ML20215J204
| ML20215J204 | |
| Person / Time | |
|---|---|
| Site: | 05002702, 07002972 |
| Issue date: | 09/23/1986 |
| From: | Johnson I COMMONWEALTH EDISON CO. |
| To: | Jennifer Davis NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS) |
| References | |
| 2153K, 27450, NUDOCS 8610240348 | |
| Download: ML20215J204 (59) | |
Text
n), One First Nabonal Plaza. Chicago, Hlinois ahURN M 3%
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Chicago, luinois 60690 - 0767 September 23, 1986 I
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Mr. John G. Davis g(({ly[h Director of Nuclear Material Safety i
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& Safeguards U.S. Nuclear Regulatory Commission
'1 2 SEP 3 01986 }.-:
Washington, DC. 20555
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.S. NUCLEAR REGutMORY.,1 9 U COMylSSION
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NMSS A
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Subject:
Byron Generating Station Units 1 & 2 9
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'Dl O Special Nuclear Materials License (SNM)-1916 E C Docket Nos. 50-454 and 50-455
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$Q w Reference (a): Letter from I.M. Johnson to Materials Licensing Section, NRC Region III, dated September 3, 1986
Dear Mr. Davis:
The above referenced document requests several changes to Byron Station Special Nuclear Materials License 1916. These changes involve the installation of Byron Unit 2 start-up sources and provisions for the wet storage of Unit 2 fuel in the spent fuel pool.
There have been several conversations with Mr. Norman Ketzlach of your staff. As a result, we are submitting a revised amendment request which is attached to this letter. This revised request incorporates all comments
~ made by Mr. Ketzlach in the course of his review.
For your convenience we are enclosing 2 copies of the requested amendment for your review.
Please direct any questions you may have regarding this matter to this office. In accordance with the provisions of 10CFR 170.11, there is no fee associated with this amendment.
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Yours very truly, DOCKETED
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I. M. Johns 9
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/,N4 & \\g' Attachment A: Pro Amendment to Byron Station Special Nuclear Materials License No. 1916 p
T B: Portions of the Byron Station
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FSAR Applicable to the Proposed N
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20 CONTROL 110.
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Attachment A Proposed Amendment to the Byron Station Special Nuclear Materials License Number 1916 Present Wording:
20 New fuel assemblies may be stored in the. spent fuel storage pool subject to the following additional conditions:
a.
The maximum U-235 enrichment shall be 3.22 w/o.
b.
The fuel assemblies shall be stored in a checkerboard pattern.
c.
The Fuel llandling Foreman shall verify correct fuel assembly location after insertion of each fuel assembly into the assigned storage rack in accordance with a prepared written procedure approved by an Operating Engineer and the Technical Staff Supervisor.
d.
An independent loading verification shall be made by a Quality Control Inspector.
e.
The fuel handling foreman and the quality control inspector shall each sign a document assuring proper storage of each fuel assembly.
f.
The fuel assemblics shall be protected to preclude damage and preserve fuel assembly cleanliness while fuel is stored in the spent fuel storage pool.
Proposed Wording:
20A New fuel assemblics may be stored dry in the spent fuel storage pool subject to the following conditions:
a.
The maximum U-235 enrichment shall be 3.22 w/o.
b.
New Fuel Assemblies shall be stored in a checkerboard pattern.
c.
When new fuel is stored in a dry condition, the Fuel llandling Foreman shall verify correct fuel assembly location after insertion of each fuel assembly into the assigned storage rack in accordance with a prepared written procedure approved by an Operating Engineer and the Technical Staff Supervisor.
v
Attachment A Page 2 d.
When new fuel is stored in a dry condition, an independent loading verification shall be made by a quality control inspector.
c.
When new fuel is stored in a dry condition, the Fuel llandling Foreman and the Quality Control Inspector shall each sign a document assuring proper storage of each fuel assembly.
f.
When new fuel is stored in a dry condition, the fuel assemblics shall be protected to preclude damage and preserve fuel assembly cleanliness.
20B New fuel may be stored in the spent fuel storage pool while the pool is being filled subject to the following conditions:
a.
Borated water of at least 2000 ppm boron concentration is utilized to fill the spent fuel storage pool.
b.
The new fuel assemblies remain in a checkerboard loading pattern until the pool is completely flooded.
c.
The maximum U-235 enrichment shall be 3.22 w/o.
20C New fuel may be stored in the spent fuel storage pool in a flooded condition subject to the following conditions:
a.
The boron concentration of the filled portions of '.he spent fuel storage pool is at least 2000 ppm.
b.
New fuel will be moved and stored in accordance with subsections 9.1.1, 9.1.2 and 9.1.4 of the Byron FSAR, Amendment 46.
29.
Two fuel assemblies may be placed in two water filled failed fuel containers for primary neutron source installation and shielded storage provided that the water is at a boron concentration of at least 2000 ppm.
Basis For Change The changes to items 20A a) through f) have been made to include the dry storage provisions specifically in the license conditions of SNM - 1916. (The provisions for dry storage of new fuel in the spent fuel storage pool were approved in the SNM License No.1916, dated March 4,' 1985. These proposed changes merely reference this provision in each license condition where it is applicable.
l
m Attachment A Page 3 x
Items 20B and 20C will permit the storage of new fuel in the spent fuel pool in the event it is flooded and is being flooded with borated water. This provision is sought in the event that Byron Unit 2 fuelis loaded after the Byron Unit I refueling outage has commenced.
Chapter 9 of the Byron FSAR addresses three different cases; 1) dry storage of new fuel with 3.22 enrichment in a checkerboard loading pattern; 2) optimum moderation with,Keff less than 0.98 and 3) flooded conditions with Keff less than 0.95. Criticality analyses have shown Keff to be less than the design limit in all cases.
Therefore, the storage of new fuel in the Byron spent fuel storage pool, while it is being flooded (item 20B) and the storage of new fuel in a flooded pool (item 20C) is bounded by the criticality analyses found in chapter 9 of the Byron FSAR.
Item 29 will modify SNM license 1916 to allow the temporary storage and installation of the startup sources. Due to ALARA considerations, it is desirable to install the two californium primary neutron sources (up to 225 millicuries each)into new fuel assemblies in the spent fuel pool. The two fuel assemblies to receive primary sources will be installed in failed fuel containers located in two nonadjacent failed fuel container rack locations 44 inches apart. The failed fuel container has the cylindrical dimensions of a 13 inch schedule 10 pipe and contains a stainless steel angle framework centered in the container to support the fuel assembly. This framework is nominally 9 inches square. Between the two failed fuel containers will be a third borated water filled failed fuel container to be used as a temporary storage location for the primary source rod. This arrangement has been analyzed by NUS Corporation, the original rack criticality analyst, and found to have no adverse affect on the original results.
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n Attachment A Page 4 Normal station radiation procedures will be used to support the movement of the primary neutron sources. The concepts of time, distance and shielding will be utilized to minimize exposure. Ilealth physics personnel will monitor personnel exposure, area dose rates, conduct area surveys and contamination swipe surveys and appropriately post any radiation areas.
Personnel exposure will be kept below 200 milliman Rem for the source installation.
(Actual exposure accumulated as a result of the Unit One activities was less than 30 milliman Rem)
Dose rates from the unshielded source are projected to be 3.5 mrem /hr at 10 m.
Expected dose rates for source loading both with and without borated water shielding are as follows for selected areas:
Area Without Water With Water Welding work area 158 mrem /hr 0.00326 mrem /hr (9 ft. above source) 426' elevation at north end of pit 4.16 mrem /hr 0.079 mrem /hr South SFP wall at source level 688 mrem /hr 50.8 mrem /hr 426' elevation at south end of pit 9.51 mrem /hr 5.9 E-6 mrem /hr
- 09/02 i
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n-Attachment B Portions of the Byron Station FSAR Applicable to the Proposed SNM Amendment O
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AMENDMENT 38 MAY 1982 Ihe spent tuel bridge crar.e transt'ers new tuel f rcm the new f uel elevator to the upending device and spent r uel f rom the apendir.g device into a spent fuel storage rack.
Ine new f uel is transferreo Irom tne new fuel racks to the new fuel elevator cy the fuel handling ou11 ding crane.
The ruel hanaling area plan view is snown in Figure 1.2-9 ano in sectional elevation in Figure 1.2-11.
fue spe n t f uel handling operation is carried out entirely under water.
The nea ruel nandling operation is alsc normally carried out anaer water except during the ir.terval when the new f uel is transterreo from the snipping contaiaers to the new tuel storace drea or new t uel elevator, or ossibly during initial Core e
loading.
Initial core fuel with up to 3.20 w/o U-235 enrichment may be stored dry in the spent fuel storage racks in a checkerboard pattern only, prior to the initial core loading, utilizing administrative controls to prevent improper loading.
Detailed discussions are given in Subsection 9.1.2.
The tuel handling building which encloses the sper.t ruel pool, new t ue l storage area and eortions or the ruel transrer system is d Seismic Cdtegory I huilding.
It is therefore capable of witnstar. ding the design-basis earthquakes, tornados, and missiles, and is not subject to flooJing.
Details of the seismic design are given in Sucsection 3.o.4.1. 2.
9.1.1 New Fuel Storage 9.1.1.1 Design Bases New f uel is otorea dry in racks uesigned to provide storage for 132 tuel assemblies.
The oesign of the normally cry new storage r act.s is sucn that the ef fective maltiplication f actor does r.ot exceed the oesign-oasis limit ot 0.9d witn t uel of a maximum enrichment of 4.00 w/o U-235 in pis cs, assuming optimum neutror.
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modera tion canditios.s (dry or fogged) in accoruance with ANSI l
N1o.2-1973.
The nea tuel storage racks are desigrad to preclude storage of a l
r uel asse mbly other than where intenlea, dr.a to withstand tr.e uplitt turce wnich could occur due to ruel assemoly hanging up l
during litting and tne impa ct lcad of a aroppea fuel assemoly.
1 2ne ne. tuel storage f acility is designed to prevent flooding.
Nevertneless, the new fuel storage racas are oesigr.ed such that ene ef rective multiplication tactor daes not exceed 0.95 with fuel or a n.aximum enrienment or 4.00 w/o U-235 in place, assuming the stored assemolies completely submergeu ir. unbor ateo water at conservative water temperature una with r.o credit tor neutron a
yoisur. t r. the t ue l assemol).
9.1-2
i.
B/B-FSAR AMENDMENT 38 MAY 1982' The new fuel storage racks are designed to withstand the safe shutdown earthquake and design-basis wind and tornado loadings.
For further information see Subsections 3.2.1, 3.3.1, and 3.3.2.
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j B/B-FSAR AMENDMENT 43 SEPTEMBER 1983 9.1.1.2 Facility Description New fuel assemblies and control rods are received in an area of the fuel handling building separate f rom other fuel and are stored in racks in the new fuel storage vault af ter removal from the shipping containers.
The fuel for the initial core loading is temporarily stored in the spent fuel pit.
This area is shown in Figures 1.2-9 and 1.2-11.
New fuel assemblies are stored in racks in a reinforced concrete pit located in the fuel handling building as shown in Figure 1.2-9.
The new fuel storage pit provides dry storage for 132 fuel assemblies which is approximately 2/3 of a core.
The new fuel storage installation is shown in Figure 9.1-1.
Each rack consists of individual lockable vectical cells of stainless steel construction.
A group of 44 cells, in rows of two, at a center-to-center spacing of 21 inches is assembled to the support structure.
The support structure is bolted and/or welded to embedded studs in the concrete walls and columns of the storage pit.
The fuel assembly rests on a self-leveling support plate.
All l
surf aces that contact fuel assemblies are austenitic stain-less steel.
The supporting structure is high strength carbon l
steel.
The racks are designed to withstand normal operating loads as well as design-basis seismic loads.
The ability to handle off-normal refueling sequences is assured by the quantity of the spent fuel storage racks which will handle 1050 fuel assemblies.
The total storage capacity is 1050 fuel assemblies in the spent fuel pool, 10 f ailed fuel assemblies in the spent fuel pool, and 132 new fuel assemblies in the new fuel storage racks.
The new fuel storage racks rest on an 18-inch floor slab with 2-foot thick reinforced concrete walls.
Access to the area is through openings in the floor at print elevation 426 feet 0 inch.
The new fuel storage racks are laterally supported by attachment to embedment plates in the walls.
9.1.1.3 Safety Evaluation Criticality Control The design of the new fuel storage racks provides for a subcritical effective mutiplication factor (k for the optimum moderation' condition of less than 0.9Ig$n)d for the flooded condition of less than 0.95 in accordance with ANSI N18.2-1973 and ANSI N210-1976, respectively.
The optimum moderation' condition exists when the new fuel racks are covered with clean water at a uniform density of 0.018 gm/cc.
9.1-3
r B/B-FSAR AMENDMENT 38 MAY 1982 The flooded condition exists when the new fuel racks are covered with cold, clean, unborated water with a full density.
The criticality analyses of these two conditions included several conservative assumptions as well as the effects of uncertainties in the calculation method, and geometric and material variations of the fuel storage rack.
The follow-ing conservative assumptions were used in the calculation:
a.
Fresh fuel of 4.00 w/o U-235 enrichment - Initially the maximum enrichment will be lower than this, but it could approach this enrichment'if an 18-month fuel cycle is used.
b.
The neutron leakages in the' vertical direction and in one of the radial directions (X) are neglected while the neutron leakage in the other radial direction (Y) is accounted for.
Fixed neutron absorbers in the fuel assembly and c.
soluble poison in the water in case of flooding are neglected.
The calculations to determine the optimum moderation condition were performed with a Monte Carlo neutron transport code, KENO, with a 123-group cross section library generated from a basic GAM-THERMOS library using the NITAWL routine in the AMPX code package.
The effects of small changes in rack parameters on rack system criticality were determined with a four-group diffusion theory code, PDQ-07, and with the neutron cross sections generated from the NUS Corporation version of LEOPARD, NUMICE.
Both PDQ-07 and the KENO cal-culation methods have been benchmarked.
Under the optimum moderation condition of 0.018 gm/cc water with full storage racks, the k as determined from KENO is 0.9250.
Calculationalunce7k$intiesweredetermined from both comparison between calculation and experiments using KENO, and a statistical evaluation of Monte Carlo The results are a calculation uncertainty for the runs.
former of ?.0086 ok and, for the latter, 0.0089 ak, at a 95% confidence level; or a total calculation uncertainty of 0.0175 Ak.
Variation in material composition and mechanical spacing and tolerances acting in a direction to increase the system koffb.were also analyzed and resulted in a reactivity increase 0342 ok.
Adding reactivity ef fects of calculation uncertain-(
ties of 0.0175 Ak and geometric and material uncertainties of l
0.0342 ak to the optimum moderation k results in a maximum eff 9.1-3a i
B/B-FSAR AMENDMENT 38 MAY 1982 k
of 0.977 with a 95% probability at a 95% confidence 18881, which satisfies the design basis of maintaining k,gg below the design-basis limit of 0.98.
When the full new fuel storage racks are flooded with cold unborated water, the nominal k is 0.8655.
Adding reactivity effects of calculation uncertaT$kies of 0.0173 ok, geometric and material uncertainties of 0.0134 ok, and the temperature increase effect (from 68' F to 170' F) of 0.0024 ok to the nominal k probabiliIhfresultsinamaximumkata95%confidencelev$kfof0.899witha95%
which satisfies the design basis of maintaining k,gf below the design-basis limit of 0.95.
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9.1-3b l
O B/B-FSAR AMENDMENT 19 MARCH 1979 New Fuel Storage Rack Structural Design In accordance with requirements for sci.smic design classifica-tion, the new fuel storage racks are classified Seismic Category I.
Structural integrity of the fuel racks when suojected to design-basis seismic loads has been demonstrated.
The loads, load combinations, and structural acceptance criteria (strength limits) that were considered are identical to those considered for the spent and failed f uel storage racks, (subsection 9.1.2), except that thermal loads were considered negligicle and therefore not included.
The seismic loading of the new fuel rack is determined from a response spectrum modal dynamic analysis in which the stif fness of the f uel assemoly is neglected.
However, the mass of the fuel assemblies is considered to be uniformly distriouted along the storage tubes in the north-south and east-west directions.
The entire mass of the fuel is assumed to be supported by the floor in the vertical direction.
Racks are modeled in detail using finite elements.
Since the cans and fuel rest on the floor, a two-dimensional model representing either the upper or lower support system is used for the vertical earthquake.
The can is vertically rigid and was analyzed for vertical earthquakes.
The seismic analysis is performed using the STARDYNE computer program.
To determine the earthquake response, STARDYNE is rirst run to determine the natural frequencies and participation factors.
For frequencies with significant modal participation, mode shapes and modal loads are calculated using the appropriate response spectra.
Closely spaced modes are combined directly and l
then combined in a square root of the sum of the squares (SRSS) i manner with other significant modes.
The results of the three di_rections of earthquake are combined in an SRSS fashion as directed in Combining Modal Responses and Spatial Components in Seismic Response Analysis (Regulatory l
Guide 1.92).
Using the previously listed loads and load combinations, stresses were calculated at critical sections of the racks.
The results of the structural and seismic analyses demonstrate that the fuel racks are structurally adequate and meet the design criteria.
Since rack integrity is maintained there would be no damage to the stored fuel assemblies and no increase in Keff uncer these loads.
New Fuel Rack Design Features The new fuel racks were designed and fabricated with a high degree of reliability and integrity.
The codes and standards used for the new fuel. racks are the same as those used for the spent and failed fuel storage racks and are listed in Subsection 9.1.2.
I 9.1-4 C
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B/B-FJAh AMENDMENT 38 MAY 1982 9.1.2 JEent Fuel Storage 9.1.1.1 Cesign aaoes r us1 storage space is provioed in the spent luel storage pool for 1050 fuel assemblies and 10 damaged fuel assemblies.
.n: spent ruel racx design precludes storage of a ruel assembly locatior.s in the lacx other than unose designeo to receive at u sae molle s.
Tha center-to-center spacing tetween stored f uel assemclies in a tally loaded spent ruel storage racx is suf ficient to malr.t ain a Ref f egaal to or less than 0. 95 tor tne normal wet conoition, and ror all abnormal and accident conlitions.
Ihis design casts is met with fresh tuel or up to an enrichment or 4.00 w/o U-235, a l
conservative water temperature (ebo F), no credit for fixed neutron poison in tne ruel assemoly or soluole neutron poison in the pool water, and no credit f or structurdi material other than the stainless steel cans.
Ine s er.t fuel racks and the failed fuel canisters were desigr.e3 e
to allow adequate cooling or tnc ag:nt tuel assemtlies.
Both tne failed fuel racks are classirled Seismic Category I.
s pe nt ano
.ney are designed to withstar.d the ef tects or the Sbt, remain functional and maintain suLcriticality.
Tne spent fuel racks were desigr.ea to withstar.d either a dropped ruel assemnly or the upward torce of a stuck assemtly without loss or f unction.
The structural analysis is discussed later in this section.
Shielding 101 the spent tuei storage $rrangement is safricient to rotect elaat personnel trom exposure to as low as reasonably e
dCnievdble ana well helow 10 cra 20 limits.
Ihe spent f uei storage facility was designed to prevent tornaao-generated missiles from causing damage to the f uel.
9.1.2.2 Facilities Description Spent Fuel Storage Facxs Ine spent r uel storage racxs t rovide a place in the spent tuel storage pool for storing tne spent ruel uischarged trom tne reactor vessel.
They are top entry racxs, designeo to maintain the spent fuel in a space geometry that precludes the poss1Lility or criticality under ooth normal ana abnormal conditions.
The design or the spent ruel storage rock assembly prevents any possioility of accidental criticality.
Ine location or the spent f arl pool within tne plant is snown in r igure 1.2-9.
A general a rrangemen t of spent tuel storage racilities is shown in lioure 9.1-2.
9.1-5
c B/b-FJAR AMENDMENT 38 MAY 1982 Ahe spent tuel storage racks stainless steel. stainless steel tubes or 1/8 inch tnicx 304(snown in ligure 9 square consist of They are held at 14 a ustenitic tyee 304 stainless steel plates.
inches center-to-center rey inen tnick, Ihe plates, which are also are welded to the sides of the square storage tute four elevations.
1/b at Ine tuces are rlsred at he com atible witn the fuel handling equipeasy storage s
e semblies, and to rack arrays are utilized to maximize use or the Two different ment.
storage space in tne pool.
Ihe racx structure available tuel elevated case which is a system or welded teams is welded to an Tht: Lase serves to support ar.o stift eners.
to alstrinate the loao on tnethe weight of the tual assemolics ar d pool floor.
opening at tne tuel assemoly lower nozzle.each fuel assembly storage location wnich aTh ccommodates watel down between the storage tuoesNatural circulation of pool cells are designeo -to provice lateral supportnozzle or ea The storage assemolleu or the Westinghouse 17 x 17 for stored as emnlles wita the same external aimensioarray design, ano otner nozzle design.
ns ano similar lower To prevent movement of the racks, particularly f citation,'the racks are bolted to embedment plates in th rom seismic ex-of the spent fuel pool.
adjustable legs to allow natural circulation flowThe racks are raised of f th e floor floor on adjusted after installation to level the racks and dist The legs are load to all legs.
embedment nlate details and plan. Refer to Figures 9.1-20 and 9.1-21 for ribute the The spent initial core fuel with up to 3.20 w/o U-235 enri hfuel storage r checkerboard pattern only, ge of c ment in a Subsection 9.1.2.3.The safety evaluation for this mode of fuel stoprior to the rage is given in ty lgd Fu al_ Jtorage P acks
(
k uel assena lies that release ' radioactivity areare damaged or tailed such tnat they may stored in special containers in the talled fuel racx.
tuel pool and provides stolage for 10 talleo tu line tailed tuel stol spor.t dssemtlies.
This racx is snown in Figure 9.1-4 e
1/4 inen tnick stainless steel tuues spaced at Its consists of to-center, and supported by stainless steel pl t22 inches center-lengths.
ihe rack structure es along their a
is welded to an elevated case to aupport th(
cas+ contains an opening atweight or the tuel assemolies on tne pool tlool it.e each storage locatior. to allow natural circolation of pool water ap through this opening to remove decay heat trom tne stored assenoly.
To prevent movement or tr e rac4s, aosi.
delsmic excitattor., this rock is alav tolted to end,edment plates particularl y due to i t. the rloor or the spent t ue l pool.
It.e rack also has ad astacle legs to level the lacks 3
tloor to and raise the lacxs ott the allow not aral circalat ion r iaw.
9.1-6
[
J
r B/b-FSAR AMENDMENT 43 SEPTEMBER 1983 Spent Fuel Storage Pool Ihe spent fuel storage pool is designed f or the underwater storage of spent tuel ass emolies and control rods af ter their removal f rom the reactor.
The pool depth is such that the surf ace dose level will ce approximately 2. 5 mrem /hr when moving a fuel assemoly over storage racks.
It is designed to accommodate a total of 1050 tuel assemolies.
The spent fuel pool general arrangements are shown in Figures 1.2-9 and 1. 2-11.
Seent f uel assemolies are handled oy a fuel handling tool suspended from an overhead monorail electric hoist and manipulated by an operator standing on a movable bridge over the pool.
The spent fuel storage pool and transfer canal are constructed of structural reinforced concrete and lined with stainless steel.
A new ruel elevator is located in the pool to allow transferring new f uel assemblies into the pool for subsequent handling with the spent f uel pool handling crane.
The elevator is equipped with alarms to inrorm the operator of malf unction of the elevator during movement of new fuel.
The elevator carriage is always celow tne level of water in the pool.
All parts in contact with tne pool water are stainless steel.
This elevator is a
structurally designed as Seismic Category II-I A se pa ra te, walled-oft area is provided at the end of the pool for storing spent fuel ctsks.
The cask storage area is provided with leaktight gates to allow isolation f rom the pool.
The design of the fuel handling ouilding crane rails precludes travel of the crane's hook over the spent fuel storace racks except for storage of fuel assemblies for initial core loading.
For operations following initial core loading, the only lifting device which can travel over the spent fuel racha is the suent fuel pit bridge crane hoist.
This hoist is equipped with a load-limiting interlock to prevent lif ting a load heavier than l
4000 pounds.
If the load is in excess of this weight, the electrical interlock will stop the fuel bridge winch drive fron moving upward.
Fuel assemblies received for initial core loading may be inserted directly into the spent fuel storage racks'u' sing the new fuel handling tool and the fuel handling building crane..
This method of fuel storage will require an override of the fuel handling l
building crane interlocks and would occur only for receipt and dry storage of first core fuel.
l The design requirements for the separating wall between the fuel cas4 pit and the spent fuel pool are:
l a.
Walls are designed to withstand increased water I
pressure wnich may De caused oy a vertical drop of the cask.
The wall thickness is 5 feet.
l l
l 9.1-7 I
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B/B-FSAR AMENDMENT 43 SEPTEMBER 1983 t,.
For a cask drop on the exterior pool. wall, the wing sall will be allowed to deform locally ceyond its elastic limit, but it has oeen shown that the cask will not fall outside the cask storage well-and thus will not affect the fuel in the fuel storage pool.
I I
9.1-7a
B/B-FSAR AMENDMENT 46 JANUARY 1985 Since the cask will always be. brought to the pool so c.
as to enter the storage well, without passing over the pool, there is no possibility of cask drop into the fuel storage pool.
d.
The spent fuel pool floor is 6-foot thick reinforced concrete resting directly on bedrock (Byron) or on soil (Braidwood) and is designed for a vertical drop of the cask in the storage well.
e.
Gates are provided to transfer f uel from the pool to the spent fuel cask pit and to the fuel transfer canal.
9.1.2.3 Safety Evaluation Criticality Control The design of the spent fuel storage racks provides for a suocritical effective multiplication factor (keff) for both normal and abnormal storage conditions less than 0.95 in accordance with ANSI N210-1976.
Normal conditions exist when the fuel racks are covered with a normal depth of water (at least 23 feet above the top of stored fuel) for radiation shielding, and with the maximum number of fuel assemblies in their design storage position.
An abnormal condition may result from accidental dropping of a fuel assemoly or horizontal movement of a fuel assembly to a location adjacent to a loaded storage rack.
The criticality analyses of the normal condition included several conservative assumptions as well as the. effects of uncertainties in the calculation method, and geometric and material variations of the fuel storage rack.
The following conservative assumptions were used in the calculation:
a.
Fresh fuel of 4.00 w/o U-235 enrichment - Initially the maximum enrichment will be ' lower than this, but it could approach this enrichment if an 18-month fuel cycle is used.
The enrichment selected is higher j
than the average enrichment of any fuel expected to 1
be stored in the spent fuel pool for conservatism.
b.
Spent f uel pool bulk water temperature - A calculation was done to determine the increase in reactivity due to a further decrease in the design-basis temperature of 680 F and this reactivity was t
added to the nominal keff for still further conservatism.
c.
The neutron leakages in all three directions were neglected.
d.
Fixed neutron absorbers in the fuel assembly and soluble poison in the pool water are neglected.
9.1-8 L-
B/B-FSAR AMENDMENT 38 MAY 1982 The majority of the calculations were performed with methods commonly used in light water reactor design, i.e.,
four group diffusion theory with all calculations using PDQ-07.
Cross sections for these calculations are generated with NUMICE, the NUS Corporation versi'on of the Westinghouse LEOPARD code.
This code uses the same cross section library tape and calculation techniques as LEOPARD.
Selected cases were checked and the final design multiplication f actors were verified with Monte Carlo calculations using KENO with a 123-group cross section library generated from a basic GAM-THERMOS library using the NITAWL routine in the AMPX l
code package.
Both the PDQ-07 and the KENO calculation methods as described above have been benchmarked.
Under normal conditions with full storage racks, the k as determined from KENO is 0.9085.
Calculationuncert$kkties l
were determined from both comparison between calculation and experiments using KENO, and a statistical evaluation of Monte Carlo runs.
The results are a calculation uncertainty for the former of 0.0086 ak and, for the latter, 0.0087 ak, l
at a 95% confidence level; or a total calculation uncertainty of 0.0173 ak.
Variation in material composition and mechanical l
spacing and tolerances acting in a direction to close the water gaps between adjacent racks were also analyzed and resulted in a reactivity increase of 0.0134 Ak.
As previously l
stated, the effect of a temperature decrease was also included.
A temperature increase to 170' F resulted in a reactivity increase of 0.0024 ok.
Adding reactivity effects of calculation uncertainties of 0.0173 ak, geometric and material uncertainties of 0.0134 ak, and the temperature decrease effect of 0.0024 ak to the nominal k results in a maximum k of 0.942 l
with a 95% proba8[$ity at a 95% confidence Ibbel, which l
satisfies the design basis of maintaining k below the eff l
design-basis limit of 0.95.
Two abnormal conditions were also considered.
They are (1) a dropped fuel assembly assumed to lay across the top of fuel rack and (2) a fuel assembly in transport in a vertical position accidentally dropped into a position parallel with the stored fuel in the most reactive corner among the racks.
Of these two, the second is most severe, as the end fittings on top of the fuel assembly prohibits the dropped fuel assembly l
on top of the rack from being nearly as close to the stored fuel as the assembly beside the rack.
The analytical techniques j
and assumptions used for the analyses of the second case are the same as used for the analysis of the normal condition r
except that credit was taken for a minimum 1950 ppm soluble boron to bring the initial conditions into the range of l
credibility.
The analysis is still concervative as the l
conservatism for water temperature reduction was not removed, even though credible water temperature is well above 32' F.
9.1-9
l i
i B/B-FSAR AMENDMENT 38 MAY 1982 Thus, the I
The results show a nominal k worst accident case as analyI$$,well below 0.762.
does not have consequence any more than the worst normal distribution case.
Dry Storage of New Fuel in Spent Tuel Racks The design of the spent fuel storage racks also allows a subcritical effective multiplication factor for dry storage of new fuel with 3.20 w/o enrichment in a checkerboard loading pattern: k less than 0.98 for the optimum moderation conditionIk$k less than 0.95 for the flooded condition inaccordancewTkbANSIN210-1976.
Under the checkerboard loading pattern, new fuel assemblies are to be placed in rack positions with four nearest adjacent positions empty.
The optimum moderation condition exists when the storage racks are covered with clean water at a uniform density of 0.05 gm/cc.
The flooded condition exists when the storage racks are covered with cold, clean, unborated water at a i
full density.
The criticality analyses of these two conditions were performed with the assumptions and calculational methods similar to those utilized for the analyses of new fuel racks in Subsection 9.1.1.3.
The only exception was that the spent fuel storage rack was assumed to be infinite in size in all three dimensions for additional conservatism.
Under the optimum moderation condition of 0.05 gm cc water for the spent fuel racks containing the 3.20 w/o enrichment
[
fuel in the checkerboard loading pattern, the k as deter-j mined is 0.8904.
Addingreactivityeffectsof8$$culational uncertainties of 0.0153 Ak and geometric and material uncer-(
taintiesof0.0479Aktotheoptimummoderationk*kka95%
results in a maximum k of 0.954 with a 95% probability confidencelev$kfwhichsatisfiesthedesignbasisofmain-taining k,gg below the design-basis limit of 0.98.
When the spent fuel racks containing the 3.20 w/o enrichment fuel in the checkerboard loading pattern are flooded with cold, unborated water, the nominal k is 0.8413.
Adding reactivity effects of calculation un8khtainties of 0.0198 ok, l
geometric and material uncertainties of 0.0056 ok, and the temperature decrease effect (from 68* F to 40' F) of 0.0014 ok to the nominal k results in a maximum k of0.868witha95%probabTkktyata95%confidenceIkhel, which satisfies the design basis of maintaining k,gg below the design-basis limit of 0.95.
l 9.1-9a
B/B-FSAR AMENDMENT 19 MARCH 1979 Fuel Storage Rack Structural Design In accordance with the spent fuel storage f acility design bases and the requirements for seismic design classification, the spent and failed fuel storage racks are designed as Seismic Category I.
Structural integrity of the racks when subjected to normal, abnormal, and seismic loads was demonstrated.
Consequently, the following loads, loac combinations, and structural acceptance criteria are considered:
a.
Dead loads - the load due to dead weight of the rack, f uel assemblies, and buoyancy.
b.
Live loads - the load due to lif ting empty racks during installation.
c.
Thermal loads - the load due to unif orm thermal expansion of the racks caused by a change in average pool temperature from 400 F to 1500 F, a thermal gradient between adjacent storage locations of 180 F for spent fuel storage racks, and 220 I for failed f uel storage racks.
l
~
d.
Severe environmental load - the load due to the operating basis earthquake (OBE).
e.
Extreme environmental load - the load due to the sate shutdown earthquake (SSE).
I f.
Accidental drop of a tuel assembly from the maximum possible height consistent with f uel handling operations.
g.
A postulated stuck fuel assemoly causing an upward I
force equal to tne fuel grapple load limit exerted on the assembly during attempted withdrawal.
The f uel storage racks were analyzed using the elastic working stress oesign methods for tne following load combinations:
a.
dead loads plus live load, b.
cead loads plus OBE loacs, c.
dead loads elus normal tnermal loads plus OBE loads, d.
dead loads plus normal thermal loads plus SSE loads, e.
dead loads plus normal thermal loads plus fuel assembly drop, and 9.1-10
/
[
B/B-FSAR AMENDMENT 32 AUGUST 1981 f.
dead loads plus normal thermal loads plus stuck fuel assembly load.
Live loads are not included in load combinations b through f, since the only live load on the rack occurs during lifting.
Lifting of the racks is performed only when the racks are ampty.-
The following strength limits apply for the above load combinations:
Load Combination Strength Limit a
1.0S b
1.0S c
1.5S d
1.6S e
1.6S (except as noted below) f l.6S (except as noted below)
Where S is the required section strength based on the elastic design methods and the allowable stresses defined in Part 1 of the AISC " Specification for the Design, l
Fabrication and Erection of' Structural Steel for Buildings,"
February 12, 1969, including Supplement Numbers 1, 2, and 3.
(Supplement 3 was effective, June 12, 1974, and revised June 30, 1975.)
For load combinations e and f, local stress-es might exceed the limits, provided there is no loss of function of the fuel rack.
Both the spent fuel and failed fuel storage racks were ana-lyzed to determine 'that these strength litd is are not exceeded, a.
Spent Fuel Storage Racks The seismic loading of a fuel rack module is l
determined from a response spectrum modal dynamic analysis in which the stiffness of the fuel as-sembly is neglected.
However, the mass of the fuel assemblies and an effective mass of water are considered to be uniformly distributed along the storage tubes.
The appropriate floor response spectra and damping values (2% and 4%) for the l
l OBE and DBE, respectively, are employed.
The STARDYNE computer program is used to perform the structural analysis of the racks.
Racks are modeled in detail using beam and plate finite 9.1-11 J
v-B/B-FSAR AMENDMENT 19 MARCH 1979 elements.
The three-dimensional finite element Model for a spent fuel rack is shown in Figure 9.1-5.
To determine the earthquake response, STARDYNE is first run to determine the natural frequencies and participation factors.
For frequencies with significant modal participation, mode shapes, and t
I 9.1-11a
P".
B/B-FSAR modal loads are calculated.
Closely spaced modes are combined directly and then combined in an SRSS manner with other significant modes.
The results of the three directions of earthquake are combined in an SRSS fashion per requirements for combining modal responses and spatial components in seismic response analysis.
In tne general seismic / structural analysis of the fuel racks, the mass of a f uel assembly is assumed to be uniformly distributed along the length of each of the fuel storage cans.
This assumption is conservative because the lower racx fundamental frequencies are calculated using a relatively stiff' rack design, resulting in a higher seismic amplified acceleration loading on the rack.
Since a gap of approximately 1/4 inch exists between the sides of a fuel assembly and the can, the fuel will actually move within the can during a seismic event and cause impact loads to be transmitted to the fuel rack.
The eff ects of this fuel-can interaction were analyzed by utilizing the ANSYS computer program.
A nonlinear dynamic analysis of a single can and fuel assembly was performed to determine the shear force and bending moment which may occur at critical sections of the can as a result of the fuel assembly impacting the can at the maximum velocity.
The can and fuel assembly were modeled by finite elements separated oy nonlinear gap elements as shown in Figure 9.1-6.
The can has stiffness characteristics representative of a can within a rack.
The fuel, which was assumed to be pinned at its base (by friction), was given an initial velocity relative to the can.
This initial velocity is equal to the SRSS summation of the floor velocity and the velocity of the ~ rack with respect to the floor.
b.
Failed Fuel Racks l
The seismic loading of the failed fuel rack was determined from a response spectrum modal dynamic analysis in which the stiffness of the fuel assembly and f ailed fuel container was neglected.
- Again, however, the mass of the f ailed f uel container and its contents and an effective mass of water are considered to be uniformly distributed along the storage tubes.
The appropriate response spectra for the OBE and SSE were employed.
The STARDYNE computer program was used to perform the structural analysis of the racks.
Failed fuel racks were modeled in l
detail using beam and plate elements.
The three 9.1-12 I
l L
f' B/B-FSAR AMENDMENT 19 MARCH 1979 dimensional finite element model for the failed fuel rack is shown in Figure 9.1-7.
As for the spent fuel storage racks, to determine the earthquake response, STARDYNE was first run to determine the natural fre uencies and participation 3
factors.
For frequencies with significant modal participation, mode shapes and modal loads were calculated.
Closely spaced modes were combined directly and then combined in an SRSS manner with other significant modes.
2he results of the three directions of earthquake was combined in an SESS fashion per requirements for combining modal responses and spatial components in seismic response analysis.
In the seismic and structural analysis of the fuel racks, the mass of a f ailed fuel container and its contents was assumed to be unif ormly distributed along the length of each of the f uel storage cans.
This assumption is conservative because the lower
~
rack fundamental frequencies calculated loading on the racx due to the relatively stiff rack design, result in higher seismic amplified acceleration.
A gap on the order of 1/4-inen exists between the sides of a failed fuel container and the can, thus the container will actually move within the can during a seismic event and caase impact loads to ce transmitted to the racx.
The effects of this fuel-Cdn interaction were considered Dy conservatively doacling the results of the frequency response analysis to account for the effects of impact.
Using the previously listed loads and load combf. nations, stresses are calculated at critical sections of the racks.
The results of the structural and seismic analyses demonstrate that the fuel racks are structurally adequate and meet the design criteria.
Since rack integrity is maintained, no damage to stored assemblies and no increase in K,ff would occur under these loads.
Fuel Rack Design Features Both the spent and failed fuel storage racks are all stainless steel, as is the spent fuel pool liner, to minimize the potential for galvanic corrosion.
Stainless steel has also been shown to be compatible with spent fuel pool water and the stored assemblies.
The fuel rack base is elevated above the floor to assure adequate flow under the rack in each fuel assembly.
Analyses have been performed and show that sufficient flow is induced by natural convection to preclude local boiling in the hottest storage location.
l The analyses were based on the following assumptions.
9.1-13
B/B-FSAR AMENDMENT 43 SEPTEMBER 1983 a.
The fuel element inlet design temperature of the pool. temperature is the mixed hot b.
A hot assembly peaking factor of 1.55 is applied to the core average assembly energy release rate of 2.3 x 105 Btu /hr.
The maximum local peaking factor is 2.33, giving a c.
maximum local neat rlux of 1706 Btu (hr-ft2).
d.
A film coefficient of 39 Btu (hr-ft2-0 F) is cased on pure conduction through a stagnant boundary layer at the fuel rod surface.
A downcomer region between adjacent cans within the e.
be generating the maximum heat rate defined inrack assumption b.
f.
One dimensional fluid flow analysis applies design value ofDuring rull core offload with the culx pool temperatur 1500 F, e at a the mixed temperature of the water exiting f rom the hottest storage location is less than 1860 Ihis is 540 inaica ting that adequate marF below the local saturation temperature of 2400 F.
F design operating conditions, gin to culk boiling exists.
Under calculated on the basis of the heat flux and film coefficiertthe fuel rod su defined above, is 110 and thus precludes local coiling.F below the local saturation temperature Other dpent Fuel Storage Facility Design Features The Byron /Braidwood stainless steel spent fuel pool liners w originally designed and constructed to Category I requirements ere Weld inspections conducted subsequent to the initial accepta of the Byron liner did not confirm the acceptability of thosence welds to Category I standards.
Because they meet the additional criteria of NUREG-0800 Revision 3, July 19 81, the liners for the spent fuel pool, transfer canal, and spent fuel cask pit do not need to be desi fuel and erected to Category I requirements.
gned terms of reduced record keeping and surveillance requirementsLong term benefits in also result from this action without degrading safety.
The design of the liner is such that the design th are much larger than the design seismic ermal stresses of this fact, and because the liner was designed and construct d (SSE) stresses.
Because as a Category I structure, gross failure of the liner as a e
result of a seismic event is not considered to be credibl Thus, neither mechanical damage to fuel nor flow bl e.
of cpent fuel in the spent fuel racks as a result of line ockage failure are considered to be credible events r
9.1-14 L
AMENDMENT 43 B/B-FSAR SEPTEMBER 1983 A failure of the liner which allows leakage of the pool water past the liner (such as the rupture of a seam) as a result of a seismic event would not have any adverse effects on fuel in the pool or on any safety-related equipment in the plant.
Five 1 inch drains are provided behind the spent fuel pool liner.
The drain piping is embedded in the concrete structure up to column-row W, where it joins the auxiliary building floor drain system.
At the point where the drain piping emerges from the concrete wall, five normally closed valves and sight glasses (one per drain line) are provided.
Any leakage past the liner would be collected by the drain piping and stopped by the valves.
No other leakage paths exist.
The auxiliary building floor drain piping is seismically supported, so failure of this piping as a result of an SSE is not postulated.
- Thus, mailure of the liner would not result in loss of significant quantities of pool water, and no safety-related equipment would be affected by any resulting leakage.
The activity release from a dropped or ruptured fuel assembly shall be limited such that the radiation dose at the site boundary will not exceed the limits stated in 10 CFR 100.
Activity releases in the event of fuel damage in the spent fuel pool would be limited by the decontamination factor provided by the pool water.
The ventilation system provided in the fuel handling building is equipped with filtration devices to limit the potential release of radioactive materials including iodine.
The design is described in detail in Subsection 9.4.2.
1 Ihe fuel storage facility is contained and the equipment is l
designed so that accidental releases of radioactivity are monitored and will not exceed the guidelines of 10 CFR 100.
Subsection 15.7.4 contains analyses of hypothetical fuel handling accidents.
l The fuel racxs are designed and fabricated with a high degree of l
reliability and integrity.
A list of codes and standards used f or the spent and failed fuel storage racks is given celow:
1 1
9.1-14a
B/B-FS AR a.
Design Codes 1.
AISC Manual of Steel Construction, 7th Edition, 1970; and 2.
AISC " Specification for the Design, Fabrication and Erection of Structural Steel for Buildings,"
February 12, 1969 and Supplements 1, 2,.and 3.
(Supplement 3 effective June 12, 1974).
3.
ANSI N18. 2-19 73, Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants.
4.
ANSI N210-1976, LWR' Spent Fuel Facilities.
b.
Material Codes 1.
ASME Specification SA-240, Specification for Stainless and Heat Resisting Chromium and chromium-Nickel Steel Plate Sheet and Strip for l
Fusion-Welded Unfired Pressure Vessels; 2.
ASME Specification SA-230 specification for Alloy Steel Bolting Materials for Low Temperature Service; and 3.
ASME Specification SFA-5.9, Corrosion Resisting Chromium and Chromium Nickel Steel Welding Rods and Base Electrodes.
c c.
Welding Codes 1.
ASME Boiler and Pressure Vessel' Code Section IX-1974, Welding and Brazing Qualifications.
d.
Quality Assurance, Cleanliness, and Packaging Requirements 1.
Commonwealth Edison Company and/or other approved quality assurance requirements; 2.
ANSI N45.2.2, Packaging, Shipping, Receiving, Storage and Handling of Items for Nuclear Power Plants; and 3.
ANSI N45. 2.11, Quality Assurance Requirements for the Design of Nuclear Power Plants.
9.1.3 Spent Fuel Pool Cooling and Cleanup System Ihe spent fuel pool cooling system (SFPCS) is designed to remove the decay heat generated by stored spent fuel assemolies from the spent f uel pool water.
This cooling is accomplished by taking 9.1-15
... ~. - -
i B/B-FSAR AMENDMENT 43 SEPTEMBER 1983 9.1.4 Fuel Handling System 9.1.4.1 Design Bases Tne Fuel Handling System (FHS) consists of equipment and structures utilized f or the refueling operation in a safe manner, meeting General Design Criteria 61 and 62 of 10_CFR 50, Appendix A.
The following design bases apply to the FHS:
Fuel handling devices have provisions to avoid a.
dropping or jamming of fuel assemblies during transfer operation.
c.
nandling equipment has provisions to avoid dropping of fuel handling devices during the fuel transfer operation.
Handling equipment used to raise and lower spent fuel c.
have a limited maximum lift height so that the minimum required depth of water shielding is maintained.
d.
Criticality during fuel handling operations is prevented by geometrically safe configuration of the fuel handling equipment.
Handling equipment will not fail in such a manner as e.
to damage Seismic Category I equipment in the event of a safe shutdown earthquake.
f.- The inertial loads imparted to the fuel assemblies or core components during handling operations are less than the loads which could cause damage.
Physical safety features are provided for personnel g.
operating handling equipment.
9.1.4.2
System Description
The FHS consists of the equipment needed for the refueling operation on the reactor core.
This equipment is comprised of a tuel assemoly, core and reactor component hoisting equipment, handling equipment and a Fuel Transfer System (FTS).
The structures associated with the fuel handling equipment are the refueling cavity, the refueling canal, and the fuel storage area.
New fuel assemblies received for initial core loading are removed one at a time from the shipping container and stored in the new fuel vault; or are lowered into the spent fuel storage area by the new fuel elevator and stored in the spent fuel storage racks; or 9.1-28
's
f-B/B-FSAR AMENDMENT 43 SEPTEMBER 1983 fuel assemblies received for initial core loading may also be inserted directly into the spent fuel storage racks usin fuel handling tool and the fuel handling building crane.g the new-This method of fuel storage will require an override of the handling building crane interlocks and would occur only for receipt fuel and dry storage of first core fuel.
3 9.1-28a l
_J
r i
B/B-FSAR AMENDMENT 43 L
SEPTEMBER 1983 1
l The f uel handling equipment is designed to handle a spent fuel assembly underwater from the time it leaves the reactor vessel until it is ready for placement in a container for shipment from the site.
Underwater transrer of spent fuel assemblies provides an effective, economic, and transparent radiation shield, as well as a reliable cooling medium for removal of decay heat.
The boric acid concentration in the water is sufficient to preclude criticality.
Ihe associated fuel handling structures may be generally divided into two areas:
the refueling cavity and refueling canal which are flooded only during plant shutdown for refueling, and the spent refueling cavity, which is kept f ull of water and is always accessiole to operating personnel.
The refueling canal and the refueling cavity are connected by a fuel transfer tube.
This tune is titted with a gate valve on the fuel storage area end.
Fuel is carried through the tube on an underwater transfer car.
Fuel is moved between the reactor vessel and the refueling canal by the refueling machine.
The RCC Change Fixture is used for transferring control elements from one fuel assembly to another fuel assembly.
The FTS is used to move fuel assemblies between the containment ouilding and the_ fuel storage building.
After a f uel assemnly is placed in the fuel container, the lifting arm pivots the fuel assembly to the horizontal position for passage tnrough the fuel transfer tube.
After the transfer car transports the fuel assembly through the transfer tube, the lif ting arm at that end of the tune pivots the assembly to a vertical position so that the assembly can be lifted out of the fuel container.
In the fuel storage building, fuel assemblies are moved about by the spent fuel pit bridge crane and the fuel handling building crane.
When lif ting fuel assemblies with the spent fuel pit bridge crane, the hoist uses a long-handled tool to assure that sufficient radiation shielding is maintained.
A shorter tool is used with the fuel handling building crane to handle new fuel assemblies initially, and for operations following receipt of assemblies for initial core loading, the new fuel elevator must be used to lower the assembly to a depth at which the spent fuel pit bridge crane, using the long-handled tool, can place the new fuel assemblies into or out of the fuel storage racks.
9.1.4.2.1 Re fueling Procedure Ihe refueling operation follows a detailed procedure which provides a safe, efficient refueling operation.
Prior to initiating refueling operations, the reactor coolant system is borated and cooled down to refueling shutdown conditions as specified in the technical specifications.
Criticality protection for refueling operations, including a requirement for daily checks of boron concentrations, is specified in the technical specifications.
The following significant points are assured by the refueling procedure:
9.1-29
l B/B-FSAR AMENDMENT 32 AUGUST 1981 The refueling water and the reactor coolant contains a.
approximately 2000 ppm boron.
together with the negative reactivity of controlThis concentration,
- rods, subcritical during the refueling operations.is sufficien l
It is also sufficient to maintain the core cluster control assemblies were removed from th core.
b.
The water level in the refueling cavity is high limits when the fuel assemblies are being removed from the core.
Ihe refueling operation is divided into four major phase prepa ration, (2) reactor disassembly, (3) fuel handling, and s:
(1)
(4) reactor assembly.
refueling operation through the four phases ~ follows:A general descrip 1.
Phase I - Preparation The reactor is shut down and cooled to cold shutdown co di as specified in the technical specifications.
n t ons down or later, following a radiation survey During cold shut-entered.
is lowered to a point slightly below the vessel flangeAt t i
sel transfer equipment and refueling machine are then The fuel proper operation.
checked for 2.
Phase II - Eeactor Disassemoly All caoles and insulation are removed from the vessel h refueling cavity is then prepared for flooding by sealing off th ead.
The l
reactor cavity; checking the underwater lights, tools, and FTS; e
closing the refueling canal drain holes; and removing the blind I
flange to create access to the fuel transfer tube i
head assembly is unseated and raised above the vess With the Water from the refueling water storage tank is pumped i ge.
causing the water to overflow into tne refueling cavity. r nto the vessel head and the water level in the refueling cavity areThe raised simultaneously, keeping the water level just below the head.
When the water reaches a safe shielding depth Subsection 9.1. 4. 3. 4),
(see taken to its storage pedestal.the integrated vessel head assembly is disconnected and, with the upper internals, are removed fThe control rod d vessel.
are now free from obstructions and the core is ready forThe rom the refueling.
9.1-30
B/B-FSAR 3.
Phase III - Fuel Handling The refueling sequence is started with the refueling machine.
The positions of partially spent assemblies are changed, and new assemblies are added to the core.
The general f uel handling sequence is as follows:
a.
The refueling machine is positioned over a fuel ass embly.
b.
The fuel assembly is lifted by the refueling machine to a predetermined height sufficient to clear the reactor vessel and still leave sufficient water covering to eliminate any radiation hazard to the operating personnel.
If the assembly to be removed contains a rod cluster c.
control unit, the rod cluster control is removed from the spent fuel assembly and put in a new fuel assembly or in a partially spent fuel assembly by the RCC change fixture.
d.
The fuel transfer car is moved into the refueling cavity from the fuel transfer canal.
e.
The fuel container is pivoted to the vertical position by the lifting arm.
I f.
The refueling machine is moved to line up the fuel assembly with the FTS.
g.
The refueling machine loads the fuel assembly into the fuel container of the transfer car.
h.
The container is pivoted to the horizontal position oy the lifting arm.
i.
The container is moved through the fuel transfer tube to the fuel transfer canal by the transfer car.
j.
The fuel container is pivoted to the vertical position.
k.
The spent fuel assembly is unloaded by the spent fuel handling tool attached to the spent fuel pit bridge crane.
1.
The spent fuel assembly is placed in the spent fuel storage rack.
m.
The fuel container is pivoted to the horizontal position and the transf er car is moved back into the refueling cavity.
9.1-31
B/B-FSAR AMENDMENT 43 SEPTEMBER 1983 n.
Partially spent fuel assemblies are relocated in the reactor core, and new fuel assemblies are added to the core.
o.
This procedure is continued until refueling is completed.
4.
Phase IV - Feactor Assembly Reactor assembly following refueling is achieved by reversing the operations given in Phase II, Reactor Disassembly.
During reassembly, the vessel head and water are lowered simultaneously until the vessel head engages the guide studs.
At this point of reassembly the water is lowered to the top of the reactor vessel flange.
This allows visual observation of drive rod insertion into the piped locations of the vessel head.
9.1.4.2.2 component Description Refueling Machine The refueling machine (Figure 9.1-12) is a rectilinear bridge and trolley system with a vertical mast extending down into the refueling water.
The bridge spans the refueling cavity and runs on rails set into the edge of the refueling cavity.
The bridge and trolley motions are used to position the vertical mast over a fuel assembly in the core.
A long tube with a pneumatic gripper on the end is lowered down out of the mast to grip _the fuel assembly.
The gripper tube is long enough so that the upper end is still contained in the mast when the gripper end contacts the fuel.
A winch mounted on the trolley raises the gripper tube and fuel assembly up into the mast tube.
The fuel is transported while inside the mast tube to its new position.
All controls for the refueling machine are mounted in a console on the trolley.
The bridge is positioned on a coordinate system laid out on one rail.
An electrical readout system on the l
console indicates the position of the bridge.
The trolley is positioned with the aid of a scale on the bridge structure which l
I is read directly by the operator at the console.
The drives for the bridge, trolley, and winch are variable speed and include a separate inching control on the winch.
The maximum speeds are 40 fpm for the bridge and 20 fpm for the trolley and hoist.
The l
auxiliary monorail hoist on the refueling machine has a 2-step magnetic controller to give hoisting speeds of approximately 7 fpm and 20 fpm.
Electrical interlocks and limit switches on the bridge and trolley drives prevent damage to the fuel assemblies.
The winch is also provided with redundant limit switches to prevent a fuel assembly from being raised above a safe shielding depth should one limit switch fail.
In an emergency, the bridge, the trolley, 9.1-32
B/B-FSAR AMENDMENT 18
' JANUARY 1979 and the winch can be operated manually using a handwheel on the motor shaft.
1 i
9.1-32a f
,,.m,-
- - ~ ' - - ' - ~ ~ - ~ ' ~ "
l' B/B-FSAR AMENDMENT 45 JUNE 1984 Scent Fuel Pit Bridge Crane l
The spent fuel pit bridge crsne (Figure 9.1-13) is a wheel-
~
mounted walkway which spans the spent fuel storage area and fuel transfer canal and it carries an electric monorail hoist on an overhead structure.
Following initial core loading the spent fuel j
pit bridge crane is used exclusively for handling fuel assemblies i
within the spent fuel storage area by means of a long-handled tool suspended from the hoist.
The hoist travel and tool length are designed to limit the maximum lift of a fuel assembly to a safe shielding depth.
. The spent f uel pit bridge crane has a 2-step magnetic conrroller for the bridge and hoist.
Bridge speeds are approximately 12 fpm and 25 fpm with inching control, and hoist speeds are approximately 7 fpm and 20 fpm.
A hydraulic coupling is used in the bridge drive to limit starting acceleration.
)
The hoist trolley is hand operated oy a chain drive.
l l
New Fuel Elevator The new f uel elevator (Figure 9.1-14) consists of a box-shaped elevator assembly with its top end open and sized to house one fuel assembly.
Following receipt of fuel assemblies for initial core loading, the new fuel elevator is used exclusively to lower a new fuel assembly to the bottom of the fuel storage area, where it is transported to the storage racks by the spent fuel pit l
bridge crane.
l l
Fuel Transf er System l
The fuel transfer system (PPS) (Figure 9.1-15) includes an i
underwater, electric-motor-driven transfer car that runs on tracks extending from the refueling cavity through the transfer l
tube and into the refueling canal.
There is a hydraulically actuated lifting arm at each end of the transfer tuce.
The fuel container in the refueling cavity receives a spent fuel assembly in the vertical position from the refueling machine.
The spent f uel assembly is then lowered to a horizontal position for
- passage through the transfer tube.
Af ter' passing through the l
tube, the spent fuel assembly is raised to a vertical position and removed from the container by a tool suspended from a hoist I
I mounted on the spent fuel pit bridge crane.
The spent fuel pit bridge crane then moves to a storage loading position and places i
the spent fuel assembly in the spent f uel storage racks.
i l
During reactor operation, the transfer car is stored in the refueling canal.
A blind flange is bolted to a tube surrounding i
i the transfer tube on the containment end to seal the reactor
{
containment.
The terminus of the tune outside the containment is closed oy a gate valve.
l r
9.1-33
p B/B-FSAR AMENDMENT 43 SEPTEMBER 1983 I
Spent Fuel Assembly Handling Tool The spent fuel assemoly handling tool (Figure 9.1-16) is used to handle new and spent fuel assemblies in the fuel storage area.
It is a manually actuated tool suspended from the spent fuel pit bridge crane which uses four cam-actuated latching fingers to grip the underside of the fuel assemoly top nozzle.
The operating handle which actuates the fingers is located at the top of the tool.
When the fingers are latched, a pin is inserted into the operating handle which prevents the fingers from being accidentally unlatched during tuel handling operations.
New Fual Assembly Handling tocl The new f uel assembly handling tool (Figure 9.1-17) is used to litt and transfer fuel assemblies from the new fuel shipping containers to the fuel storage vaults or the new fuel elevator.
Fuel assemblies received for initial core loading may be inserted directly into the spent fuel storage racks using the new fuel handling tool and the fuel handling building crane.
This method of fuel storage will require an override of the fuel handling building crane interlocks and would occur only for receipt and dry storage of first core fuel.
The new fuel assembly handling tool is a manually actuated tool suspended from the fuel building crane and uses four cam-actuated latching fingers to grip the underside of the fuel assembly top nozzle.
The operating handles which actuate the fingers are located on the side of tool.
When the fingers are latched, the safety screw is turned in to prevent the accidental unlatching of the fingers.
Integral Reactor Vessel Head Assembly The integral upper head cooling shroud is a system which combines the head lifting rig, seismic platform, lift columns, reactor vessel missile shield, Control Rod Drive Mechanism (CRDM) forced air cooling system, and electrical and instrumentation cable routing into an efficient, one package reactor vessel head design.
Cooling Shroud Structure The cooling shroud structure provides support for the CRDM cooling system fans and the stud tensioner hoists.
Cooling air is directed through openings in the shroud, down along the mechanisms, back up the shroud through the CRDM cooling fans, and is finally exhausted upward into the containment atmosphere.
Four f ans are provided on the shroud to deliver the required flow.
Two f ans will provide the design flow rate, while the other two f ans are held in reserve as standby spares that are aVdilable to service the CRDM.
The shroud. structure is bolted to d sdpport ring on the reactor vessel head and is also attached to the three lif t columns.
The shroud also provides support for the CROM power and instrumentation (Reactor Protection Instrumentation and Thermocouple) cables.
Cables are routed from the mecnanisms to the cable tray platform which is attached to 9.1-34
~~
i B/B-FSAR AMENDMENT 43 SEPTEMBER 1983 I
the shroud.
Connectors are provided on the platform so that the caole tray with the cables may be easily removable.
Access is also provided through the shroud for use of a thermocouple.(T/C) column loading tool.
Missile shield The reactor vessel rissile shield is used to prevent any postulated missiles from the reactor vessel head appendages from penetrating other reactor coolant system pressure boundaries and/or containment structure.
In addition to this function, the missile shield also transfers the reactor vessel head load to the lifting rig.
The missile shield also provides seismic support for the CRDM's.
Attached to the three lift rods during plant operation, the missile shield has the ability to be properly levelea for the lift operation as well as to be easily detached from the lift rods to provide access to the control rod drive mechanisms.
Cable Tray The cable tray is a structure which is attached to -the cable tray platform on the cooling shroud and pivots on the steam generator wall or another appropriate support structure.
The cable tray serves to support the power and instrumentation cables from the cable tray platform to the terminal boxes.
It also provides a method of easily disassembling and storing the cables in preparation for head removal.
Stud Handling System By providing the capability of handling studs independent of the main polar crane, the stud handling system permits more efficient and smoother stud handling.
Studs and stud tensioners are handled by the hoists supported from a monorail on the shroud structure.
Radial travel of studs and stud tensioners is also provided through transfer beam assemblies to improve the flexibility of stud movement.
Cable support System l
To provide a support system for the CRDM power and
~
instrumentation (RPI and T/C) cables.
Cables are routed from the mechanisms and T/C columns to the cable tray to the te,rminal boxe s.
Connectors would be provided on the cables so that the cable tray with the cables may be easily removable as desired.
9.1-34a
i t
l l
B/B-FSAR AMENDMENT 18 JANUARY 1979 Reactor Vessel Head Assembiv Lifting Device The reactor vessel head-lifting device consists of a welded and I
bolted structural steel frame with suitable rigging to enable the crane operator to lift the head and store it during refueling operations.
The lifting device is permanently attached to the reactor vessel head.
Reactor Internals Lifting Device 9
The reactor internals lifting device (Figure 9.1-18) is a structural frame suspended from the overhead crane.
The frame is lowered onto the guide tube support plate of the internals and is mechanically connected to the support plate by three breech-lock-type connectors.
Bushings on the frame engage guide studs in the vessel flange to provide guidance during remov' l and replacement a
of the internals package.
Reactor Vessel Stud Tensioner The stud tensioners (Figure 9.1-19) are employed to secure the head closure joint at every refueling.
The stud tensioner is a hydraulically operated device that uses oil as the working fluid.
The device permits preloading and unloading of the reactor vessel closure studs at cold shutdown conditions.
Stud tensioners minimize the time required for the tensioning or unloading operation.
Tensioners are provided and are applied i
simultaneously to studs located 1200 apart.
A single hydraulic pumping unit operates the tensioners, which are hydraulically connected in series.
The studs are tensioned to their operational load in two steps to prevent high stresses in the flange region and unequal loadings in the studs.
Relief valves on each tensioner prevent overtensioning of the studs due to l
excessive pressure.
9.1.4.3 Safety Evaluation 9.1.4.3.1 Safe Handling Design Criteria for the Fuel Handling System 1.
The primary design requirement of the equipment is reliability.
A conservative design approach was used for all load-bearing parts.
Where possible, components were used that have a proven record of reliable service.
Throughout i
the design, consideration was given to the fact that the equipment will spend long idle periods stored in an atmosphere of 1200 F and high humidity.
l i
2.
Except as otherwise specified, the refueling machine and spent fuel pit bridge crane were designed and constructed in l
accordance with Crane Manuf acturer Association of America (CMAA) Specification No. 70.
t 9.1-35
[
B/B-FSAR AMENDMENT 45 JUNE 1984 3.
Design load for the refueling machine bridge crane shall be normal dead and liand spent fuel pit hoist load.
ve loads plus maximum 4.
The allowable stresses for the refueli fuel pit bridge crane structures supp ng machine and spent fuel, shall be as specified in Subarticlorting the weight of the Appendix XVII of the ASME Code Sectio e XVII-2200 of AISC Steel Construction Manual. n III, (1974), or i
Hoisting components leaded in simple t 1
allowable stress of 0.20 ultimate stress maxiensio are used:
each shall be assumed to carry one-half the l mum.
Two cables S.
All components critical to the operatio oad.
located so that parts can fall into then of the equipment or loosening under vibration. assembled with the fasteners posit reactor shall be rained from i
Fandlina EQuiccentIndustrial codes and Standards Used n the Desian of the Fuel 1.
Applicable sections of Crane ManufactuRefueling Machin r
ge Crane:
Specification No. 70.
rer Association of America 2.
Structural:
refueling machine),a. AISC Steel Construction Manual Section III, Appendix XVII,which is equivalent to ASME Code, (for the i
Edition; and b. ASME Code, Subarticle XVII-2000, Subarticle 2200 1974 Section III, Appendix XVII (for the spent fuel pit bridge cran ),
3.
Electrical:
e.
279, National Electric Code, Nation lApplicable standards a Association ents of IEEE Association - ((NEMA)
NFPA)
Fire Protection a
70, and National Electrical Manufact Standards MGI and ICS shall be used i design, installation, and manufactu i urers equipment.
n the r ng of all electrical 4.
Materials:
Materials conform to the specificati standards.
ons of ASTM S.
Safety:
subpart N of the Occupational SafetyThe applicable re and General Design Criteria 61 and 62 code, the (OSHA)
., Regulatory Guide 1.29 Refuelina Machine The refueling machine design includ ensure safe handling of fuel assembliees the following provisions to s:
I 9.1-36
___w-V W"
s B/B-ESAR AMENDMENT 45 JUNE 1984 a.
Electrical Interlocks
.1.
Bridge, Trolley and Hoist Drive Mutual Interlocks Bridge, trolley and winch drives are mutually interlocked using redundant interlocks to prevent simultaneous operation of any two drives and can therefore withstand a single failure.
2.
Bridqo Trolley Drive - Gripper Tube Up Bridge and trolley drive operation is prevented except when the gripper tube up position switches are actuated.
The interlock is redundant and can withstand a single f ailure.
3.
Gripper Interlock An interlock is supplied which prevents the opening of a solenoid valve in the air line to the gripper except when zero suspended weight is indicated by a force gauge.
As backup protection for this interlock, the mechanical weight-actuated lock in the gripper prevents operation of the gripper under load even if air pressure is applied to the operating cylinder.
This interlock is redundant and can withstand a single failure.
4.
Excessive Suspended Weight Two redundant excessive suspended weight switches i
open the hoist drive circuit in the up direction when the loading is in excess of 110% of the combined weight of a fuel assembly and the gripper mast.
The interlock is redundant and can withstand a single failure 5.
Hoist-Gripper Position Interlock An interlock in the hoist drive circuit in the up direction permits the hoist to be operated only when either the open or closed indicating switch en the gripper is actuated.
The hoist gripper position interlock consists of two separate circuits that work parallel such that one circuit must be closed for the hoist to operate.
If one or'both interlocking circuits fail in the closed position, an audible and visual alarm on the console is actuated.
The interlock is therefore not redundant but can withstand a single failure, since Doth an interlocking circuit and the monitoring circuit must fail to cause a hazardous condition.
l 9.1-37 l
I B/B-FSAR AMENDMENT 45 JUNE 1984 b.
Bridge and Trolley Hold-Down Devices Both refueling machine bridge and trolley are horizontally restrained on the rails by two pairs of
- i guide rollers, one pair at each wheel location on one truck only.
The rollers are attached to the bridge truck and contact the vertical -faces on either side j
of the rail to prevent horizontal movement.
Vertical restraint is accomplished by antirotation bars l
located at each of the four wheels for both the oridge and trolley.
The antirotation bars are bolted to the trucks and, for the bridge restraints,
'xtended under the rail flange.
For the trolley l
restraints they extend beneath the top flange of the bridge girder which supports the trolley rail.
Both horizontal and vertical restraints are adequately i
designed to withstand the forces and overturning l
moments resulting from the safe shutdown earthquake.
1 i
c.
Design Load The design load for structural components supporting the f uel assembly is the dead weight plus 4500 pounds l
(approximately 3 times the fuel assembly weight).
d.
Main Hoist Braking System The main hoists are equipped with two independent braking systems.
A solenoid release-springset electric brake is mounted on the motor shaft.
This j
Drake operates in the normal manner to release upon application of current to the motor and set when current is interrupted.
The second brake is a mechanically. actuated load brake internal to the hoist gear box that sets if the load starts to overhaul the hoist.
It is necessary to apply torque from the motor to raise or lower the load.
In i
raising, this motor came the brake open; in lowering, the motor slips the brake, allowing the load to lower.
This brake actuates upon loss of torque from the motor for any reason and is not dependent on any electrical circuits.
On the main hoist, the motor brake and the mechanical brake are rated at the l
capacity of the hoist.
e.
Fuel Assembly Support System l
The main hoist system is supplied with redundant l
paths of load support such that failure of any one l
component will not result in free fall of the fuel assembly.
Two wire ropes are anchored to the winch drum and carried over independent sheaves to a load-equilizing mechanism on the top of the gripper tube.
In addition, supports for the sheaves and equalizing 9.1-38 l
c.
B/B-FSAR AMENDMENT 45 JUNE 1984 mechanism are backed up by passive restraints to pick up the load in the event of failure of this primary support.
Each cable system is designed to support 13,750 pounds or 27,500 pounds acting together.
The working load of the fuel assembly plus the gripper is l
approximately 2500 pounds.
The gripper itself has four fingers gripping the fuel, any two of which will support the fuel assembly weight.
- The gripper mechanism contains a spring-actuated mechanical lock which prevents the gripper from opening unless the gripper is under a compressive load.
During each refueling outage and prior to removing fuel, the gripper and hoist system are routinely load tested to 3563 l
pounds.
j Fuel Transfer System The following safety features are provided for in the fuel l
transfer system:
I a.
Transfer Car Permissive Switch Since the transfer car controls are located in the I
f uel storage area, conditions in the containment are i
I not visible to the operator.
The transfer car permissivo switch allows a second operator in the l
containment to exercise some control over car movement if conditions visible to him warrant such control.
Transfer car operation is possible only when both lif ting arms are in the down position as indicated by the limit switches.
The permissive switch is a backup for the transfer car lif ting arm interlock.
Assuming the fuel container is in the upright position in the containment and the lifting arm interlock circuit fails in the permissive condition, the operator in the fuel storage area still cannot operate the car because of the permissive switch interlock.
The interlock can therefore withstand a single failure.
b.
Lifting Arm - Transfer Car Position Two redundant interlocks allow lif ting arm operation only when the transfer car is at the respective end of its travel and can therefore withstand a single failure.
9.1-39 l
l
o.
E/E-FSAR AMENDMENT 18 JANUARY 1979 Of the two redundant interlocks which allow lifting arm operation only when the transfer car is at the end of its travel, one interlock is a position limit switch in the control circuit.
The backup interlock is a mechanical latch device on the lifting arm that is opened by the car moving into position.
c.
Transfer Car - Valve Open An interlock on the transfer tube valve permits transfer car operation only when the. transfer tube valve position switch indicates the valve is fully open.
d.
Transfer Car - Lifting Arm The transfer car lifting arm is primarily designed to protect the equipment from overload and possible damage if an attempt is made to move the car when the fuel container is in the vertical position.
This interlock is redundant and can withstand a single failure.
The basic interlock is a position limit switch in the control circuit.
The backup interlock is a mechanical latch device that is opened by the weight of the fuel container when in the horizontal position.
e.
Liftino Arm - Refuelino Machine i
The refueling cavity lifting arm is interlocked with the refueling machine.
Whenever the transfer car is located in the refueling cavity, the lifting arm cannot be operated unless the refueling machine mast is in the fully retracted position or the refueling machine is over the core, f.'
Liftina Arm - Spent Fuel Pit Bridge Crane The lifting arm is interlocked with the spent fuel pit bridge crane.
The lifting arm cannot be operated unless the spent fuel pit bridge crane is not over the lifting arm area.
l Spent Fuel Pit Bridae Crane The spent fuel pit bridge crane includes the following safety features:
a.
The spent fuel pit bridge crane controls are interlocked to prevent simultaneous operation of bridge drive and hoist.
l l
l 9.1-40 l
I
_ _ _._._ ~ _ _- _. _,
B/B-FSAR AMENDMENT 45 JUNE 1984 b.
Bridge drive operation is prevented, except in the jog mode, when the hoist is in the full up position.
c.
An overload protection device is included on the hoist to limit the uplift force which could be applied to the spent fuel storage racks.
The protection device limits the hoist load to 100% (4000 pounds) of the rated 2-ton hoist capacity and can withstand a single failure.
d.
The design load on the hoist is the weight of one fuel assembly (approximately 1600 pounds), weight of one failed fuel container (approximately 1000 pounds), and the weight of the tool, which gives it a i
total weight of approximately 3000 pounds.
e.
Restraining bars are provided on each truck to prevent the bridge from overturning.
f.
Two independent wire ropes support the load and can withstand single failure.
I Fuel Handling Tools and Equipment All fuel handling tools and equipment handled over an open reactor vessel are designed to prevent inadvertent decoupling from machine hooks (i.e., lifting rigs are pinned to the machine hook, and safety latches are provided on hooks supporting tools).
Tools required for handling internal reactor components are designed with fail-safe features that prevent disengagement of the component in the event of operating mechanism malfunction.
These safety features apply to the following tools:
a.
Control rod drive shaft unlatching tool:
The air cylinders actuating the gripper mechanism are equipped with backup springs which close the gripper in the event of loss of air to the cylinder.
Air-operated valves are equipped with safety locking rings to prevent inadvertent actuation.
b.
Spent fuel handling tool:
When the fingers are latched, a pin is inserted into the operating handle and prevents inadvertent actuation.
The tool weighs approximately 385 pounds and is preoperationally tested at 1255 percent the weight of one fuel assembly (approximately 1600 pounds).
c.
New fuel assembly handling tool:
When the fingers are latched, a safety screw is screwed in, preventing inadvertent actuations.
The tool weighs approximately 100 pounds and is preoperationally 9.1-41
B/B-FSAR AMENDMENT 18 JANUARY 1979 tested at.125% the weight of one fuel assembly (approximately 1600 pounds).
t 0
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9.1-41a
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'4 BYRON /BR AIDWOOD STATIONS FIN AL S AFETY AN ALYSIS REPORT FIGURE 9.1-2 l
SPENT FUEL STORAGE RACK ARRANGEMENT J
FUEL CELL
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LEVELNG LEO RACK BASE 4 HOLO 00NN BOLT BYRON /BR AIDWOOD STATIONS FINAL S AFETY AN ALYSIS REPORT l
FIGURE 9.1-3
(
TYPICAL SPENT FUEL RACK IS0 METRIC
F ALED FUEL STOR AGE CELL 22 00 NCH PITCH TYP m
h
- c
's
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$UPPORT PLATE 184 50 NCH NOM.
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LEVELNG LEO 4 HOLD OCWN BOLT RACK SASE BYRON /BRAIDWOOD STATIONS FINAL S AFETY AN ALYSIS REPORT FIGURE 9.1-4 4
FAILED FUEL CONTAINER RACK l
~ -
B/B-FSAR AMENDMENT 44 DECEMBER 1983 15.7.3.3.2 Input Parameters and Initial Conditions The tank failure is evaluated in accordance with the following sets of assumptions and conditions:
a.
One hundred percent of the liquid volume of the boron recycle holdup tank is released into the boron recycle holdup tank cubicle.
b.
The liquid enters the groundwater environment through postulated cracks in the auxiliary building.
15.7.3.4 Radiological Consequences The radiological consequences of this accident are presented in Subsection 15.0.12.
The concentrations of any postulated accidental release of radioactive effluents from the boron recycle holdup tank would not exceed 10 CER 20 limits at the nearest surface water intake.
15.7.4 Fuel Handling Accidents 15.7.4.1 Identification of causes and Accident Description The accident is defined as dropping of a spent f uel assembly onto the spent fuel pit floor resulting in the postulated rupture of the cladding of all the fuel rods in the assembly despite many administrative controls and physical limitations imposed on tuel handling operations.
All refueling operations are conducted in accordance with prescribed procedures under direct surveillance of a supervisor.
15.7.4.2 Analysis of Eff ects and Consequences The f uel assembly from the core region discharged which has the peak inventory is the assembly assumed to be dropped.
The assembly inventory is determined assuming maximum full power t
operation at the end of core life immediately preceding shutdown.
The gap model discussed in Regulatory Guide 1.25 (May 1972) is used to determine the fuel-cladding gap activities.
Thus 10% of the total assembly iodines and noble gases, except for 30% for Kr-85, are assumed to be in the fuel-cladding gap.
The remainder of the assumptions used to determine the gap activity of the i
assembly are listed in Table 15.7-5.
The radial peaking factor given in this table is from Regulatory Guide 1.25.
The total assembly and fuel-cladding activities at the time of reactor shutdown are given in Table 15.7-6.
I I
15.7-6 l
B/B-FSAR AMENDMENT 43 SEPTEMBER 1983 15.7.4.3 Radiological Consequences of a Postulated Fuel Handling Accident Two analyses of a postulated fuel handling accident are perf ormed:
(1) a realistic analysis, and (2) an analysis based on Regulatory Guide 1.25.
The parameters used for each of these analyses are listed in Table 15.7-7.
The activity release to the atmosphere is given in Table 15.7-8 f or both the realistic and Regulatory Guide 1. 25 analyses.
Ihe short-term, i.e., 0-2 hour, atmospheric dispersion factor at the site boundary and the dose models, presented in Attachment 15A, are used in the analysis.
The thyroid, and gamma doses from a postulated fuel handling accident at the site boundary and low population zone are given in Tables 15.0-11 and 15.0-12 for the These doses are realistic and Regulatory Guide 1.25 analyses.
much less than the 10 CFR 100 reference values of 300 rem to the thyroid and 25 rem to the whole body.
15.7.5 Spent Fuel Cask Drop Accident 15.7.5.1 Identification of Causes and Frequency Classification A spent fuel cask will follow the path outlined in Figure 15.7-1.
Any single failure of any of the major components of the fuel handling crane would result in an undesirable condition, however, only f ailure of the main book would result in dropping of the cask.
A failure of one wire rope (one of four), one of the wheels for either the end truck or the bridge trolley, or an electrical fault would result only in the inability to safely J
l move the casx horizontally.
A failure in the lifting gear would result in an unsafe condition but would not affect (sling) the ability to raise or lower the cask.
The following items are assumed to be incapable of failure:
a.
the cable drum; b.
any structural component of the crane (bridge beams, rails, etc.) ;
any support for the crane rails which is tied into l
c.
the building steel; l
d.
crane interlocks; and stops which prevent movement of the cask over the e.
spent fuel pool.
Therefore, the only single failure in the crane which could cause a f uel cask drop is a f ailure in the main hook.
This accident is expected to occur with the frequency of a limiting f ault.
15.7-7
B/e-IS AR AMENDMENT 30 MARCH 1981 15.7.5.2 Evaluation and Analysis t igure 15.7-1 shows the general arrangement of the fuel handling butiding and tne recommended route tor the cask to move the spent fuel.
ouard walls are provided around the cask loading area in'the spent f uel storage pool as shown on the plan (Figure 15. 7-1) and in Section A-A (Figure 15.7-2).
These walls wnich surround the cask loading area, rise the f ull height of the pool and are structurally designed to withstand the impact force due to a falling cask.
If tre cask As positioned over the cask loading area and tips and falls, it will land on the guard walls.
Since the center of gravity of the cask is 'within the loading area, as noted on Figure 15. 7-2, the cask cannot tip over into the spent fuel storage pool.
In addition, tne f uel building crane will ne restricted from-o erating over the spent fuel storage pool by providing stops on e
the crane rails as indicated in Figure 15.7-1.
Th tuel handling building crane is designed so that travel of the crane will ce restricted so as to ensure that neither the hooxs nor the loads eney may carry can extend over the spent fuel storage racas.
Since the crane is required to travel over the spent fuel pit to remove casxs Irom the leaktight spent fuel cask storing area, the rails of the crane must be extended far enough to allow the crane to operate in this area.
Therefore, while the rails may extend over the spent fuel storage racks, tne crane's hook with or without load will not have this aDility.
Figure
- 15. 7-2 snows the hook's limit ot travel.
The crane interlocks, administrative controls, and structures are provided to prevent movement of the cask over the fuel storage areas, and provide added assurance that the integrity of the new f uel and spent fuel will not be compromised.
15.7.5.3 Barrier Performance The spent ruel pool walls are designed to withstand increased water pressure caused by a vertical drop of the cask.
If a cask drops on the exterior pool wall, and tips, it will land on the guard walls which are designed to withstand the impact torce due to a f alling cask.
The cask will not fall outside the cask storage well and thus will not aftect the fuel in the fuel storage pool.
15.7-8
)
AMENDMENT 30 B/B-FSAR MARCH 1981 Modeling of Accident Analysis 15.7.5.4 15.7.5.4.1 Mathematical Model A free drop of a fuel cask from a height of 30 f eet or more ontois not possible an unyielding surface (see 10 CFR 71, Appendix B)Therefore there will be no at the Byron /Braidwood Stations.
resulting damage to the cask that will cause a release of radioactive materials to the public.
Parameters and Initial Conditions 15.7.5.4.2 Input There are no input parameters and initial conditions of variables relevant to the evaluation of barrier performance that were not presented in Subsection 15.7.5.3.
i 15.7.5.4.3 Results be free dropped onto an unyielding Inasmuch as the cask cannot surf ace f rom a height of 30 teet or more, the cask will not be damaged and there will be no release of radioactive materials to the public.
Moreover, if damage to the cask compartment liner resulted from a the fuel pool would drop of the cask from maximum hook height, This is due to not become dewatered as a result of such damage.the watertight h
main portion of the fuel pool.
Since the cask will always be brought to the pool so as to enter is no the storage well without passing over the pool, there possibility of cask drop into the fuel storage pool.
or on stiff The ruel pool slab rests directly on rock (Byron)This slab was designed f l
soil (Braidwood).
the cask in the storage well.
f 15.7.5.5 Radiological Consecuences Since potential cask drop distances are less than 30 feet and appropriate impact limiting devices are employed during caskfue l
I severe than the fuel handling accident analyzed in Subsection movement, a spent 15.7.4.
l l
15.7-9
B/B-FSAR AMENDMENT 30 MARCH 1981 TABLE 15.7-6 ACTIVITIES IN PEAK INVENTORY DISCHARGED ASSEMBLY AT TIME OF REACTOR SHUTDOWN (BASED ON REGULATORY GUIDE 1.25)
ASSEMBLY ACTIVITY FRACTION OF GAP ACTIVITY ISOTOPE (Curies)
ACTIVITY m GAP (Curies) 5 4
I-131 7.52 x 10 0.10 7.52 x 10 6
5 I-132 1.15 x 10 0.10 1.15 x 10 6
5 I-133 1.68 x 10 0.10 1.68 x 10 6
5 I-134 1.98 x 10 0.10 1.98 x 10 6
5 I-135 1.53 x 10 0.10 1.53 x 10 3
2 Xe-131m 5.71 x 10 0.10 5.71 x 10 6
5 Xe-133 1.74 x 10 0.10 1.74 x 10 4
3 Xe-133m 4.41 x 10 0.10 4.41 x 10 5
4 Xe-135 4.75 x 10 0.10 4.75 x 10 4
Xe-135m 4.67 x 10 0.10 4.67 x 10 0
5 Xe-138 1.53 x 10 0.10 1.53 x 10 5
4 l
Kr-83m 1.40 x 10 0.10 1.40 x 10 3
3 Kr-85 8.54 x 10 0.30-2.56 x 10 5
4 Kr-85m 3.38 x 10 0.10 3.38 x 10 5
4 Kr-87 6.49 x 10 0.10 6.49 x'10 5
4 l
Kr-88 9.23 x 10 0.10 9.23 x 10 6
5 Kr-89 1.20 x 10 0.10 1.20 x 10 15.7-15
=
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TABLE 15.7-7 PARAMETERS USED IN FUEL llANDLING ACCIDENT ANALYSES REGULATORY GUIDE REALISTIC ANALYSIS 1.25 ANALYSIS Time between plant shutdown and accident 26.5 days
- 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> Maximum fuel rod pressurization i 1200 psia 1 1200 psia Minimum water depth between top of damaged 1 23 feet 1 23 feet fuel rods and pool surface Damage to fuel assembly One row of rods (17)
All rods ruptured ruptured m
Fuel assembly activity Average of fuel Highest powered fuel 3)
- w assemblies in core assembly in core 4
4 region discharged region discharged g
w Activity release to spent fuel pool Gap activity in Gap activity in ruptured rods ruptured rods **
Radial peaking factor 1.0 1.65 Form of iodine activity released to spent fuel pool elemental iodine 100%
99.75%
methyl iodine 0.0%
0.'25%
oz
- E wm mz Time to transfer one-half of the fuel assemblies in the core region discharged during m HI refueling, based on Westinghouse PWR operating experiences.
"w o
10% of the total radioactive iodine and 10% of the total noble gases, except for 30% for Kr-85, in the damaged rods at the time of the accident.
~
TABLE 15.7-7 (Cont'd)
REGULATORY GUIDE REALISTIC ANALYSIS 1.25 ANALYSIS Decontamination factor in spent fuel pool elemental iodine 760 133 methyl iodine 1
noble gases 1
1 Filter efficiencies in auxiliary building elemental iodine 90%
90%
m l
N methyl iodine 70%
I 4
m tn 8
Amount of mixing of activity in None None U
auxiliary building Meteorology See Attachment 15A See Attachment 15A mQ2 k
wm
@Z CD d ta O
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B/B-FSAR AMENDMENT 30 MARCH 1981 TABLE 15.7-8 ACTIVITY RELEASES TO ATMOSPHERE FROM FUEL HANDLING ACCIDENT REALISTIC ANALYSIS REGULATORY GUIDE 1.25 ISOTOPE ACTIVITY RELEASE (Ci)
ANALYSIS ACTIVITY RELEASE (Ci)
I-131 4.1 (-2) 5.3 (+1)
I-133 NEGLIGIBLE 6.0 (+1)
I-135 NEGLIGIBLE 5.0 (-1)
Xe-133m NEGLIGIBLE 7.5 (+2)
Xe-133 2.2 (+2) 1.0 (+5)
Xe-135m NEGLIGIBLE 1.2 (+3)
Xe-135 NEGLIGIBLE 4.6 (+1)
Kr-85 3.3 (+1) 3.3 (+3) 2 Note:
4.l(-2) = 4.1 x 10 l
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