ML20107A729
Text
'
,ge
(
UNITED STATES
,,b ATOMIC ENERGY COMMISSION waswinoton, o.c. aoses F
July 16, 1973 Docket No. 50-219 Jersey Central Power & Light Company l
ATTN: Mr. R. H. Sims Vice President
'~
Madison Avenue at Punch Bowl Road Morristown, New Jersey 07960 Gentlemen:
l.
By letter of November 20, 1972, the Commission's Regulatory staff requested-
)
that.you provide the necessary analyses and other relevant data for deter-mining the consequences of densification and its effects.on normal operation, anticipated' transients, and accidents,. including the loss-of-coolent accident.-
Your response of February 22, 1973, stated that the General Electric report NEDM-10735, "Densification Considerations in BWR Fuel Design and Performance,"
December 1972, serves as your answer to our request.
l
~
As you are aware, five additional proprietary supplements to NEDM-10735 have been submitted by General Electric Company in response to questions rai. sed i
by the staff as a result of our review of NEDM-10735 and the succeeding supplements.
By letter dated April 3, 1973, we requested additional information concerning the fuel densification analyses performed for all types of fuel in the Cycle 3 core, including that supplied by General Electric Company and that supplied by Exxon Nuclear Corporation. In response, you submitted Supplement No. 3 to Facility Change Request No. 4 dated April 17, 1973.
Enclosures A and B represent the staff's conclusions on BVR fuel densifica-tion for the GE fuel and Exxon fuel respectively and provide the essential elements to be included to account for the effects of fue.1 densification in the Oyster Creek core.
Therefore, we. request that you provide the necessary analyses cnd other relevant' data for determining the consequences of densification and the effects on normal operation, anticipated transients, and accidents, including the postulated loss-of-coolant accident, using the guidance provided in the enclosures..If-the analyses indicate that changes in design or operating conditions are necessary to maintain required margins, you should submit proposed changesand operating limitations.with the analvses.
3
/
o
'Y C5+"
. ; ~;.
9604150096 960213 PDR FOIA-DEKOK95-258 PDR y
-}
~
. =..
(
4 Jersey Central Power & Light Company July 16, 1973 To permit the Regulatory staff to conduct an expeditious and orderly review of these matters, we request that you submit the analyses and additional information within thirty days from the date of this letter. It is requested that this infermation be provided with one signed original and t
j thirty-nine additional copies.
Sincerely, h
Robert J. Schemel, Chief Operating Reactors Branch No. 1 Directorate of Licensing i
Enclosures:
A - Model for Fuel Densification, GE fuel B - Model for Fuel Densification, Exxon fuel k
cc: see next page 1
4 l
s -
e e
e
(.
Jersey Central Power & Light Company 3
July 16, 1973 J
cc:. George F. Trowbridge, Esquire Shaw, Pittman, Potts, Trowbridge
& Madden 910 - 17th Street, N. W.
Washington, D. C.
20006 CPU Service Corporation ATTN:
Mr. Thomas M. Crimmins Safety 6 Licensing Manager 260. Cherry Hill Road Parsippany, New Jersey -07054 J. Lester Yoder, Jr., Esquire 206 Horner Street i
Toms River, New Jersey 08753 4
Mr. Kenneth B. Walton Brigantine Tutoring 309 - 21st Street, South Brigantine, New Jersey 08203 Miss Dorothy R. Horner Township Clerk Township of Ocean Waretown, New Jersey 08753
~
Ocean County Library 15 Hooper Avenue Toms River, New Jersey 08753 i
I 1
)
i j
{
1 A
n
=-
.- e I
ENCLOSURE A f
g u
GE FODEL FOR FUEL DENSIFICATION
., The General Electric fuel densification rccdel is described in flEDM-10735 and Supplements 1, 2, 3, 4, and 5 to tiEDM-10735 (see references 1through6). The GE model when modified as described below is considered 4
5 to be suitably conservative for the evaluation of densification effects '
in BWR fuel.
4 i
Possible effects of fueU.densification ares (1) power spikes due to,
axial gap formation; (2) increase,in. (HGR because of' pellet length shorten-ing; (3) creep collapse uf the cladding due to axial gap formation; and (4) changes in stored energy due to increased radial gap size. Similarly, the GE model for fuel densification consists of four parts:
power spike model, linear heat generation model, clad creep collapse model and stored 4
energy model. The required modifications to each of these models are/
listed below.
Power Spike Model 1
The GE power spike model is acceptable as it is described in NEDM-10735 and Supplement 1 to NE0M-10735 and modified in Supplement 5 of NEDM-10735 as long as it is used in conjunction with a maximum axiak gap size given by the following equation:
AL' =(0.965
+0.004)L
~
2 where AL = maximum axial gap length L = fuel column icngth f = mean value of measured initial pellet density (geometric) i t
0.004 = allowance for irradiation induced cladding growth and axial strain caused by fuel-clad mechanical interaction i
1 I
4
..i
~
f'
)
]
I -
Linear Heat Generation Model The following ex'pression should be used to calculato the decrease in fuel column length in determinations of the 1'inear heat generation rate:
g, 0.065 - @, t 2
where: AL = decrease in fuel column length L = fuel column leng h N
t Q = mean value of measur6d initial peliet density (geometric)
Credit can be taken for fuel column length increase due to thermal expansion, and for the actual measured length of the fuel column.
Clad Creep Collapse Model Examination of exposed BWR fuel rods (Ref. 5) and Regulatory staff calculations show that clad collapse will not occtrin typical BWR fue'l during the first cycle of operation.
Consequently, no additional creep collapse calculations are required for the first cycle of typical BWR fuel.
For reactors in subsequent cycles of operatio,n the GE creep collapse model, described in NEDM-10735 and its supplements, should be used with the following modifications:
1.
The equation used to calculate the~cnange'in ovality due to the increasing creep strain should account for the ovality change due to change in curvature as well'as for the ovality change due to change in roa circumference.
4
=.-
. s
(
2 r
's.
2.
A conservative value should be used for the clad temperature.
Axial temperature variations -in the vicinity of a fuel gap as affected by thermal radiation from the ends of the pellets and by axial heat conduction should be taken into account.
Effects from any buildup of oxide and crud on the clad surfaces should als be considered.
i 3.
The calculations should be made for the. fuel rod having the worst combination of fast r}eutrop flux and clad temperature, g
4.
No credit should be taken for fission gas pressure buildup.'
i S.
No credit should be taken for end effects. An infinitely long, unsupported length of cladding should be assumed.
4 6.
Conserntive values for clad wall thickness and initial ovality 1
l should be used. An acceptable approach is to use the two stan(6cd 1j.
deviation limit of as fabricated dimensions.
Stored Energy Model i
The GE stored energy model is based on V0 thermal conductivity and 2
heat capacity given in Section D of Reference 6, a flux depression factor 2
of 1.0, and a gap coefficient of 1000 Btu /hr-ft F applied to aach fuel rod within the hot fuel assembly. The selection of the. gap, coefficient in this model should be modified as follows.
(1) Cnanges in gap conductance due to variations in LHGR,. gap i
size (or g/D) and initial fuel pellet density should be accounted for.
d i
4 3
4 4
... s
}
4-(2)Agapconductancevs.LHGRcurvethatbasedonavailable experimental data predicts with 95 percent confidence that -
90 percent of future events wt11 exceed predictions, should be used.
(3) instantaneous densification should be assumed, i.e., pellet OD and gap size should be calculated using the following equation:
g 7, 0.965. Pc' + 20' r t
(
where: Ar = reduction in pellet radius r = initial pellet radius T = standard deviation in the measured probability distribution of pelhet density f4 = mean value of measured initial pellet density (geome'tric)
~
The gap size and pellet, 00, corrected for instantaneous densi-fication, should be used for the selection of the gap conductance vs. LHGR curve.
(4) The fuel pellet located at the most critical position for 4
r.:rmai operation, anticipated transients and postulated accident conditions should be analyzed with the densified pellet size as given by the equation under item (3).
(5)
In calculations which are sensitive to bundie stored energy, for the 43 r,cighi)oring policts in the same horizontal planc, the standard deviation used in the equation can be replaced by the standard deviation in mean boat pellet density.
e
s I,
g.
- .e -
(6) Since the assembly average stored energy is one of the most important inputs to BWR LOCA evaluation, a Tcc; 4 cal Specification limit should be imposed on maximum perm.itted asse.Tbly power.
\\
.. s O
S, 6
4 6
4 8
e h
7 I.
9 S..
4 E
i T. !.
a e
I
.s
(
)
References D. C. Ditmore and R. B. Elkins:
"Densification Considerations in BWR Fuel Design' and Performance" NE014-10735, Decerrber 1972.
2.
" Responses to AEC Questions - NEDM-10735,* "NE0M-10735 Supplement 1, April, 1973.
3.
Responses to AEC Questions NEDM-10735 Supplement 1, "NEDM-10735
^
Supplement 2, May,1973,.,
4.'
" Responses to AEC Questions NEDM-1,0735 Supplement 1 "NE0M-10735 Supplement 3 June 1973.
5.
" Responses to AEC Question NEDM-10735" NEDM-10735 Supplement 4, July 1973.
6.
"Densification Considerations in BWR Fuel," NE0M-10735 Supplement 5 ' July 1973.
7.
B. C. Slifer and J. E. Hench, " Loss-of-Coolant Accident & Emergency Core Cooling Models for General Electric Boiling Water Reactors,"
NE00-10329, April 1971.
6 4
4
.s
(
)
i Enclosure B EXE:.:UCLEM: liODEL FOR Fi:
iE!!S I F I C.^.T I C':
The Exxon !uclear fuci densification model ic described in Reference 1.
The Exxon model when raodified as described below is considered to oc suitably conservative for the evaluation cf densification offects in BNR fuel.
Possible effects of fuel densification are:
(1) povier spikes due to
,s axial gap formation; (2) increase in L:!GR because o# pellet length shorten-I ing; (3) creep collapse of the cladding duc to axial gap formation; and (4) changes in stored energy due to increased radial gap size.
Similarly, the Exxon nodel for fuel densification consists of four parts:
power spike mode l linear heat generation model, clad creep collapse model anJ ttored energy model.' The required modifications to each of these models are listed belo.v.
Posser Scike ?lodel The Exxon pover spike mocel is acceptable as it is described in Reference 1 as long as it is used in conjunction with a naximum gap size given by tae follotving equation:
til = (.955
' 0'034) '
2 uhere AL = maximum axial gap length I. = fuel coluun length p(= nean value of measured initial pellet d?nsity (geometric) 0.004 = alio.zance for irradiation induced ciadding gravtn and axial strain causad by fuel-clad nochanical interaction m
's p
2 Linear Heat Generation Model The following expression should be used to calculate the decrease
[
in fuel column length in determinations of the linear heat generation -rate.
A L = 0.965
,A L p
where: AL = decrease in fuel column L'= fuel column lengh g,
e
@= mean value of measured initial pellet censity (geometric)
Credit can be taken for fuel column length increase due to thermal expansion, i
and for the actual measured fuel column length.'
Clad Creep Collapse Model i
Examination of exposed BWR fuel rods and Regulatory staff calculations shows j
that the clad collapse will not occur in typical BWR fuel during the first cycle of operation.
Consequently, no additional creep collapse calculations are required for the first cycle of typical BWR fuel, i
1 For reactors in subsequent cycles of operation the Exxon creep collapse 2
- model, described in Reference 2 should be used with the following assumptions:
1.
A conservative value shoula be used for the clad temperature.
Axial temperature variations in the vicinity of a fuel gap as affected by thermal radiation from the ends of the pellets and by axial heat conduction should be taken into account.
Effec ts-from any buildup of oxide and crud on the clad surfaces should i
also be considered.
i 4
45h T.*
%W+'*-
W **' gutepergspeawee *4
- . eq o ge N,,, n.
,,,eg,9, e9
,p,
,,.w,,,,,
,,,'o.
I 2.
The calculations should be man for the.el rod naving tne worst combination of fast neturon flux and ciad temperature.
3.
No credit snould be taken for fission gas pressure buildup.
4.
40 credit snould be taken for cnd effects.
A.i infinitely long, unsupported lengt;i of cladaing should be assumad.
5.
Conservative values.{or clad wall thickr. css and initial ovality c.
should be used. An acceptable approach is to use the two standard deviatica limit of as fabricated dineasions.
Stored Enerny Modal taermal conductivity T.le Exxca storeJ caergy model is cased on UOg heat capacity given in p.aforence 3, a fiux giv:n in 3eference 1, UG2 2 p, depression factor of 1.0, and a gap coefficient of 1000 Stu/hr-f t appliad to cach fue'.
d within the hot fuci assr.c;iy.
The selection 'of t'le gap coefficient in this model should be modified as folicus:
(1) Changas in gap conductance due to variations in ' MG2, gap size (or g/3) aad ir.itial fuel pallet density shcuid be accounted for.
(2) A gap conductance vs. UiGR curve that bas 2d on availabia experinental data predicts with 95 perce".t confidence that 93 p?rcent of ft.ure evcnts will ?xcccc pr? dictions, should b0 used.
t J
l 7-
)
. (3)
Instantaneous densification should be assumed, i.e... pellet 00 and gap size should be calculated using the following equation:
r = 0.965 - 91,+ 26 r where: Ar = reduction in pellet radius r = initial pellet radius N
{;= standard deviation in the measured probability i
distribution of pellet density d = maan value of measured initial pellet density (geometric)
The gap size and pellet, OD, corrected for instantaneous densi-fication: should be used for the selection of the gap conductance vs. LHGR curve.
(4) The fuel pellet located at the most critical position for, /
norral operation, anticipated transients an3 postulated accident conditions should 'oe analyzed with the densifisd pellet size as given by the equation under item (3).
(5)
In calculations which are sensitive to bundle sto ec energy, the j
initial den:;ity of the 48 neighboring pellets in the same horizontal plane, should be equal to the lowest mean value of the individual
' pellet lot densities. To calculate the densified values, the ecuation under item (3) can be used substituting the lowest a
t.can pellet lot density for}i and setting tr.c 2(I value ec.uci i
to zero.
- \\
i i
s
[
.4
. =.
/
..e
(
)
l (6) Since the assembly average stored energy is one of the most important inputs to DWR LOCA evaluation, a Technical Speci-l fication limit should be imposed on~ maximum permitted assently p o.te r.
i 4
.s, O
e 1
l
(.
\\
References 1.
Oyster Creek Nuclear Generating Station, Docket No. 50-219, Supplement No. 3 to Facility Change Requ'est No. 4, April,1973.
2.
Merckx, K. R.:
Cladding Collapse Calculational Procedure,"
JN-72-23, November 1,1972.
3.
Braxfield, H. C., et. al:,," Recommended Property and Reaction Kinetics Data for Use in Evaluating a Light-Water-Cooled Reactor Loss-of-Coolant Incident Involving Zircaloy-4 or 304-ss-Clad U0 "
2
~
GEMP-482, April,1968.
e 4
muere Jan.u==,
4.
.. e i
C.(.2/$
I In reply refer to:
RO:RPB 30-219 JUL i 6 W73 Jersey Central Power and Light Company ATTN:
Mr. Donald A. Ross Manager, Nuclear Cencrating Stations Madison Avenue at Punch Dowl Road-Morristown, New Jersey 07960 Centlemen:
This will acknowledge receipt of your letter dated June 5, 1973, reporting the e.sposure of an individuril to radioactive material. This catter will be examined during a future inspection.
Very truly yours, cr:t>:' u ee t, Nafi I 4 fr.:,tg
/
Jolm C. Davis, Deputy Director for Field Operatic.s Directorate of Regulatory Operations bec: w/cpy ler dtd 6/5/73 D. J. Skovholt J. G. Keppler J. P. O'Reilly H. D. Thornburg C. F. Eason PDR Local PDR NSIC-DTIE-8 4 9 ' DR Reading pf/U DR Central Files-J [-
Incident Files RO Files RO RPB RO RO suamut >
.M.f.0.c,k,e,t,t.,:,e f,.,;, CB.9,y,,,,,,,,_,,,,Jg.Qg,y $.s.,,,,,,,
~ 7../.11../. 7. 3 '.
..].
un>
' Form AIC488 gAev. 9 33) AECM 0240
+44-se-siesM ** His
(-
.p h 3-e
. c
.t " ;
=.'
.. Jersey Contra 1l Power'& Light Company ll i
. M ADi$oN AVENU E' AT PUNCH BOWL. Ho AD e MoRRisToWN. N. J. 07960 e 539 611 t June 5, 1973 1
Mr. Frank E. Kruesi, Director Directorate of Regulatory Operations United States Atomic Energy Commiss' ion
' Washington, D..C.
20545 l
Dear Nr. Kruesi:
Subject:
Oyster Creek Station Docket No. 50-219 Personnel Exposure The purpose of this letter.is to advise you that during the performance of control rod drive. modification and replacement, an individual,. under the i
employ of an outside contractor, received a whole body exposure in excess of.
~
3.0 roms.
This exposure is in excess of the applicable limits as set forth in 4
'10 CAR 20.101.B.1 and, as such, is being. reported per 10CFR20.405.
The individual of concern was assigr.ed to a work crew performing the
~
modification and replacement of the control rod drives, and received the incremer.t of excessive exposure, while engaged in the removal of a drive under the reactor vessel.. In the performance of this specific job, the man was exposed to IcVels of radiation which ranged from 60 mr/hr to 800 mr/hr.
The following controls were in effect at the time of the incident:
The area was restricted, a Radiation Work Permit (RWP) had been issued and the job was being supervised.
In retracing the incident to determine the cause of.the exposure, the following information was determined:
E f,
1.
The individual, employed by the contractor, arrived at Oyster Creek en Friday, Apr31 27, 1973, was issued a film badge and attended an orientation course in Radiation Protection.
I 2.-
lie was assigned to a crew scheduled to perform work within.the scope of the control rod drive modification and replacement-program. The work was conducted under the supervision of contractor personnel.
l
. 3.: llis tot al accumulated exposure through May 5,1973 was.1210 mr
'as determined from film badge results.
At this time, after re-viewing his exposure, the individual was given permission to-accumulate additional exposute to a 1cvel of 1700 mr, which was g'
Jaccording to established guidelines.
h
[
'~
(
c Mr. Krucsi June 5, 1973
...Ilis total exposure on May 7,1973 was 1615 mr (1210 mr film badge
-4.
and 405 mr self-reading dosimeter) as recorded on the daily log.
i sheet.
At this time, the individual wa assigned to.a work crew I
scheduled to remove a control rod drive. The area in which the work was performed was adequately surveyed and the crew was under contractor supervision.
Aft $r performing the necessary drive work, the individual discovered 5.
I that his self-reading dosimeters (200 mr, 500.mr and 1R) had all pegged upscale indicating an exposure in excess of I rem. The job had b.cen performed in a high radiation area located under tl e 3
reactor vessel.
6.
His. film badge was immediately processed and the results indicated 1810 mr for the period May 6 through May 8, 1973 inclusive, indica-ting the individual received approximately 1400 mr while performing the work.
After evaluation of the above information, the conclusion was reached that the cause of the overexposure was twofold; firstly, the failure of the individual of concern to periodically check his scif-reading dosimeters to deterrine the aucunt of exposure he was receiving and, secondly, the failure of the contractor supervisor to,- (being aware of the allowable exposure limits) periodically check 4
the individual's exposure and to use more care in the assignment of trork censidering the man's previous accumulated,cxposure.
Immediately upon discovering that the i
i overexposure had cecurred, a eccting uns conducted betuccn the contractcx and Jersey Central Power 6 Light Company's staff to determine corrective action needed and to initiate measures of control to prevent recurrence of similar incidents.
i Corrective action taken involved the use of health physics personnel to more closely observe exposure of individuals engaged in work in Radiation Tod Permit (Rh'P) arcas. This was accomplished by having the health physics personnel perform the following:
i 1.
De aware of exposure limits for all contractor personnel request-ing entrance to RWP areas prior to admittance.
f 2.
Assure that all contractor personnel are informed as to the RWP requirements, are properly clothed, protected, monitored and record allowable exposure.
3.
Monitor and record exposures of contractor personnel at least hourly, more frequently if required, and remove any individual 4
from the area who reaches his allowable limit.
In addition, more stringent administrative rec,uirements have been imposed on all: contractor personnel to preclude the recurrence of this event.
These requirements include daily meetings to discuss work to be performed in light of-necessary radiation protection, the restriction from work in high radiation areas -of all contractor personnel who receive an accunulated cxposure of 2.0 rems,.
and the processing of. film badges daily for all contractor personnc! who are 4
e r
.,s.
(.
l Fir. Kruc,i June 5, 1973 engaged in work in high radiation areas.
It is felt that the above actions will assure Jersey Central Power G Light Company's management that a recurrence vill not be experienced.
Jersey Central Power G Light Company had prepared and implemented radiological control of personnel engaged in work during the outage, through the establishn.ent of administrativc guidelines, the maintaining and reporting of all personnel exposure on a daily basis, and the orientation of all personnel in radiation protection.
In addition, a supplemental systen of memorandum writing was instituted to alert the contractor supervisois of personnel who were approach-ing pre-established limits.
It is the feeling that Jersey Central Power G. Light Company had naintained proper administrative control to prevent an occurrence of this nature and the reason for the incident was the failure of the contractor personnel involved to observe the rules and follow the proper safety practices, h'e are enclosing forty (40) copics of this letter.
V (gy truly yours,pphh,-~ f,2
/
<.i4L
~ ~
Donald A. Ross blanager, Nuclear Generating Stations DAR:cs Attachment cc: Fir. J. P. O'Reilly, Director Directorate of Regulatory Operations, Region 1
-