1 B 3/4 9-4 Amendment No.
PROPOSED
..-- ~-
- I'nstrt 1
-- Rp s e s 314.' 9.' L4 FUEL STORAGE - SPENT FUEL STORAGE. POOL The~ reracked ~ spent fuel consists of two discrete regions.- Region'1 is configured' to ~ store. ' fuel with a maximum enrichment-of 5.0 w/o.
The most reactive of the Westinghouse-17 X 17 STD/ Vantage'SH and'OFA-fuel -assemblies yielded a maximum Keff of 0.940 including all biases and uncertainties.
Region 2 racks sum 3 designed to store fuel with burnup consistent _with its initial enrichment.
A table of enrichment and corresponding required burnup is provided in the Technical Specification.
A conservative value of the required burnup is given by the following 1
linear equation:
Minimum -burnup for unrestricted storage in Region' 2'in.
12100 E%
- 20500, where E
is the initial MWD /MTU
=
enrichment in w/o.
The maximum reactivity in Region 2 is 0.945 if all cells are loaded with fuel with minimum allowable burnup.
This includes all biases-and uncertainties and appropriate allowance for uncertainty in depletion calculations.
Storage cells in Region 2
which face the. pool wall are capable of.-
maintalning the Keff below 0.95 with fuel which does.not-meet the foregoing burnup restriction.
A separate calculationLto establish the admissibility of storing low burnup fuel in a Region 2 peripheral cell will be required on a case-by-case basis.
The calculation to demonstrate subcriticality for the proposed storage of low burnup fuel will be performed using the same analytical models and. computer codes which were used in the high density rack design.
i il' l-l l
1 N
DPR-66 DESIGN FEATURES i
i 1
5.4 REACTOR COOLANT SYSTEM DESIGN PRESSURE AND TEMPERATU3E 5.4.1 The reactor coolant system is designed and shall be maintained:
a.
In accordance with the code requirements specified in Section 4.2 of the
- FSAR, with allowt.nce for normal degradation pursuant to the applicable Surveillance Requirements, b.
For a pressure of 2485 psig, and c.
Fot a
temperature of 650'F, except for the pressurizer which is 680*F.
VOLUG 5.4.2 The total water and steam volume of the reactor coolant system is 9370 cubic feet at a nominal Tavg of 525'F.
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.
5.6 FUEL STORAGE CRITICALITY 5.6.1 The spent fuel storage racks are designed uith a minimur of>
12.002';
inch cantar--t; center dictance between ft:01 Occcmblice placed in th:
teragc ::cke.
The fuel will be stored in accordance with the provisions described-in UFSAR Sections 3.3 and 9.12 to keff equivalent to 50.95 with the storage pool filled with ensure a
unborated water.
DRAINAGB 5.6.2 The spent fuel storage pool is designed and shall 'ce maintained to prevent inadvertent draining of the pool belcw elevation 750' - 10".
IN A lIv0 ft lo
% - %y $yp#1 Con ht0 u tQ htdn.
k oon l tach of t Ch b t $0150ne
- e. Setts ed a+ io.a., ' +c4
- e. unA.ne s+org& As porie yl noe %-hp a << a p h k. rke b an a caeb a re g
eo consivac% wiA o cell-ti-cell puhb d toA inches.
BEAVER VALLEY - UNIT 1 5-5 Amendment No.
hA0f0 SED
4 DPR-66 DESIGN FEATURES CAPACITY 5.6.3 The fuel storage pool is designed and shall be maintained with a storage capacity limited to no mota than 449-fuel assemblies.
l If R pf -
5.7 SESIMIC CLASSIFICATION 5.7.1 Those structures, systems and componen?s identified as Category I
Items in Appendix "B" of the FSAR she.11 be designed and maintained to the original design provisione.
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.
4 BEAVER VALLEY - UNIT 1 5-6 Amendment No.
Profoseb a
s-
~
, =
ATTACHMENT B Beaver Valley Power Station, Unit No. 1 Proposed Technical Specification Change No. 202 REVISION OF TECHNICAL SPECIFICATIONS 3.9.14, 5.6.1, 5.6.3.
RERACKING SPENT FUEL POOL A.
DESCRIPTION OF AMENDMENT REQUEST The proposed amendment would allow the fuel rack storage capacity in the spent fuel pool to be increased to 1625 locations. A total of 13 free-standing rack modules in a Discrete Zone Two Region storage system will be installed in the BV-1 fuel pool.
Two modules containing a total of 162 storage cells shall be defined as Region 1,
which would permit storage of fresh fuel up to 5%
(nominal)
U-235 enrichment.
The balance of the cells, Region 2, will have an enrichment /burnup restriction on them.
B.
BACKGROUND Beaver Valley Power Station Unit 1 (BV-1) has a spent fuel pool (SFP) which at the present time contains spent fuel storage racks with 832 total storage cells.
The present racks provide adequate capacity for storage of spent fuel while maintaining reserve full core discharge capacity through 1996. Therefore, to ensure that sufficient spent fuel storage capacity continues to exist at BV-1, we have contracted for high density spent fuel storage racks from Holtec International whose design incorporates Boral as a
neutron absorber in the cell walls. The new racks have an ultimate storage capacity of 1625 fuel assemblies-(excluding two storage cans for defective fuel), which is expected to extend the full core reserve storage capability until the year 2013.
The new free-standing high density spent-fuel storage racks will store fuel in two discrete regions of the SFP.
Region 1 includes two modules having a
total of 162 storage cells. Each cell is designed for storage of fuel assemblies with Uranium-235 initial enrichments up to 5.0 wt%
while maintaining the required subcriticality (keff 5
0.95).
Region 2
includes 11 modules having a
total of 1463 storage cells, which are available for storage of spent fuel assemblies.
This region is designed-to store fuel which has experienced sufficient burnup such that storage in Region 1 is not required.
The high density spent fuel storage rack cells are fabricated from 0.075" thick type 304L stainless steel sheet material. In Region 1,
panels of the.Boral neutron absorber material are between-the cell walls and stainless steel retainers, and the-cells are separated by a specified water gap. In Region 2, the Boral panels are located between the stainless steel walls without a water gap. The cells are welded together in a specified manner resulting in a
free-standing structure which-is structurally qualified for all postulated seismic events. The nominal center-to-center spacings of the cells within Region 1-are 10.82" in both orthogonal directions.
The nominal pitch in-Region 2 is 9.02".
~.,
A
-NTTACHMENT B,-continued Proposed' Technical' Specification Change No. 202-lPage 2 Since spent. fuel is presently: stored in the BV-3 SFP,' special:
administrative controls and/or procedures: will bel developed:-to minimize radiation exposure during.tbe. installation of the-new spent fuel racks.
The evaluation of postulated accidents.with respect to nuclear criticality and/or-radioactivity release has.
shown acceptable
- results, in that k gg does not exceed 0.95, g
including uncertainties, and postulated radiological releases do not exceed 10 CFR 100 acceptance criteria..
C.
JUSTIFICATION The proposed change to the Technical Specification-' entails increasing the storage cell count to
- 1625, increasing the allowable fuel enrichment to 5%,
and modifying the-burnup restrictions imposed on fuel stored in Region 2.
The new storage system will differ from the existing storage in
~
the following key aspects:
1.
The existing racks are of the End Connected Construction (ECC) which are quite limited-in their structurallstrength.
As
- such, the existing racks provide a rather small margin against increased seismic loadings or increased-fuel weights (such-as a -consolidated fuel. bundle).
The new racks'are designed to withstand considerably higher seismic loads'than those postulated for DV-1.
2.
The existing racks are of anchored construction; the new racks are free-standing.
The anchored racks induce thermal' stresses in the pool structure due to differential. expansion between the racks and-the. pool slab.
The new racks will free the pocl structure from this avoidable loading.
3.
The new racks can withstand considerablyL higher impact' loads than the existing racks.-
4.
The fuel storage. pattarn in the new racks is simpler-than that for the existing-racks because it does not require checkerboarding of fuel in Region 1.
neution absorber materiali 5.-
The - new racks employ a
proven (Boral) with prior npplication in over-30-plants.
Thefold-racks relied on_ water slone for neutron-attenuation.
~
.w.
+
r N
.iNTTACHMENT B;fcontinued.-
16
_ProposedfTechnicalLSpecification Change No. 202 Page_3-In
- summary, the reracked BV-1 pool will' embody numerous features which--enhance isafety and reliability.
As the safetyJanalysis in the following section indicates, theresis adequate' justification for proceeding with the proposed changes to the BV-1l storage:
system.
I D.
SAFETY ANALYSIS 1.
Affected Systems The
_following systems and subsystems are potentiallym affected by the proposed modification:-
a.
Storace Racks The spent fuel storage array in the pool is'directly-affected since the existing racks will be replaced with free-standing new high density racks.
b.
Spent Fuel Pool Coolina System Because the stored quantity of fuel _in'the'BV-1 fuel.
pool will be greater than thel presently licensed inventory, the decay heat' load will be greater than the licensing basis value.
A reassessment of the fuel pool.
cooling. system has indicated its adequacy.
y 3
c.
Pool Struc.tura-No
. modification to the pool structure itself
.is-necessary.
to accommodate-the proposed ' expansion.-
- However, the wall attachments will be trimmed:if under the top of the racks.
d.
HVAC System The rate ~of water evaporated from..the' fuel pool'will~
increase due to elevated = pool water temperature. The increased moisture load - cnt the HVAC: system has been determined toobe1 acceptable.-
e.
Purification System The radionuclides ! released to the pool water-may increase due to: the' ' increase Ein the ' stored fuel inventory.
This may affect the ability of the purification'= system to maintain water purity.in the fuel-pool. _RadiologicalE ' evaluations
~ how-that the s
existing : system is adequate to: handle'theTpurificatinns load.
f t
_,d..'
r
.-r-
I$TTACHMENT B, continued Proposed' Technical Specification Change-No. 202 Page 4 2.
Safety Functions The safety function or functions of_the.affected systems, subsystems or components listed are described-herein.
a.
Storage Racks:
The storage racks' provide for: vertical upright storage of new or spent fuel assemblies in prismatic cell openings.
The racks are designed to i
maintain structural integrity during and after a DBE or j
OBE event.
b.
Cooling System:
Two trains of the. spent fuel pool j
ooling system remove decay heat from the spent fuel discharged from the reactor.
The single train of the cooling system maintains.
the-pool water
. bulk-temperature below 165*F, low enough to prevent boiling under normal' full core offload condition.
c.
Spent Fuel pool:
The spent fuel pool provides wet storage for spent fuel which is stored inside;the rack.
The racks are designed to-store the spent fuel in such a
manner as to maintain subcriticality during normal and abnormal conditions.
The pool-floor alab provides the requisite support for the storage rack and fuel assembly system during-normal and seismic conditions, d.
HVAC System:
The HVAC system removes _the heat and humidity generated by the diffusion of water vapor into the pool environment.
e.
Purification System:
The Purification System removes particulate and ionized impurities from the spent fuel pool to ' maintain pool water visibility. This system also helps maintain the desired pH balance in,the fuel-pool.
3.
Safety Evaluations The safety
-evaluations are detailed in - the attached licensing
- report,
" Spent Fuel Pool Modifications 'for Increased Storage Capacity",
Beaver Valley Power Station Unit'1, Docket No. 50-334,--Holtec Report HI-92791.
The mechanical design:of the racks meets all_ required safetyL functions stipulated for this. equipment in-the BV-1-UFSAR.
Details of the mechanical configuration-'are described in Section 6.0 of the licensing report.
The proposed rack arrays have. been analyzed to establish their structural integrity-under -OBE and DBE loadings.
Details of the analysis are. described in Section 6 0 of.the licensing report.
l I'
A
NTTACHMENT B, continued Proposed'Technica) Specification Change No. 202 Page 5 The proposed storage expansion will increase the heat load in the pool.
However, analysis has shown that the maximum cladding temperature is kept low enough by the existing spent fuel pool cooling system such that nucleate boiling or voiding of coolant on the surface of the fuel rod cladding is precluded.
The increased pool bulk temperature increases the thermal loading on the reinforced concrete structure and liner of the fuel pool.
The increase occurred for both normal and abnormal conditions.
Reanalysis of the pool structure,
- however, demonstrated that the integrity of the pool structure and the pool liner is maintained. Details of the analysis are presented in Section 8.0 of the licensing report.
It has been determined that the existing purification system is adequate to handle the increased radiological burden in Section 9.0.
The increased evaporation from the pool due to elevated water temperature is safely handled by the HVAC System in Section 11.0.
The planned expansion will not increase crud deposition in the spent fuel pool since crud deposi_'on occurs during refueling outages and new fuel racks will not affect operation of clean-up system and/or handling of fuel during refueling outages.
The pool clean-up system effectively maintains water clarity and no increase in act:.vity of the clean-up system filters or resins is anticipated.
4.
Safety Margins The NRC Staff Evaluation Review process has established that
~
the issue of margin of safety, when applied to a reracking modification, should address the following areas:
a.
Nuclear criticelity considerations b.
Thermal-hydraulic considerations c.
Mechanical, material and structural considerations The established acceptance criterion for criticality is that the neutron nultiplication factor in spent fuel pools shall be less than or equal to 0.95, including all uncertainties, under all conditions.
This margin of safety has been adhered to in the criticality analysis methods for the new rack design.
~.
OdTTACHMENT B',
continued
-Proposed' Technical-Specification Change-No. 202-
^
Page 6
- I The-methods -used in the criticality analys~s conform 1to the i
applicable portions of the appropriate-NRC_ guidance and industry
- codes, standards, and specifications.
In meeting.
the acceptance criteria for criticality in.the spent fuel-
- pool, such that k
is always less than 0.95 at a 95%/95%'
eff probability tolerance level, the proposed amendment does not-involve a
reduction in the margin of safety for nuclear criticality, as defined in the UFSAR.
The finite element method was used to evaluate the margins of the spent fuel pool concreto-structure.
The analysis is in accordance with SRP 3.8.4. The evaluation demonstrates-that the strength margin of safety of the fuel pool structure is maintained.
Conservative methods were used to calculate the maximum fuel temperature and the increase in. temperature of the water in the spent fuel' pool.
The thermal-hydraulic. evaluation used the methods previously -employed for evaluations of the present spent fuel racks to demonstrate that the temperature-margins of safety are maintained._ The-proposed modification will increase the heat load in the spent fuel pool.
The_.
evaluation shows that the existing spent fuel cooling system.
will maintain the bulk pool water temperature at or-belowJ 165'F.
- Thus, it is demonstrated that the peak value of the pool bulk temperature is considerably lower than the pool bulk boiling temperature (212'F). The. evaluation also shows that maximum local water temperatures along-the hottest fuel f
assembly are below the nucleate boiling-condition value.
- Thus, there is no reduction in the margin _tf safety for thermal-hydraulic or spent fuel cooling concerns,-as defined in the UFSAR.
The main safety function of.the spent fuel racks is.to maintain the spent fuel l assemblies in a safe configuration 7
L through all normal or abnormal loadings. Abnormal loadingi l
which have been. considered are the_effect of an earthquake and the impact due to the-drop of a spent fuel assembly. Thel mcchanical, material, and structural design of the new spent fuel racks; is in accordance with applicable portions of NRC "0T Position for Review and Acceptance ofLSpent< Fuel-Storage and Handling Applications",_ dated. April 14, 1978,.
as-modified January 18, 1979;_and=other applicable NRC guidance r-I and -industry codes.
The rack materials used are compatible with 'the spentHfuel pool and thefspent fuel assemblies. The structural ; considerations of-the new racks address' margins
~
of -safety - against tilting and deflection or_ movement, such that' tl -
racks do not impact each. other.during_ the-postulated seismic events.
In addition, ~the spent fuel-assemblies remain intact and no-criticality concerns 1 exist.
L
-Thus, the margins of safety as defined in the UFSAR are-not reduced by.the_propoced rerack.
m
~-e-
-e
"+-,w f
~-w w-we w
n
-4 ITTACHMENT B,1contjnued'.
6
_ Prop'osed' Technical Specification Change No. 202 j
i Page_7 E.
NO SIGNIFICANT HAZARDS EVALUATION The-no significant hazard considerations involved with -the proposed amendment.have been evaluated, _ focusing:on the three standards set forth in'10 CFR 50.92(c) as quoted'below:
The Commission may make a final determination, pursuant to the procedures in paragraph 50.91, that-a-proposed amendment to an operating license for a
facility licensed under j
paragraph 50.21(b) or paragraph 50.22 or for a testing 1
facility involves no significant hazards consideration,'if operation of the facility in'accordance_with the' proposed amendment would not*
I (1)
Involve a
significant increase in the probability or consequences of an accident previously evaluated; or (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 following evaluation is provided for the no significant hazards consideration standards.
1.
Does. the proposed amendment involve a significant increase in the probability or consequences-of an accident previously evaluated?
In the course of the
- analysis, the _following: potential accident scenarios have-been considered:
~
~
~
a.
A spent fuel assembly drop in the spent fuel pool b.
Loss of spent fuel pool cooling system flow c.
A seismic event d.-
A spent fuel cask drop-e.
A construction accident The increased storage capacity. of the BV-1 Spent Fuel Pit has been -analyzed for the existing fuel handling equipment and procedures,. spent. fuel pit cooling system, and seismic events.-
The movement of-the spent fuel shipping caskfis conducted by the cask ' crane, - which passes only over the.
. isolated-cask loading ~ area..Mo cask movement;is contemplated as part of this modification. The cask crane will'betused to:
bring the racks into the. Fuel Building. Thus, the: proposed mcdification.does' pot increase the probability offany of the firnt four accidents.
4
- $TTACHMENTlB,. continued LProposed* Technical' Specification Change-No. 202 Page 8 With regard to the construction accident,-_the BV-1 Technical Specifications prohibit loads
- heavier than the. weight of;a single ; spent fuel assemb?y plus the toolzfor moving that assembly from'.being carried over fuel stored in-the spent.
fuel' pool.
All work in the spent l fuel pool area will be controlle'd and performed in strict accordance with specific written ' procedures and administrative controls to preclude the. movement of a rack directly over any fuel. Therefore the-probability of a
construction accident occurring.is'not significant as a-result of the propotad reracking.
In
- addition, Sections 5.1.1, 5.1.2 and 5.1.6'of_NUREG-0612, entitled
" Control of Heavy Loads at Nuclear Power Plants",.
provide guidance for heavy load handling operations pursuant to a
spent fuel storage rack replacement.
Section 5.1.2 provides four alternatives for assuring the safe handling of heavy loads:
_during a
fuel ' storage rack replacement.
Alternative (1) of Section 5.1.2 proviu s that-the control of heavy loads guidelines can be satisf.ed by' establishing that the potential fer a heavy load drop is extremely _small..
The provisions of alternative (1) will. be met during-implementation of the. subject application.
The load handling system over the BV-1 pool consists of a movable platform with two 5-ton electric hoists.'A temporary 1 crane with an ultimate load. capacity greater than'10 times the maximum lift load will be installed for the-reracking operation.-
The maximum weight of an individual new rack or existing rack is less than 13 tons.
The weight of;the lifting fixture is less than. 2. tons.
Therefore, the temporary crane will have-an ultimate capacity.of--more than 150 _ tons.
As' per NUREG-0612, Appendix B, the substantial-safety factor margin ensures that~the probability of a; rack drop is extremely-low.
Accordingly, the proposed modification 'does not involve it significant increase in the probability-of a
load drop accident since' NUREG-0612 guidelines for defense-in-depth to-prevent load drop accidents have been satisfied.
The-consequences - of a spent. fuel assembly drop in the spent
~
fuel
_ pool were evaluated -and it_fwas found- '+kat: the criticality acceptance criterion,- keff s
0.9",
is.not-violated.
In
- addition, it was found that there:Was no significant-change in the radiological consequences'of a fuel assembly drop from the previous analyses, our annlyres found that the. calculated doses are well within 10 CFR 100 guidelines. 'The results of an analysis show that a dropped.
spent fuel assembly on-the racks willunot distort the racks such that they would not perform their safety function.
-Thus,
-the-consequences of this type of accident are-not- -
significantly_ changed from the previously evaluated spent fuel assembly drops.
- $TTACHMENT B, continued' P r o p'o s e d 'T e c h n i c a lt Specification Change No. 202
- Page 9
]
The consequences of a loss of-spent fuel pool cooling system q
flow have been evaluated and it was found that sufficient time is still available to provide an alternate' means-for cooling in the event of a-complete-failure of the cooling system.
- Thus, the consequences of this type accident are not significantly increased from previously evaluated loss-of-cooling system flow accidents.
1 The consequences of a seismic event have been evaluated. The new racks will be oesigned and fabricated to meet the requirements of applicable portions of the NRC Regulatory-Guides and published standards.
The_new free-standing racks-are
- deLigned, as are the existing-racks, so that the-integrity of the racks and the pool structure is. maintained during and after a seismic event.-Thus, the consequences of a
seismic event are not increased fron previously evaluated events.
The probability and consequences of a spent fuel cask drop will not be affected by the replacement of the racke.
The design of the cask loading crane preventL any cask movements over any region of the spent fuel pool which contains irradiated fuel.
The consequences of a
construction accident. have been considered.
A heavy load will not be carried in the spent fuel pool area until all fuel in the pool-has decayed for'a minimum of two months.
This provides sufficient time for decay of gaseous radionuclides in the fuel (gap activity) such-that an assumed accidental release.of gases from fuel damaged' by a
construction accident would result in-a potential off-site dose less than'10% of 10 CFR 100' limits..
In addition, there is no equipment which is essential =to.the safe shutdown of the reactor or employed to mitigate the consequences of an accident which is beneath, adjacent to,-
or otherwise with4 % the area of influence of any loads 1that-
~
will be handled during the expansion modification..
Therefore, the-consequences of a construction accid 9nt are not significantly.
increased from previously evaluated events.
Therefore, it is concluded Ethat the proposed amendment to replace the spent-fuel racks in the spent fuel.poolidoes not involve a
significant increase in the probability or consequences of an accident-previously_ evaluated.
y
,~,.
,7,
- - - ~ ~,
4 5TTACHMENT B, continued 202 Prop'osed'TechnicalLSpecification Change No.
Page 10' 2.
Does the proposed' amendment-create the possibility of a-new or different kind. of accident from any accident previously evaluated?
The proposed modification was evaluated in-accordance with i
the guidance of the NRC_. Position Paper _ entitled,-
"OT l
Position for Review and Acceptance of Spent Fuel Storage and j
H Handling Applications",
appropriate NRC-Regulatory Guides, appropriate NRC Standard Review
- Plans, and appropriate industry cot s and standards.
In addition, several previous <
NRC Safety Evaluation Reports for rerack applications similar to this proposed modification have been reviewed.
1 No unproven technology will be. utilized either in the-construction process or in the analytical techniques necessary to justify the -planned fuel storage expansion.
1 2
The basic reracking technology in this instance Ima 'been developed and demonstrated in over 80-applications for' fuel pool capacity increases previously_ approved by.the NRC.
The change to a ctwo-region spent fuel pool requires the performance _ of additional evaluations to ensure that the criticality criterion is maintained.-
These include the:
evaluation of the limiting criticality-condition, i.e.,
misplacement of an 'unirradiated fuel _ assembly of
-5.0%E enrichment- -into a
Region 2
storage cell pr-outside'and-adjacent to a Region 2 rack nodule.
The evaluation for this case shows that when the' boron concentration meets the proposed technical specification requirement,-
the criticality criterion is satisfied.
Although-this-change does-pose the need to address additional aspects _ of a previously analyzed
- accident,
--it - does not-create.the-possibility of a previously unanalyzed accident..
Based upon the foregoing, it is concluded that the proposed reracking does not create the possibility of
-a:
new or different type _
accident
-from-any acciaent previously evaluated.
3.
Does the proposed amendment involve.a significant reduction.
in a margin of safety?
The-NRC_
Staff _
Safety Evaluation Review :-process __ hasf established. that the issue cf margin of safety, when applied
-to a -reracking--modification, :should? address 1theJfollowing areas:
a.
Nuclear criticality considerations b.
Thermal-hydraulic considerations c.
-Mechanical, material and structural considerations m.
=
R 15TT'ACHMENT B, continued froposed Technical Specification Change No. 202 Page_111 The established acceptance criterion for criticality is that the neutron multiplication factor _in spent fuel pools shall be less than or equal to 0.95, including all uncertainties, under all conditions.
This. margin of safety has been adhered to in the criticality analysis methods for the new rack design.
The methods used in the criticality analysis conformed.to the applicable portions of the appropriate NRC guidance.and industry
- codes, standards, and specifications, as listed in n
the Licensing Report. In meeting the acceptance criteria for criticality in the spent fuel pool, -such that k
is eff always less than 0.95, including-uncertainties at a-95%/95%
probability confidence
- level, the proposed amendment does not involve a significant reduction in the margin of safety for nuclear criticality.
Conservative methods were used to calculate the maximum fuel temperature and the increase in temperature of the water in the spent fuel pool. The thermal-nydraulic evaluation used the methods previously employed fer evaluations of-the present-spent fuel racks to demonstrate that the temperature murgins of safety are maintained.
The proposed _ modification will increase the heat load
-in the spent fuel pool.
The evaluation shows' that the existing spent fuel cooling system will maintain the bulk pool water teinperature below 165'F.
- Thus, a
margin of safety exists such that the maximum allowable temperature for bulk boiling is not exceeded.for the calculatod increase in pool heat load.
The' evaluation also shows that maximum local water temperatures along the hottest.
fuel assembly are below the nucleate boiling condition value.
Thus, there is no significant reduction in the margin of safety for thermal-hydraulic or spent fuel cooling concerns.
The main safety function of-the spent fuel pool and.the racks is to. maintain _the spent fuel assemblies.in a safe configuration
- through -all. normal or abnormal loadings.
Abnormal loadings which have been considered are the effect of-an: earthquake, the impact due to a spent fuel cask drop, the drop of a' spent fuel 1 assembly, or the drop of any object used in the rerack modification. The mechanical,. material, and _ structural -design of the new spent ~ fuel' racks is in accordance with_ applicable -portions of.
"oT' Position for Review and Acceptance -of Spent Fuel Storage'and Handling Applications",
dated April 14, 1978,_as modified January 18, 1979; Standard Review Plan 3.8.4; and other. applicable _NRC.
l guidance.and' industry codes.
The rack-materials used are compatible with the spent fuel pool and the spent fuel assemblies.
The -structural considerations of the new racks address margins 'of safety against-tilting and deflection or
~
- movement, such-that the racks do not impact each other-in L
-n
+,,
__~
ikTTACHMENTgB,. continued
- Proposed' Technical Specification Change Nc. 202 Page 12 the -cellular region during the-postulated seismic events. In
- addition, the spent fuel assemblies remair, intact and no criticality concerns exist.
Thus, the margins of safety are not significantly reduced by the proposed rerack.
Additionally, the proposed amendment most closely-resembles example (X) of
" Amendments That Are Considered Not Likely to Involve Significant Hazards Considerations" as provided in the final NRC adoption of 10 CFR 50.92, 51 FR 7751 (March 6,-1986).
This example indicates that an amendment is not likely to involve a sianificant-hazards consideration as follows:
(X)
An expansion of the storage capacity of a spent fuel pool when all of the following are-satisfied:
1.
The storage expansion method _ consists of either-replacing existing racks with a-. design which allows closer spacing between stored spent ael assemblies or placing additional racks of the original design on the pool floor if space permits.
The BV-1 spent fuel pool rerac'. involves the replacement of the present capacity racks with a-design
- which, by requiring only burned fuel be-stored in Region-2, allows closer spacing of the stored spent fuel cells.-
Region 1 is designed for i
allowing safe storage of fuel enriched to 5.wt%.
2.
The storage expansion method does not involve rod consolidation or double tiering.
The BV-1 racks are not double-tiered and'all racks will sit on the
' spent fuel pool floor.
Additionally, the amendment application-does not involve consolidation of spent fuel.
3.
The k
of the pool is maintained less than or eff equal to 0.95.
The design of the-spent fuel-racks contains a neutron
- absorber, Boral, -to ensure that the k,tfi gf remains less than 0.95 under all conditions (wi unborated water in the pool). Additionally,' the water in the spent fuel pool contains at least 1050 ppm of - boron, providing further assurance that. k remains less _than_
0.95. The analysis egg demonstrates that 400- ppm boron is required to meet the reactivity requirement for the accident condition.
~...
?NTTACHMENT[B, continued
- Proposed Technical 1 Specification-Change No. 2021
- Page-13' 4.-
No new technology.or unproven technology is utilized-in either the-- construction process or: the -
analytical-techniques.necessary' to ' justify..the.
expansion.-
i The rach design has beeni licensed at least 15 times.
The technology.
for the ; construction 1 processes and~. analytical
. techniques-remain substantially the same as these other 15 storage rack projects.
Thus no new or unproven technology-is utilized in the construction or analysislof;the proposed BV-1 spent fuel racks.
- Thus, this submittal meets example -(X) _ = as presented in.the supplementary information accompanying publication of the Final Rule as an example' of-situations which are considered not to' involve significant hazards considerations.
Based on the foregoing,- it is' concluded that all criteria for issuance of a no significant hazards statement are_ satisfied.
F.
NO SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION Based on the. considerations expressed'above, it is. concluded that the activities. associated with this license amendment request satisfies the no significant hazards consideration standards ~of 10 CFR 50.92(c)
- and, accordingly, a.
no.significant hazards' consideration finding is justified.
3 a
v um
ATTACHMENT C Beaver Valley Power Station, Unit No. 1 Proposed Technical Specification Change No.-202
. Typed Pages:
3/4 9-14 3/4 9-15 B 3/4 9-3 B 3/4 9-4 5-5 5-6 l
l 1
1
-l
.bPR-66 l
REFUELING OPERATIONS 3/4.9.14 SPENT FUEL STOPJaGE POOL 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 1
b.
Fuel assembly storage in Region 1 restricted to fuel with an enrichment less than or equal to 5.0 w/o U235; and l.
c.
Fuel assembly storage in Region-2 restricted to fuel which has been qualified in accordance with Table 3.9-1 orja criticality analysis..
APPLICABILITY:
During storage of fuel in the spent fuel pool.
ACTION:
a.
Suspend all actions involving movement of. fuel in the 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 acsembly.
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 10$0 pp until such time _as'the baron concentration is increased to 1050 ppm or greater.
c.
The provisions of Specifications 3.0.3 and-3.0.4Lare 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, by burnup analysis and_ comparison with Table 9-1, or.by a
criticality analysis
_that_ fuel assemblies to be placum into or_ moved infthe. spent-fuel pool are within the above enrichmentflimits.
4.9.14.2 Verify _the-spent fuel pool; boron concentration is E 1050 ppm:
a.=
Within 8
hours prior to and.
at-__ leastHonce per 24-heurs o
during movement of fuel in_the spent fuel 7 pool, and'
- b. ;
At least once-per 31 days.
L BEAVER VALLEY - UNIT-1
.3/4 9-14 Amendment-No.
Ll PROPOSED l
n T
L
i l
[DPR-66 1
TABLE 3".9-1 BEAVER VALLEY FUEL ASSEMBLY MINIMUM-BURNUP'VS. INITIAL U2351 INRICHMENT FOR' STORAGE IN REGION'2 SPENT FUEL RACKS
-i Initial U235
-Assembly Discharge-Enrichment Burnuo (MWD /MTU) 2.0 2585 2.5 9551 3.0 15784-i 3.5 21643 4.0 27260 4.5 33710 5.0 40000 NOTE:
The data in the above table' may be either interpreted linearly or may be calculated by theLconservative equation below._
This equation provides a linear fit to the design burnup-limits.
Limiting burnup, MWD /MTU = 12100
- E% - 20500 Where E = Enrichment-(E-s 5%)
l i
L L
BEAVEREVALLEY - UNIT 1 3/4 9'-15 Amendment'No.
PROPOSED
.bPR-66 ILEFUELING OPERATIONS BASES 3/4.9.10 A!Q___3]_4. 9.11 WATER LEVEL - REACLOR VESSEL AND STORAGF POOL The restrictions on minimum water level ensure that sufficient water depth is available to remove 99%
of the assumed 10% indine 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 SY STE_fi The limitations on the storage pool ventilation system cnsure 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 o^ this system and the resulting iodine removal capacity are consistent with the assumptions of tt 1 accident analysis.
The spent fuel pool area ventilation system is non-safety related and only recirculates air through the fuel building.
The SLCRS portien of the ventilation system is safety-related and maintains a negative pressure in the fuel building.
The SLCRS flow is normally exhausted to the atmosphere without filtering, however, the flow is diverted through the main filter banks by mannal actuation or on a high radiation signal.
3/4.9.14 FUEL _ STORAGE - SPENT FUEL ST6 RAGE POOL The requirements for f uel storage in the spent fuel pool ensute that:
(1) the spent fuel pool will r : main 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 16.1-1975.
The value of 0.95 or less for keff which includes all uncertainties at the 95/95 probability / confidence level is the 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 2 1050 ppm (this includes a 650 ppm l
conservative allowance for uncertainties) daring 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 Keff is s 0.95; and (2) the boron concentration meets the 1050 ppm limit.
BEAVER VALLEY - UNIT 1 B 3/4 9-3 Amendment No.
PRO"OSED
. - - - - ~ ~.
- DPR-66
-REFUELING OPSEM1911e BASES
-FUEL STORAGE'- SPENT FUEL STORAGE POOL (Continued)
The reracked spent fuel consists of.two discrete regions.
Region 1 is configured' to store. fuel with a ma) ' mum enrichment of 5.0 w/o.
-The moet reactive of the Westinghouse 17 X
,7 STD/ Vantage SH and OFA fuel assemblies yielded a maximum Keff of 0.510 including all biases l
and uncertainties.
The enrichment limitations for storage of fuel in Region 1 of the spent fuel pool is based on a nominal region average enrichment with indivicual fuel assembly tolerance of
+
or - 0.05 w/o of 5.0 w/o U-235.
Region 2 racks are designed to store fuel with burnup consistent with its initial enrichment.
A-table of e iclment and corresponding required burnup is provided in the Tew.nical Specification..
.A conservative value of the required burnup is given by the following-linear equation:
Minimum burnup for unrestricted storage in Region 2-in
- 20500, where E-is the; initial 12100 E%
MWD /MTU
=
enrichment in w/o.
The maximum reactivity in Region 2 is 0.945 if all cells are loaded with fuel viith minimum allowable burnup.
This includes all biases and uncertainties and appropriate allowance for.-uncertainty in depletion calculations.
Storage cells in Region 2
which face the pool wall are. capable of-maintaining the Keff below 0.95 with fuel which does not meet the:
foregoing burnup restriction.
A separab calculation to establish the admissibility of storing low burnup-fuel in a Region 2 peripheral-cell will be required on a case-by-case basis.
The_ calculation to:
demonstrate subcriticality for the proposed storage of low burnup fuel will be performed-using the-same analytical models and computer z l
codes which were used in the-high density rack design.
l l
3/4.9.15 CONTROL ROOM EMERGENCY HABITABILITY SYSTEMS i
The OPERABILITY of the v.
'ol room emergency habitability system ensures that the control will' remain habitable for operations personnel ~during and followa' mil credible accident conditions =
~The o
!~
amb4 t
air temperature is controlled to-prevent exceedingL-the L
allw.able-equipment qualification temperature:for'the equipment and instrumentation. in the control room.
The OPERABILITY of this system in conjunction with control room design provisions is based on
- limiting the whole body radiation exposure to personnel' occupying the control room to 5 rem or less, or its equivalent.
This limitation-is consistent with the requirements of General Design Criteria 19 of.
Appendix "A",
-BEAVER VALLEY - UNIT 1 B 3/4 9-4 Amendment No.
PROPOSED
~
- DPR-66 DESIGN FEATURES
_ = =
~
n
- =- -
.- - =-
- - = -
- -.a I
RI2LCIDlLC99LhML.METW DESI9N PREEEURii_AtRLlTMPrinTURE 5.4.1 The reactor coolant system in designed and shall be maintained:
a.
In accordance with the code requirements specified in Section 4.2 of the
- FSAR, with a1lowance Ior norma 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 the pressurizer which
~
is 680"F.
Y0lMit 5.4.2 The total water and steam volumn of the reactor coolant system is 9370 cubic feet at a nominal Tavg of 525 F.
LLDIERG ERRLCDB E_CDo blitG_JdHTIES 5.5.1 The emergency core cooling systems are designed and shall be w
maintained in accordance with the original design provisions contaired in Section 6.3 of the FSAR with allotance for normal degra.lation pursuant to the applicable Surveillance Requirements.
~
L.6.ltE kliloBAg1 CRITTcALITl 5.6.1 The spent fuel corage racks are designed in a two region configuration.
Region 1
racks are of the F71soned flux-trap type with the storage cells arranged nt 10.82 inen pitch.
The Region 2 racks are of the poisonec:
non-flux trap construction with a
cell-to-cell pitch of 9.02
.nches.
The fuel will be stored in accordance with the provisions described ir. UFSAR Sections 3. 3 and 9.12 to enttre a
keff equiv-lent to
$0.95 with the storage pool filled with
..> orated water.
DEMEA" 5.6.2 TPe spent fuel storage pool is designed and shall be m6.i rita i ned to nrevent inadvertent draining of the pool below elevation 750' -
10".
BEAVER VALLE'i - UNIT 1 5-5 Amendment No.
PROPOSED
.~
4
=
DESIGN FEATURES
.4m e 4 CAPACI_T.1 5.6.3 The-fuel storage pool in designed and shall be maintained with a storage capacity limited to no more than 1627 fuel assemblies.
l 5.7 _FEISMIC CLASSIFICATION 5.7.1 Those structures, systems and components identified as Category I
Items in Appendix "B" of the FSAR shall be designed and-maintained to the original design provisions. with allowance for-
- l normal degradation pursuant to' the applicant Surveillance Requirements.
1 i
b8 11[;.TLQILOLOGICAL TOWEJLLOCATION I
5.8.1 The meteorological tower shall be located as shown on Figure 5.1-1.
b 1:
l BEAVER VALLEY - UNIT 1 S-6 Amenument No.
PROPOSED l
~. -
A T T A Cll!4 E!1 T D Beaver Valley Power Station, Uni t llo. 1 Proposed Technical Specification Change 11o. 202 Licenning Report SPE11T PUEL POOL liODIFICATIO!1 for 111CitEASED STORAGE CAPACITY DEAVER VALLEY POWER STATIO!1 Ut31T 1 r
l
-~