ML20094S517
| ML20094S517 | |
| Person / Time | |
|---|---|
| Site: | Limerick  |
| Issue date: | 08/14/1984 |
| From: | Kemper J PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC |
| To: | Schwencer A Office of Nuclear Reactor Regulation |
| References | |
| OL, NUDOCS 8408210303 | |
| Download: ML20094S517 (28) | |
Text
-_
PHILADELPHIA ELECTRIC COMPANY 2301 M ARKET STREET P.O. BOX E699 PHILADELPHIA. PA.19101 (215)841 4502 AUG 141984 vicr#ntsmrNT ENGINEE N8MG ANC %ESE ARCH Mr. h. Schwencer, Chief Docket Nos.:
50-352 Licensing Branch No. 2 50-353 Division of Licensing U. S. Nuclear Regulatory Conmission Washington, D.C. 20555
Subject:
Limerick Generating Station, Units 1 and 2 Information for Auxiliary Systems Branch (ASB) and Accident Evaluation Branch (AEB) Regarding Light Loads / Fuel Handling Accident
References:
(1) Telecon between L. Bell /R. E. Martin (NRC) and J. H. Arhar (PECO) on 4/24/84.
(2) Letter, J. S. Kenper 'PECO) to A. Schwencer (NRC), dated 8/13/84.
File:
GOVT 1-1 (NRC)
Dear Mr. Schwencer:
Attached are draft changes to FSAR Section 15.7.4 and response to RAI 410.37.
These changes evaluate a more conservative fuel-handling accident than that presently described in the FSAR.
As discussed in Reference (1), the resulting post accident offsite doses remain a small fraction of 10CFR100 1Imits.
Pursuant to our conmitment noted on SER page 9-9, the evaluation contained in the revised response to Q410.37 demonstrates that the maxinun kinetic energy resulting from the drop of each object weighing less than a fuel bundle and grapple assenbly that could be handled over spent fuel will not result in consequences in excess of the revised fuel-handling accident.
The revised fuel-handling accident evaluation is consistently applied to the " Limerick Overhead Handling Systems Final Report" (Revision 3), transmitted to you in Reference (2).
k B40021Oh384081405000352 O
i g
I
a-The information contained on these draft FSAR changes will be incorporated into the FSAR, exactly as it appears on the attachments, in the revision scheduled for Septenter 1984.
Sincerely, n
JHA/gra/07268405 Attachment cc: See Attached Service List
T cc: Judge Lawrence Brenner (w/o enclosure)
Judge Richard F.' Cole (w/o enclosure)
Troy B. Conner, Jr., Esq.
(w/o enclosure)
Ann P. Hodgdon, Esq.
(w/o enclosure)
Mr. Frank R. Romano.
(w/o enclosure)
EMr. Robert L. Anthony
-(w/o enclosure)
Charles W. Elliot, Esq.
(w/o enclosure)
Zori G. Ferkin, Esq.
(w/o enclosure)
Mr. Thomas Gerusky (w/o enclosure)
Director, Penna. Emergency (w/o enclosure)
Management Agency Angus R. Love, Esq.
(w/o enclosure)
David Wersan, Esq.
_ (w/o enclosure)
Robert J. Sugarman, Esq.
(w/o enclosure)
Spence W. Perry, Esq.
(w/o enclosure)
Jay M. Gutierrez, Esq.
(w/o enclosure)
Atomic Safety & Licensing (w/o enclosure)
Appeal Board Atomic Safety & Licensing (w/o enclosure)
Board Panel Docket & Service Section (w/o enclosure)
Martha W. Bush, Esq.
(w/o enclosure)
Mr. James Wiggins (w/o enclosure)
Mr. Timothy R. S. Campbell (w/o enclosure)
Ms. Phyllis Zitzer (w/o enclosure)
Judge Peter A. Morris (w/o enclosure) i i
1
.)
F5 - Co %
r,:m u
lu O "iE. $ __
OUESTION 410.37 (Section 9.1.2, 9.1.4) 4 Verify that the maximum potential kinetic energy resulting from dropping each object of less weight than a spent fuel assembly and its handling tool, which will be handled over spent fuel, will not exceed the effects of the fuel handling accident described in Section 15.7.4 of the FSAR.
Provide a list of all objects considered and a discussion of the analysis.
RESPONSE
[A4pfot9_4 RAP 94 ASSEAS ki As noted in Lthe discussion of th design basis fuel handling accident in ection 15.7 th maximum kinetic energy of a dropped fuel (cundleAis Q ft-lb.
A' review has been made to V
determine whether there are any potential drops of loads lighter f
g than a fuel bundleAthat could have a higher kinetic energy due to Ig a higher carryino/heicht.
The.following conclusions have been reached: (ht0D FOEL 4R.M M AGG S LY (,1. E., IL*o h D
No load that weighs less t Ib can develop a higher kinetic energy than a fuel bundleAlf ropped__over spent uel.
This value i J is based on a potential energy of 6",000 ftM 4 with the load at
$g the maximum lift height of the reactor enclosure crane and ap i
relative to the reactor core (worst case).
The majorit of light li.
loads carried over spent fuel weig less than Ib. 5 4-o p
bit. it.e s-ccyti;.. ;f tr.;;; it;;; Alisted in Table 410.37-1, the,
w potential.
gy of the remainino few light loads which(weign to l
more tha Ib is less than "" ^^^ ft-lb because their maximum d
dr?.pjeigggsage1gssgpangpe gase. g }:gg_}{gr W M
g.
3 a.
The load is carried only by the refueling platform
~ "
hoist.
b.
The load is carried only over the spent fuel pool.
l l
c.
The load is very long (i.e. the bottom of the load is t
close to the top of the fuel).
l The following light loads may develop a higher kinetic energy i
than a dropped fuel bundle.
The approximate potential energies j
at normal and maximum load carrying heights are listed, relative to the elevation of the top of the spent fuel in the core or in the spent fuel pool as appropriate.
l It is reasonable to assume that the consequences of a light load drop will be no worse than those of the design basis fuel bundle I
drop for the following reasons:
l 410.37-1 Rev. 23, 08/83
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The actual kinetic energy developed during the drop of a.
the light loads above will be less than their maximum potential energy due to bouyancy and drag of the wat over the fuel.
While no calculations have been ma ductions in kinetic energy due to drag should si ificant for steam line plug and the in-ves sto e rack due to their relatively large s ace areas.
b.
The loads ill usually be carried at le than maximum height (i.e the bottom of the load I normally be near the ref ling floor or bottom o the reactor well, as applicable).
c.
As discussed in Se ion 15.7.4 all of the fuel rods in the dropped spent f I bundle are assumed to fail, representing 50% of t res.. ting fission product release.
Because no fi n products are released from a dropped light load (i ding an unieradiated new fuel bundle), all releases ust me from the impacted spent fuel.
Thus, for th case of light load drop, the impacted fuel can sorb more ergy without exceeding the releases cal lated for the nt fuel bundle drop.
d.
Some of the pact energy would be a orbed by components her than the spent fuel the dropped load (e.g., th fuel storage rack, core top ide or other impacte items) that would further reduce he energy avail e to cause fuel failure.
The three 1 ht loads listed in Table 410.37-1 will be eated as heavy loa in accordance with the guidelines of NUREG-06 until e culations are performed which demonstrate that the effect of dropping these objects will not exceed the effects f
the el handling accident described in Section 15.7.4 of the FS W (6T6 IM 9%E
- i l
Rev. 23, 08/83 410.37-2 l
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- - - _,,,, -. - -..,... ~
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L Table 410.37-1 Approx. Combined Potential Energy, Wt., Handling (ft-lb)
Tool Plus Load Normal Max.
Load (Ib)
Height Height 1)
New fuel bundle or 700 21,000 29,000 dummy bundle
[ Reactor enclosure crane relative to spent fuel pool]
2)
In-Vessel storage 600 21,000 33,000 rack
[ Refueling platform hoist relative to core]
ff%AML 3)
Steam line plug and 450 16,000 33,000 installing tool
[ Reactor enclosure crane relative to core]
410.37-3 Rev. 23, 08/83
E6 -W(c LGS FSAR 15.7.4 FUEL-HANDLING ACCIDENT tr?
O'r'C b
I.
b 15.7.4.1 Identification of Causes and Frecuency Classification 15.7.4.1.1 Identification of Causes The fuel-handling accident is assumed to occur as a consequence of a failure of the fuel assembly lifting mechanism resulting in the dropping of a raised fuel assembly onto other fuel bundles.
A variety of events that qualify for the class of accidents termed " fuel-handling accidents" has been investigated.
The accident that produces the largest number of failed spent fuel rods is the drop of a spent fuel bundle into the reactor core g
when the reactor vessel head is off.
Mp -rm Fost. 6RAppg_ Aum6i.1 Tag _ m W e4 (_e.% 5TS cf A OF T+E. SESroEu h PJT FOR M TEt E.decfic K AW A.No w A% mt55'(.
15.7.4.1.2 requency Classification 55g_MSLQ This accident is categorized as a limiting fault.
15.7.4.2 Sequence of Events and System Operation 15.7.4.2.1 Sequence of Events The sequence of events folicwing this failure is shown in Table t15.7-15.
15.7.4.2.2 Identification of Operator Actions The operator actions are as follows:
a.
Initiate the evacuation of the refueling area and the locking of the refueling area doors.
b.
The supervisor in charge of fuel handling should instruct his employees to go immediately to the radiation protection personnel decontamination area.
c.
The supervisor in charge of fuel handling will alert the
(
control room operator to the acci, dent.
Rev. 15, 12/82 15.7-10
.f6 - & f 4~
DRAFT
(
d.
Determine if the normal ventilation system.has isolated and the SGTS is in operation.
Initiate action to determine the extent of potential e.
radiation doses by measuring the radiation levels in the vicinity of or close to the refueling area.
f.
Appropriate radiological control methods should be implemented at the entrance of the refueling area.
g.
Before entering the refueling area, a careful study of l
4 conditions, radiation levels, etc, will be performed.
l 15.7.4.2.3 System Operation Normally, operating plant instrumentation and controls are assumed to function, although credit is taken only for the isolation of the normal ventilation system and the operation of the SGTS.
Operation of other plant or RPS or ESF systems is not expected.
15.7.4.2.4 The Effect of Single Failures and Operator Errors The automatic ventilation isolation system includes the radiation monitoring detectors and isolation valves.
The SGTS is designed to the single failure criterion and safety requirements.
Refer to Sections 7.6, 9.4 and 15.10 for further details.
15.7.4.3 Core and System Performance 15.7.4.3.1 Mathematical Model The analytical methods and associated assumptions used to evaluate the consequences of this accident are considered to provide a realistic yet conservative assessment of the consequences.
+ lMt.t.77 W N MD i %t.
- To estimate the expected number of failed fuel rods in each impact, an energy approach is used.
15.7-11 Rev. 15, 12/82
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The fuel assemblyA m expected to impact on the reactor core at a k
small angle from the vertical, possibly inducing a bending mode It is of failure on the fuel rods of the dropped assembly.
assumed that each fuel rod resists the imposed bending load by a couple consisting of two equal, opposite concentrated forces.
Therefore, fuel rods are expected to absorb little energy prior to failure as a result of bending.
Actual bending tests with 1 concentrated point loads show that each fuel rod absorbs Each rod that (g approximately 1 ft-lb prior to cladding failure. fails as a result of gross absorb approximately 250 ft-lb before cladding failure (based
_J The energy of upon 1% uniform plastic deformation of the rods).
8 the dropped /assembljd is conservatively assumed to.be absorbed by I
Because a fuel only the cladding and other core structures.
assembly consists of 72% fuel, 11% cladding, and 17% other structural material by weight, the assumption that no energy is absorbed by the fuel material results in considerable conservatism in the mass energy calculations that follow.
The energy absorption on successive impacts is estimated by considering a plastic impact.
Conservation of momentum under a plastic impact shows that the fractional kinetic energy absorbed during impact is:
1-M, Mi + M, where M is the impacting mass and H, is the struck mass.
i 15.7.4.3.2 Input Parameters and Initial Conditions The assumptions used in the analysis of this accident are listed below:
% 69.+PPuE A456MEAM AEG-The fuel assemblysts. dropped from the maximum height a.
allowed by the fuel-hand!!n; equiprent '10cc ther 20Tfo9.%(A
'MfLEPOEt t At= PL A:
S G.P>EA~ Chi >+1D 10 bE PG0oE4T UO LTS.
y Mon _y peg t.
The entire amount of potential energy, referenced to the b.
is available for application to top of the reactor core, This 4,c,yg, the fuel assemblies involved in the accident.
l i
assumption neglects the dissipation of some of the Ausset mechanical energy of the falling fuel assembly in the water above the core._fnd requires the complete V the assembly from the fuel-hoistingf (Detachment ot Rev. 15, 12/82 15.7-12
FSM b DRAFT
" ' ' ' ^ "
c
[ equipment.
This is only possible if the fuel assembly Lhandle, the fuel grapple, or the grapple cable break, None of,the energy associated with the dropped @um-c.
assembl/isabsorbedbythefuelmaterial(uranium dioxide).
d.
The minimum water depth between the top of the fuel rods and the fuel pool surface is 23 feet.
Maximum fuel rod pressurization is 2072 psia.
e.
f.
The peak linear power density for the highest power assembly discharged is 13.4 kWf t and the corresponding maximum centerline operating fuel temperature is 34120F.
6RAPPLE.
p SprAvse Tee. g AsE14HrT5. cv THE. Ft,s., A65Ghs'y Atap%d Hli.M ASsEngsy Act 15.7.4.3.3 Resu1ts MSVMu To gg. 'THE. SN foc f5cn% Ass %iss, 15.7.4.3.3.1 Energy Available
@OM h
Dropping a fuel assembly onto thg reactor core from the maximum height allowed by the refueling C v,.c.... (
afeet) results in an impact velocity of Rf t/sec. 45 A pt Age 44 T'ne FCEL 4e#FLE. ACSEM&by CATc T6t. RE.ALTD p. Cos2d Ft.p u TM MAy.lM0 g HE.iGar hu cCE.D TWE.94LFoEL-tm 9ATPogn L47 FEET) Pe3WT5 i4 An iHtw.T 4Etc<.4ry of 55.0 FT/sec.
Thedinetic energy acquired by the falling i " _ _.. ~,
6xft-lb and is dissipated in one or more impacts.g:
r g
(A55EMtk.LES15 AfPILc% i Mkt c.cy 45)oD 15.7.4.3.3.2 Energy Loss Per Impac Tpg Lo% dAnud SHAPf W' h
g, Burn oe.certo A%exsLies l
l Based upon the fuel geometry in the reactor cor d four fuel l
assemblies are struck by M e dropped assembly.
The fractional s
i energy loss on the first impact is approximately 80S..
A The secon impact is expected to be less direct.
The broadsid ropped assembly impacts approximately 24 more fuel tr.s TM l
of 240 assemblies, so that after the second impact r.ly 135fft-lb (approximately 1% of the original kinetic energy) is available for a third impact.
Because a single fuel rod is capable of it is absorbing 250 ft-lb in compression before cladding failure, unlikely that any fuel rod will fail on a third impact.
s EcoAs.s %E. SOM of THE k.inTst E 4 ERG 4 CF TM.
T+e. Tcn'4 kww.Tse h4.R4y(,,'3% FEET % 100 kB ::: 2;2 400 FT-L.8) Mp Tpg.
t4t.opeso poet ASSsagvf 7
15.7-13 Rev. 15, 12/82 Of2cerE.o post. 6p. APP 4 A55me>s7 (.47 FEET x 500 t-6 = A3,6co Pr-i Q,
~
p-m
}h f.
ChG^"
(
w If p dropped 4 set assembly strikes'only one or two fuel assemblies on y first impact, the energy absorption by the core support structure results in approximately the same energy dissipation on the first impact as in the case where four fuel The energy relations on the second and assemblies are struck.
third impacts remain approximately the same as in the original Thus, the calculated energy dissipation is as follows:
case.
First impact 80%
Second impact 19%
Third impact 1% (no cladding failures).
15.7.4.3.3.3 Fuel Rod Failures 15.7.4.3.3 3.1 First Impact Failures 45/foo 36,W The first impacts dissipate [0.80 x 17,000 or E,000 f t-lb of l
It is assumed that 50% of this energy is absorbed by the dropped, f uel asse_ bly..and that the remaining 50% is absorbed by energy.
m the struck fuel assemblies in the core.A Because the fuel rods of the dropped fuel assembly are susceptible to the bending mode of failure and because 1 ft-lb of energy is sufficient to cause cladding failure as a result of bending, all 62 rods of the Because the eight tie dropped fuel assembly are assumed to fail. rods of each struck fuel 0
it is assumed that they fail failure than the other 54 fuel rodsMtierods(totalinfour on.the first impact.
Thus G x 8 = 3 V.* 4 V B e b4 assemblies,) are assumed to fai1.
2 lSTt0u BY EAc.w ce -tM. Two A55s.M6t.atS g Because the remaining fuel rods of the struck assemblies are held { g rigidly in place in the core, they are susceptible only to the To cause cladding failure of one compression mode of failure.
fuel rod as a result of compression, 250 ft-lb of energy is To cause f ailure of all the remaining rods of the our required.
struck assemblies, 250 x 54 x 4 or 54,000 ft-lb of energy would l
Thus, it is clear that have to be absorbed in cladding alone.
not all the remaining fuel rods of the struck assemblies can fail The number of fuel rod failures caused by on the first impact.
compression is computed as follows:
O$o CF AoprriodAO-y ; lT is Gtal5EMATIVcQ /h30MEp 'THRT
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4+-
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=
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.= 36 GPAPPLE. ASSF M641,e
~ 49 Thus,duringthefirst3cmpact, fuel rod failures are as follows:
2.
poet.
Dropped assembly 62 rods (bending)
Struck assemblies (d -R tie rods (bending) g Struck assemblies 49 5+ rods (compression) 4ea.r failed rods
\\'70 hF
])
15.7.4.3.3.3.2 Second Impact Failures Q;
s e.e d 2.
Because of the less severe nature of the second impact and the ropped -deet-assembly, it is assumed that distorted shape of 24fe in only two of the 24 struck assemblies are the tie rods S-I Qtie rods are subjected to bending f ailure. -Th ; 2 :The number of fuel rod failures caused by assumed to fail.
compression on the second impact is computed as follows:
Q(LeppE.o
- o J.L N> b 11 3
N ^^0 X 11 + 17
=
p m A55W69 3
DdL*F9Ew Fun t/"
wp0E. >65E MBW 'I**"
w
/o Thus, during the second impact, the fuel rod failures are as follows:
Struck assemblies 33 -Mr tie rods (bending)
Struck assemblies 10
,_-G rods (compression) y-2. -3$- f a i l ed rods 15.7.4.3.3.3.3 Total Failures is as The total number of failed rods resulting from the accident l
follows:
j
)
First impact l ro +0'ir rods i
Second impact 441. 49-rods
'O rods Third istpact
_L W total failed rods
)
7, t 2.
\\
15.7-15 Rev. 15, 12/82
f.S-6. Wc II,5 FSAR 74 Avdr No CMht4bE
{
~~~~
15.7.4.4 Barrier Performance 1
The reactor coolant pressure boundary and primary containment are assumed to be open.
The transport of fission products from the refueling area is discussed in Sections 15.7.4.5.2.1 and 15.7.4.5.2.2.
15.7.4.5 Radioloolcal Consecuences Two separate radiological analyses are provided for this accident The first is based upon conservative assumptions a.
considered to be acceptable to the NRC for the purpose of determining adequacy of the plant design to meet 10 CFR, Part 100.
This analysis is referred to as the
" design basis analysis."
b.
The second analysis is based upon assumptions considered to provide a realistic conservative estimate of radiological consequences.
This analysis is referred to as the " realistic analysis."
For both analyses, the fission product inventory in the fuel rods assumed to be damaged is based upon 1000 days of continuous operation at 3458 MW.
A 24-hour period for decay from the above power condition is assumed, because it is not expected that fuel handling can begin within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following initiation of reactor shutdown.
Figure 15.7-1 indicates the leakage flow path for this accident.
15.7.4.5.1 Design Basis Analysis The design basis analysis is based upon NRC Standard Review Plan 15.7.4 and NRC Regulatory Guide 1.25.
The specific models and assumptions and the program used for computer evaluation are described in Section 15.10.
Specific values of parameters used in the evaluation are presented in Table 15.7-16.
k Rev. 15, 12/82 15.7-16
FS -41Y6 r
/4eo MYy LGS FSAR Qo gg 15.7.4.5.1.1 Fission Product Release from Fuel The fission product inventory of a core average rod is adjusted by a peaking factor of 1.5 to establish the inventory of each damaged rod.
Ten percent of the noble gases inventory (30% for Kr-85) and 10% of the iodine inventory are assumed to be released to the reactor water and water in the reactor well.
The activity airborne in the refueling area is presented in Table 15.7-17.
15.7.4.5.1.2 Fission Product Transport to the Environment The transport pathway consists of mixing in the reactor well, migration from the flooded well to the refueling area atmosphere, and release to the environment through the SGTS.
All of the noble gas and 1% of the iodines in the flooded well are assumed to become airborne in the refueling area.
The airborne activity is released to the environment over a 2-hour period after filtration by the standby gas treatment system (SGTS) (99% removal efficiency for iodine).
The release of activity to the environment is presented in Table 15.7-18.
15.7.4.5.1.3 Results The calculated exposures for the design basis analysis are presented in Table 15.7-21 and are well within the guidelines of 10 CFR, Part 100.
15.7.4.5.2 Realistic Analysis 1
The realistic analysis is based upon a realistic but still conservative assessmer.t of this accident.
The specific models and assumptions and the program used for computer evaluation are described in Ref 15.7-1.
Specific values of parameters used in the evaluation are presented in Table 15.7-16.
i l -
t 15.7-17 Rev. 15, 12/82
~
MS FSAR
[ {
15.7.4.5.2.1 Fission Product Release from Fuel Fission product release esti.ates for the fuel-handling accident m
are based on the following assumptions:
The reactor fuel has an average irradiation time of 1000 days at 105% nuclear boiler rated (NBR) up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> a.
This assumption results in an prior to the accident.
equilibrium fission product concentration at the time the reactor is shut down.
Longer operating histories do not increase the concentration of biologically The 24-hour decay period allows significant isotopes.
time to shut down the reactor, depressurize the nuclear system, remove the reactor vessel head, and remove the It is not expected tht.t reactor vessel upper internals.
these operations could be accomplished in less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and probably will require at least 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, An average of 1.8% of the noble gas activity and 0.32%
of the halogen activity is in the fuel rod plena and b.
This assumption is based upon available for release.
{
fission product release data from defective fuel experiments (Ref 15.7-2).
Because of the negligible particulate activity available for release from the fuel plena, none of the solid c.
fission products are assumed to be released.
MF This is It is assumed that.htt-fuel rods f ail.
considered to be conservative, because it is expected d.
that fewer than-tst rods would be damaged.
b.G i
Fission Product Transport to the Environment 15.7.4.5.2.2 j
l The following assumptions and conditions are used in calculating the release of activity to the environment.
All of the noble gases released to the fuel pool become a.
f l
airborne in the refueling area, The iodine activity airborne is in proportion to the t
b.
partition factor and the ratio of the volume of the refueling area (Va) to the volume of fuel pool water
)
l Rev. 15, 12/82 15.7-18 i
i
S
(
above the core (Vw).
It is assumed that a partition l
factor of 100 and Va/Vw of 10 is applicable for this event.
It should be noted that the volume assumed for Va is not equal to the total volume of air in the refueling area, but is a conservative estimate of the l
volume of air that may form an equilibrium condition vith the activity in the refueling pool.
c.
The ventilation rate from the refueling area to the l
environment through the SGTS is 0.5 volume change per day (99% removal efficiency for iodine), assuming 764 sefm inleakage to the refueling area.
Based upon these assumptions, the activity airborne in the refueling area is as shown in Table 15.7-19.
The release rate of activity under normal ventilation conditions is sufficient to cause a trip of the refueling area discharge plenum radiation monitors, which results in refueling area isolation and SGTS startup.
The activity released to the environment is presented in Table 15.7-20.
15.7.4.5.2.3 Results The calculated exposures for the realistic analysis are presented in Table 15.7-21 and demonstrate the margin of conservatism in the design basis analysis.
l 15.7.5 SPENT FUEL CASK DROP ACCIDENT l
l l
The spent fuel cask will be equipped with redundant sets of lifting lugs and yokes compatible with the single-failure-proof reactor enclosure crane and main hook, thus precluding a cask drop due to a single fail ~ure.
Therefore, an analysis of the spent fuel cask drop is not required.
Refer to Section 9.1.5 for a description of the reactor enclosure crane and the interlocks that prevent moving the spent fuel cask over the fuel pool.
I li.
15.7-19 Rev. 15, 12/82 i
F6- &%
No NW (
M!4 M g
\\
15.
7.6 REFERENCES
15.7-1
- Nguyen, D., " Realistic Accident Analysis - RELAC Code,"
October 1977 (NEDO-21143).
15.7-2
- Horton, N.R.,
W.A. Williams, and J.W. Holtzclaw,
" Analytical Methods for Evaluating the Radiological Aspects of the General Electric Boiling Water Reactor,"
March 1969 (APED 5756).
l l
/
i
's f
Rev. 15, 12/82 15.7-20
pOgi.C n
re TABLE 15.7-15 SEQUENCE OF EVENTS FOR FUEL-HANDLING ACCIDENT TIME-MIN EVENT 0
Fuel assembly is being handled by refueling equipment.
Theg ssembly drop onto the top of a
g the core.
Fh W Th WPte A%Ek6W 0
Some of the fuel rods in both the droppeo assembly and reactor core are damaged, resulting in the release of gaseous fissior, products to the reactor coolant and eventually to the refueling area atmosphere.
l
<1 The refueling area ventilation radiation l
monitoring system alarms to alert plant personnel, isolates the ventilation systen and starts operation of the SGTS.
<5 Operator actions begin.
L Rev. 15, 12/82 c..
,..:* hm 1
c:
c.. l _ h b
_~
(.
TABLE 15.7-16 (Page 1 of 2)
FUEL-HANDLING ACCIDENT:
PARAMETERS TABULATED FOR POSTULATED ACCIDENT ANALYSIS DESIGN REALISTIC BASIS BASIS ASSUMPTIONS ASSUMPTIONS I.
Data and Assumptions used to Estimate Radioactive Source from Postulated Accidents A. Power Level 3458 3458 B. Radial Peaking Factor 1.5 1.0 C. Fission Products Released dHbe rods 444 rods From Fuel (fuel damaged) 1,Lt.
EdL D. Release of Activity by Table 15.7-17 Table 15.7-19 Nuclide E. Iodine Fractions (1) Organic 0
0 (2) Elemental 1
1 (3) Particulate 0
0 F. Reactor Coolant Activity NA NA Before the Accident II.
Data and Assumptions Used to Estimate Activity Released A. Primary Containment Leak NA NA Rate (%/ day)
(EE Emde,/4f4 B. Secondary Containmen elease 100% for 50%/ day Rate (%/ day) 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> C. Valve Movement Times NA NA D. Adsorption and Filtration (1) Organic iodine 99%
99%
(2) Elemental iodine 99%
99%
(3) Particulate iodine 99%
99%
(4) Particulate fission NA%
NA products E. Containment Spray Parameters NA NA (flow rate, drop size, etc)
F. Refueling Area Containment 2.2 x 108 2.2 x IO*
l Volumes (fts)
G. All Other Pertinent Data None None l
and Assumptions i
III. Dispersion Data A. EAB/LPZ Distances (m) 731/2043 731/2043 s
Rev. 15, 12/82 l
Mo WE ( /6-U(sf,g 4 p LGS FSAR TABLE 15.7-16. (Cont'd)
(Page 2 of 2)
DESIGN REALISTIC BASIS BASIS ASSUMPTIONS ASSUMPTIONS B. X/Os for Time Intervals of (1) 0-2 hrs - EAB 2.9 x 10-*
1.2 x IO-*
(2) 0-8 hrs - LP2 4.0 x 10-8 2.0 x 10-s (3) 8-24 hrs - LPZ NA 1.6 x 10-8 (4) 1-4 days - LPZ NA 9.0 x 10-*
(5) 4-30 days - LP2 NA 4.2 x 10-*
IV.
Dose Data A. Method of Dose Calculation Section 15.10 Ref 15.7-1 B. Dose Conversion Assumptions Section 15.10 Ref 15.7-1 C. Peak Activity Concentrations NA NA in Containment D. Doses Table 15.7-21 Table 15.7-21 I
I 1
I l(
lL Rev. 15, 12/82 l
..c
PS-C-9 to PR CN, O pur..:
t.
l'
- ' i l6 L a
._ _ a LGS FSAR TABLE 15.7-17 FUEL-HANDLING ACCIDENT:
ACTIVITY AIRBORNE IN REFUELING AREA l
_ DESIGN BASIS ANALYSIS ISOTOPE ACTIVITY
_ (Ci)
I-131 I-132 5.2 5
.;M91" x 10+2 I-133 5 97 Wx 10-2 I-134 5,97 Wx 10+2 I-135 t. 0)
.5.#5"' x 10 +/2 Kr-83m Kr-85m t.33 1=*'f 1 :^
-5,6 5 W x 10+2 Kr-87 2.60.L,@r x 10+3 Kr-88 9.~1 5 W x 10-2 Kr-89 j 8l L,4r6" x 10+2 Xe-131m Xe-133m M.03 3==i>?" x 10+2 Xe-133
- 2. 4 b.1.ssM x 10+3 i
Xe-135m g i g' JWP'T x 10 +# 5 Xe-135 Xe-137 Z.o2.1.,,Mr x 10+*
1 Xe-138 i
Rev. 15, 12/82 I
95 -b %
z [7 [', f..L:,;
LGS FSAR d b'Cl.7 [___ i TABLE 15.7-18 FUEL-HANDLING ACCIDENT:
ACTIVITY RELEASED TO THE ENVIRONMENT DESIGN BASIS ANALYSIS ISOTOPE ACTIVITY (Ci)
I-131 I-132 52.5" W 6,g,9 W x 10-3 I-133 gg W I-134 I-135
(.ol 2
- *: 2 a'
Kr-83m l
Kr-85m t.33 " 2^ _.:- 1 Kr-85 F,66 W x 10+2 Kr-87 2..bo W x 10+3 Kr-88 9.16 W x 10-2 Kr-89 g,si W x 10+2 Xe-131m i
Xe-133m 9.o1 W x 10+2 Xe-133 2.4bL M*x 10+3 Xe-135m g,( y(pff x 10+f 5 Xe-135 Xe-138 202 A x 10+4 h
i Rev. 15, 12/82
-,.e.g m
r_,
h
[g[
d FUEL-HANDLING ACCIDENT:
ACTIVITY AIRBORNE IN REFUELING AREAC 1)
REALISTIC ANALYSIS i
ISOTOPE 2 HRS 8 HRS 1 DAY 4 DAYS 30 DAYS I
1-131 4.o7 ;l,,.aex 101 3.n 1,ac x 101 1,5i a,.,ef x iO.
5,aca 45 x 10-1 5.is 3,eg"'x 10-13 2
},,7f" 3.t*c 3,,4!T x 10-1
- ' g S.ae'x 10-r3 4.r2.?,hP'x 10-1*
0.0 I
I-132 4,.'t 3 49'x 10-3 0.0 1.99,2 g.neg' y,u 7 3,,,#'5
( yl L : _ :: -
9 I-133 I-134 1,i1.f>,e"5 x 10-/7 7/I6 b.49'x 10-10 1.24 ~/, Gi!' x 10-*#3 0.0 C.O 3
- 6. L-z!(7 3.,eir" x 10-1 2,w h,ef x 10-2 B.24 G,,ee x 10-7 0.0 0.0
.i.-3>M,,eg' x 10-1 I-135 73,pg6 x 10-1
- 2. fl'i 3 72* x 10-2 4 9o3,e5 x 10-5 456.%.@9 x 10-18 Kr-85 8.49 g,e'3" x 102 g,g(, A x 102 332 g,ef x 102
(.65 S.69 Y-lO' 8.M g,96' x 10-11 l
Kr-83m 0.0 Kr-85m
- 7. 14 1 3,.eig" x 1 0 1
- c. vt 3,,eg' 2.0-11.,66' x 10-1 1.9l Sai!'f' x 10.F7 Kr-87 7.5 4 S,spe' x 10-*
z.3) L,ne x 10-s 234L,et x 10-*
0.0 0.0 0.0 Kr-88 3,13 L,.esr' 5.9-(3,4e*x 10-2 5.04 2 8e' x 10-3 3.ao 3.=iM x 10-12 0.0 0.0 0.0 0.0 i.vl3,.etx10-ik Xe-131m 9.e3 W x 101 7.~df4,@f x 101 3.bb3,9f x 101 g,g 3 _ ^ '.. :: -
g."13 3,,pg"x 10-12 Kr-89 Xe-133m
.2.f1,, set x 10/ 5 1.18 %,@T x 102 3.Eb 2,@e x 102 7.76 SweW
t,5% %,4@'x 10-191 Xe-133 1.12.doset x 10/4 6.GG ee'x 103 3.T)3 ae*x 103 g,3y 3,@f x 10f 2 2.,I"L.LeiMf*x 10-11 Xe-135m 7,7l g,,s,M=x 10-3 6.eI(kspeT x 10-10 0.0 0.0 0.0
~
0.0 Xe-135
- 2. 4'/1,@W x 103 I.*t l ib,@f' x 10,53 2..vf 3 @f' x 10 2 4.66 2,.iPf x 10-a Xe-137 g,o g g, ps' x 10-154 0.0 0.0 0.0 0.0 0.0 0.0 0.0 Xe-138 3,56 W x 10-7 (y[ 6,,@S' x 10-*f3 TOTAL l.bi S,se'f' x 1074
(.ito f>,4"f x 10['i q,9 c.2. 90 x 103
(.6o S,,e'f' x 10/2.
i,08 [h,A9'x10-1(
il (1) Units for activities are in curies.
l e
l
\\
F9-(c%
TABLE 15.7-20
(,j [' /4 a J
FUEL HANDLING ACCIDENT:
ACTIVITY RELEASED TO THE ENVIRONMENT <13 REALI ANALYSIS ISOTOPE 0-2 HR 2-8 HR 8-24 HR 1-4 DAYS 4-30 DAYS I-131 3.% 2.08 x 10-2 8.D.S 42"'x 10-2 1.4~] W x 10-4 t.33 3ee0 x 10-/;
6.L'2-3,sef x 10-3 I-132 3.To 2.05 x 10-3 3.13 L,ef x 10-3 47oanff x 10-*
2,071 r't x 10-6 5',tB 3 dPT x 10-17 i
I-133 c.GB 3.85 x 10-3 1.y~7 8.,ef'*x 10-/2 i.81 L.MP x 10-2 7.63Gnef x 10-3 3.u) 3s*M x 10-5 4
0.0 I-134 4551.49 x 10-10 6.'f o.3eg'$ x 10-11 4.oc],,iP6 x 10-13
- c. 433.#ff x 10-19 I-135
- 6. 43 3.76 x 10-*
14 6.e90 x 10-/3 1,19Enff x 10-4 7,G 4seg x 10-5 Z..~3 6 b @f x 10-S Kr-83m 4.31 2.52 x 10-2 3.t5 ker x 10-2 3.o33, g'f* x 10-3 y,e32eff x 10-6 cf.<e-3 WiMT x 10-2 9 Kr-85 7.20 4.21 x 102
[.e31 67'* x 102 3,62,eg x 102 3,g 1,,eg x 102 s.sg g,eg ggo' Kr-85m 2..L'l1.33
.3.z.) L ee 1.323milehusemm90mme 5.513esf4 x 10-2 2.Vtamff" x 10-e Kr-87 1.t8 6.89 x 10.H 6.183,49 x 10-5 i M Lee"f x 10-6 t.7S1 4W"x 10-10 0.0 Kr-88 3.9f2.07 x 10-a 3 6't,3,,,gs x 10- 3 7.684n#9 x 10-2
~/.27 Gwf3 x 10-4 6.M 3sef x 10-13
- 0. 0 0.0 9.to2egtfx 10-2*
0.0 Xe-131m 8.39 4.91
- 2..st L.,,iy x lo s 3,W 7,,JEF x 101 3.30 aw@f""x 101 IM4 L.-.- M ^
Kr-89 g.oi 5.88 x 10-#7 Xe-138m 4,;2-6.55 x 10fg z.12 3,,gi,y x 102 4.to3.,g6 x 102 Z.'ll L.3e" x 102 g,'i'1 3.s#9 3.953 6T x 103 3.3</ W x 103 i.g6 6.seMf" x 10N Xe-133 9.76 5.72 x 102 Z H L,*S*x 103 0.0 i.M 6 95 x 10-[4
(.ob Ge'f9* x 10-18
~J.42.Ese*P x 101
- 1. bb &sf'f x 10-%2.,
0.0 4,651.55 x 10-2 Xe-135m q,y 2,@5" x 10 2 Xe-135 2.171.63 x 102 f.333e97 x 102 Xe-137 8334.87 x 10-e q t.*-3v4'T x 10-17 0.0 0.0 0.0 0.0
- 0. 0 Xe-138 t4.9 5.01 x 10-7 5f?S3,4tr x 10-*
g,q')1,9"f*x 10-15 j.qs 8.49 x 10/g 3.'t92,GP x 103 6,t6 3,,eT'x 103
%.go 3.s#6 x 103 t.42.Swetf*x 10)%.
TOTAL
- 02) Units for activities are in curies.
Rev. 15, 12/8~
O 1
~
{S-Vib
~
b f
TABLE 15.7-21 FUEL-HANDLING ACCIDENT: RADIOLOGICAL EFFECTS I
DESIGN BASIS ANALYSIS WHOLE-BODY INHALATION DOSE (rem)
DOSE (rem)
Exclusion Area Boundary
~7,[ S gx 10-2 Mx 10-2 l
(731 meters hr dose)
Q,s't Low Population Zone 9,93 g x 10-2 Jieff y 10-[I l
(2043 meters hr(2) dose)
(.30-REALISTIC ANALYSIS WHOLE-BODY INHALATION DOSE (rem)
DOSE (rem)
Exclusion Area Boundary 3.16 ba f x 10-3 WbMPP x 10-3 l
r (731 meters hr dose) 2.3//
i Low Population Zone 3.d p x 10-3 aw rf x 10-3 l
a (2043 meters day (2) dose) 7,,39 (1) Section 15.7-4 gives a discussion of accident duration times i
i t
I Rev. 15, 12/82 l
4 fh~&Wb*
[
s/f*
[ #/ /
O WQ d
ENvtRONVENT H
REFUELING AREA SGTS JL POOL l
l l
l l
LIMERICK GENER ATING STATION i
I UNITS 1 AND 2 L
FINAL SAFETY ANALYSl$ REPORT LEAKAGE PATH FOR FUEL-i HANDLING ACCIDENT l
s 1
REV.15,12/82 f
FIGURE 15.71
.