ML20024G247
| ML20024G247 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 10/04/1973 |
| From: | Mayer L NORTHERN STATES POWER CO. |
| To: | Oleary J US ATOMIC ENERGY COMMISSION (AEC) |
| References | |
| NUDOCS 9102080412 | |
| Download: ML20024G247 (24) | |
Text
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NORTHERN 5TATES POWER COMPANY MIN Nt pOU S. MIN N E S OTA 95401 October 4,1973 g
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M Mr. J F O' Leary, Director d'
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Directorate of Licensing m m:
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Office of Regulation v
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I'h U S Atomic Energy Commission T/[7'jT Ms L,
,/j Washington, D C 20545 Y
Dear Mr.
O' Leary:
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s MONTICELLO NUCLEAR GENERATING PIANT N w' Docket No. 50-263 License No. DPR-22 Additional Information Concerning Supplement No. I to Technical Spectiication Change Request No. 3 On September 22, 1972 we submitted proposed Technical Specification changes which, among other items, covered Control Rod Worth and the Rod Worth Mini-mizer. At our May 17, 1973 meeting in your Bethesda offices, General Electric representatives summarized the nature of the Control Rod Drop Accident and its relationship to the limit on control rod worth stated in Technical Specification 3.3.B.3.(a).
It was resolved that a limit on rod worth should be established which applies to the present core loading as well as anticipated reloads of Monticello and similar plants while allowing sufficient flexibility for operation within the constraints of the Specifi-cation.
Since such a limit is affected by a number of parameters, it was decided that the Technical Specification Bases should discuss bounds for those parameters such that fuel would not exceed 280 calories per gram in the unlikely event of the postulated Rod Drop Accident.
l The September 22, 1972 submittal was written for the Monticello core that existed at that time.
The attached Exhibit A provides alternative wording l
for Technical Specification Bases sections 3.3.B.3 and 4.3.B.3.
This basis supports a limiting control withdrawal increment of.013 delta k rather thc.n.015 delta k as previously proposed for Technical Specification 3.3.L.
- 3. (a). Also attached is a document prepared by General Electric entitled
" Technical Basis for Changes to Allowable Rod Worth Specified in Technical Specifica tion 3. 3. B.3. (a). " This document provides additional information on the subject and is referenced in the alternative wording for the basco.
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9102030412 731004
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l PDR ADOCK 05000263 l
P PDR 1
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N O R'. ARN OTATED POWGR CC DANY In this submittal we continue to maintain our earlier position concerning the Rod Worth Minimizer; that is, that the Rod Worth Minimizer may be bypassed and its function performed by an independent individual anytime during a startup or shutdown.
Our September 22,197'. proposal increanes the frequency of surveillance required of the second individual to further assure an effective substitute for the Rod Worth Minimizer function. The revised Rod Drop Accident analysis shows that during a startup, more out-of-sequence control rods than previously predicted may exist which, if involved in the unlikely postulated accident, could cause an excess of 280 calories per gram. However, the design basis for the Rod Worth Minimizer remains unchanged from that discussed in the FSAR.
The availability of the Monti-cello Rod Worth Minimizer during reactor. startups and shutdowns since comencing commercial operation has been maintained in excess of 957..
There has been no need to excessively bypass the Rod Worth Minimizer and turn to the second individual.
To require the Rod Worth Minimizer to be available for reactor startups is inconsistent with its design basis and places unreasonable demands on the reliability of non-redundant equipment.
In the intervening time prior to formal issuance of these changes, we are conforming to the more restrictive Limiting Conditions for Operation stated herein and in our September 22, 1972 submittal.
Yours very truly, h.
[
L 0 Mayer, PE Director of Nuclear Support Services LOM/MHV/br 9
cc: J G Reppler G Charnoff Minnesota Pollution Control Agency Attn K Dzugen L
EXHIBIT A Proposed wording to replace Technical Specification Bases Section 3.3.b.3 and 4.3.B.3 (Page 84):
3.'
Control rod withdrawal and insertion sequences are established to assure that the maximum in-sequence individual control rod or
. control rod segments which are withdrawn could not be worth enough to cause the core to be more than 0.013 delta k supercrit-ical if they were to drop out of the core in the manner defined for the Rod Drop Accident. (3) These sequences are developed prior to initial operation of the unit following any refueling outage and the requirement that an operator follow these sequences is backed.up by the-operation of the RWM.
This 0.013 delta k Itmit, together with the integral rod velocity 1Lniters and the action of the control rod drive system, limit potential reactivity insertion such that the results of a control rod drop accident will not exceed a maximum fuel energy content of 280 cal /gm.
The peak f*ici enthalpy content of 280 cal /gm is below the encrgy content at which rapid fuel dispersal and primary sys-tem damage have been found to occur based on experimental data as is discussed in reference 1.
Recent bmprovements in analytical capability have allowed more refined analysis of the control rod drop accident.. These tech-niq)ues have been described in a topical report and two supplements.
(1 (2) (3)- By using the analytical models described in these reports coupled with conservative or worst-case input parameters, it has been determined that for power levels less than 107, of rated power, the specified limit on in-sequence control rod or control rod segment worths will limit the peak fuel enthalpy l
content to less than 280 cal /gm. Above 10% power even singic operator errors'cannot result in out-of-sequence control rod vorths which are suf ficient 'to reach a peak fuel 'enthalpy con-(
tent of 280 cal /gm should a postulated control rod drop accident occur.
(1) Paonc, C J, Stirn, R C and Wooley, J A, " Rod Drop Accident Analysis for.Large Boiling Water Reactors," NED0-10527, March 1972.
(2) Stirn, R C, Paone, C J, and Young, R M, " Rod Drop Accident Analysis for Large BWR's," Supplement 1 - NEDO-16527, July 1972.
f:
(3) Stirn, R C,Paone, C J, and Haun, J M, " Rod Drop Accident Analysis for Large Boiling Water Reactors Addendum No. 2 Exposed Cores," Supplement 2-NEDO-10527, January, 1973, i
1 EXHIBIT A
, The following conservative or worst-case bounding assumptions have been made in the analysis used to determine the specified 0.013 delta k 1 Lait on in-sequence control rod or control rod segment worths.
The allowable boundary conditions used in the analysis are quanti-ficd in reference 4.
Each core reload will be analyzed to show conformance to the limiting parameters, a.
A startup inter-assembly local power peaking factor of 1.30 or less, b.
An end of cycle delayed neutron fraction.
c.
A beginning of life Doppler reactivity feedback.
d.
The Technical Specification rod scram insertion rate, The maximum possibic rod drop velocity (3.11 ft/sec).
c.
1.
The design accident and scram reactivity shape function, g.
The moderator temperature at which criticality occure.
It is recognized that these bounds are conservative with' respect to expected operating conditions.
If any one of the above conditions is not satisfied, a more detailed calculation will be done to show com-pliance with the 280 cal /gm design limit.
In most cases the worth of in-sequence rods or rod segments will be substantially less than 0.013 delta k.
Further, the addition of 0.013 3
delta k worth of reactivity as a result of a rod drop and in a conjunc-tion with the actual values of the other important accident analysis parameters described above would most likely result in a peak fuel enthalpy substantially less than the 280 cal /gm design limit. However, the 0.013 delta k 1Lmit is applied in order to allow room for future reload changes and ease of verification without repetitive Technical Specification changes.
Should a control rod drop accident result in a peak fuel energy content of 280 caugm, less than 660 (7 x 7) fuel rods are conservatively esti-mated to perforate.
This would result in offsite doses twice that previously reported in the FSAR, but still well below the guideline values of 10CFR100.
For 8 x 8 fuc1, less than 850 rods are conserva-tively estimated to perforate, which has nearly the same consequences as for the 7 x 7 fuel case because of the operating rod power differences.
(4) Report entitled, " Technical Basis for Changes to Allowable Rod Worth Specified in Technical Specification 3.3.b.3.(a)" transmitted by letter from L 0 Mayer (NSP) o J F O' Leary (USAEC), dated October 4,1973.
l EX111 BIT A 3
The RR1 provides automatic supervision to assure that out of sequence control rods will not be withdrawn or inserted; i.e. it limits opera-tor deviations from planned withdrawal sequences.
Reference Section 7-9 FSAR.
It serves as an independent backup of the normal withdrawal procedure followed by the operator.
In the event that the RRi is out of service, when required, a second independent operator or engineer can manually fulfill the operator-follower control rod pat-tern conformance function of the RM1.
In this case, procedural control is exercised by verifying all control rod positions after the withdrawal of each group, pr or to proceeding to the next group.
Allowing substitution of a second independent operator or engineer in case of RMI inoperability recognizes the capability to adequately monitor proper rod sequencing in an alternate manner without unduly restricting plant operations. Above 10*/. power, there is no require-ment that the RW1 be operable since the control rod drop accident wi+.h out-of-sequenc'e rods will result in a peak fuel energy content of less than 280 cal /gm I
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T F G ll N 1 C A L 11 A S I S F0R C 11 A :1GES T0 A L L 0 '.' A !> L E R0D L' O R T ll S l' E C 1 F I E D IN EL !
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ye ar and a nalf which dment n'.a t chniqu. s and mojals being u: 2d to enal' ::o the p 'd trop /,cci dent (RDA).
The infon 'tiun in the:.,e documents has l'nen used for the d zelcp.ent of dnign apprcaches on nc., prcjects to vale t' e ccayuences of the I'DA acceptable to all cencerned.
In the case of the ci'eratin j plen+ -
,,aere safety analyses :.nJ resulting Tet.hnical Speci fications wer' previousi, tc.blished with the old approaches, the neu informaticn in the tcpical ecports was not easily applied.
The purpose of this document is to briCjo that gap by L;ing the inf on at ion and techniquct, develeped in the referenced reparts to
, rc/iu e to r nical ~;:is at.d<"
ndad Txhrical Spcci fic t' .;i'" ti,e curren t de,ign basis <.af: ty ;-hilcso;t a; pli;d to operating lants l a th. IE a re a.
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c cr, tr al rada tcaej e. jud-cnt p'.icn tica of recent EDA ucrh The 1.5"Ak v;1u. acid typically be drive 3 f ecn detailed calculation: cn al ca t-by-pl an t basis.
F0, ever, in vie.. of the f act that tnis avull not be practical tc do on all plant:, a "ucr;t case" ccepreheisive value of 1.3% Ak is rec.omanded for general and ir.;diate c;'pi'ca tion at all cperating plants.
Thir recc m r.datien ccrwican of avai'able speci fic plcnt calct-!:tient, t, a s e ti i t. obtait.ed fr
c c;: era ting ccta, to tnan taec in d: riving a 2LO cal /g.. pe; tuel >.'
4.f RDA.
bcundary f er the RCA Lith th > '.ey parameters af fecting the outctw of 1.h!
The 1.3,;4k value represents a combination of conservauive inputs which are inherenth rixed (e _
c;u cf tL C p;mler cm f ficient ccerespor. Jing
'r a
- qirnin y F-tife ~
u'<itic
. id,.ill alucys Lu '..m e rv a t i ve) Ti judgemen t inpt.t ',hich could vary significantly in the future Lut are not e, acted ci.~
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i By us ing the con ervative boundary a,mroach, it has 'bcen dMernined that t h " o r P. i i,
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thn.evc;r, this increase would be less than 1
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g ;,; ;. g g guiJuline valves of 10 CFR 100.
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III DIS CUS SIO.1 i'
A.
-Dasign Basis The d0 sign basis for evaluatin] the consequences of the RDA arc daribcd in the t-topical ~ reports (pgs. 3/4 of ref. 3).
The dif ference in the application of these bases between the ne, projects and the cperati5U jilants is in the j
dafinition of the worst single inadvertent cperator error or equipment palfunctiol L
to cause the RDA.
Previously for new projects. :nd currently for the operating
'1
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1 7 : qts, *- Rad Ucrth iM rricer (' '.:) and cpera tor were the ecdundan t centrols on rod selection so that a single failure could not cause the drop of an out-of-sequence. rod; if the R'..H were out of service, a second independent operator Nas acceptable as a subs titute.
This has not been accepted on new projects' and
{
. th i rd 7 tc~, N Psj Sewcc Centra' Cyste- -(OG), has Ler_q applied.
S'nce i
j tnis n?.i system ci:ang its at of cperation be/end -the 50l rod density ::sint.
i
-the design basis for nt:u prep;.ts has sh1f ted so that the drop of an cut-of-t u
l sequ:nce rcd at this point.is analyaod. If i t cannot be assumed that the fd!!!
i cr operator will prevcnt tne selecti0n cf an cut-of-sequcnce rod, then the wors t case accident for new projects beccmes the drop'of an out-of-sequence rod at
- ~
the point v;here the RSCS chenges its icode of operation.
' Since the contents of the topical report supplements were dev21oped in a
conja: t k r,
- 2. thc new design basis on new projects, it became necessary to 4
4'
.revien and provide otnar means for applying the new RDA results to the current ~
i Technical Specification application on operating plants. i.e.
The current 1
i i
Technical Specifications on operating plants are applici on the basis 'that the taximun reactivity.value of any insequence rtd mu:t be limi ted in crdar tu s
maintain the ccnscquences of a RDA uithin those analyzed and accepted.
The i
tr ical re:M rt: clso covared enly Nrticulc, plants at ; art;wier re; '
y/
eg.asure conditions, and since this added more variable para:1eters to an f
analysis that already contained many variables, it' Lccame recessary to l
devclop 'orst caso valuc.: that ;ould assuredly cover a wide range of
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con di ti ons.
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In this cac,c, available data frc'a calculatiens perf ormerJ for i.,
particular operatig; plants and conditions was ton.pered with the sue o
.n a.6 i r.
. i ui a t.' ; i m.i t; m;ix.
fu U,etyjul p:
reperts.
lnese para" :lers and cc: parisons are described In detail I
i b e la.1.
4 i
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C.
Paratelers Consim': red ?. Enion kuntions Used l
]:
Although there cre.ar./ input pare.matcrs to the ecd.decp accident analysis, the resultant peak fuel enthalpy is nost sensitive to the 4
falleaing input parar.aters:
I l
1.
Steady state accidcnt reactivity chape function
'l L
2..
Total centrol rod reactivity worth n
U' 3.
Maximum inter-assembly local' power peaking factor [P -normalized L
j over four bundles]
4, Delayed neutren fraction F
l E.
3rrt? rc.?c Li'.9 t; i
+
6.
L apler recc-c i t:. facdaack i
7.
."odarctor te perature For ft: sed contrcl rod drcp vclecitj and scran incertion rate, these t
[
parameters can be varied and cchined to yield a peak fuel enthalpy of 200 cal /gn.
icd drop volc ci ts. -
m.:, um d to l'e th a t j us ti f i e d by.the s te.t i s t i c a l-evaluatien in the appendi: of P.ef. (1) i.e., the maxibun velocity
- of 3.11 f t./sec. was used.
Also, the current standard Technical Specification scram times tabulated below were used in devalopuig tne screa reactivity curves for the 200 cal /gn design iic.it bcundary cerrcapr> din ic tu t';.. d t, ;ic conditicn t;2ecified r
be l o.. :
I
'of Pod Insertion Tiue fren Ee-Energization of Scrent SolenoidJalve(<.tc.)
5
- n. m 20 1.10 EC 2.0 o-L.0
_____.m
)
in order to tac t' the TE dn.i;n limi t of Ef" cal /g:. the m c - pwic La s cre cabined to UNt thrs li ic conditions.
Th.:w are ( A) the l
accit':mt t euctivi ty duroclaris tics, @) the typier r eactivi ty f eed-t an, and (C) the scra. reactivi ty feedback, li anj one of these i
conditions are not satisfied, then a tore detailed analysis would have to be periorud to establish cot.:pliance with the 280 cal /gm design limit.
l l
C.
Three r>asic conditiom 1.
Accident Reactivity Charact:ristics - Accident reactivity shape functichtotal control red reactivity worth, i n t e r-as s e.rbl y_.
local pouer rw'<irm f actor, end tM delayai neutr n fraction The sensitivity of_ the rod drop accident to the first three paraneters at cold startup and hot startup are shown by Figures 1 cnd 2 and the effect of the delayed neutrcn fracticn (twta) can it seen l:y ccmparing Figures 1 and 2 uith Figure; 3 and 4
- respeativeij.
~c d; tar,,i:w whetner or not a spe;ific conditicn will rect tha 2Y) cal./gm design linit at cald startup or hot startup, the accident reactivity chbracteristics (i.e., accident shcpe function, local peaking, etc.) for the plant being analy:cd should be matched to those presented in Figures 1 through 5.
If the accident reactivity characterptic curves are ecual to or less than U u shcun as solid lines in Figures I through 4, then one of the toree conci tions needed to conservatively ensure RDA peak fuel enthalpy equal to or less than 200 cal /gm is satisfied.
If the actual plant accident reactivity characteristics are
/
greater, a more detailed cnclysis would have to be performed.
Uhan applying thesa functions a lir car interpolation can be
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[.2 {ls
[> t a
! fl l'Q}a }'
s I O I "
l ',' ^ [*N tO C'bL local peal.ing factor and beta variables.
Sama example curves resulting frau calculations ui th u;]erating plant deta are also plotteo as dotteu lines on f igures 3 and 4 to daarc tratt tonlience,ti th thc condit kn, including the
./.,.
4
.s
.. - ~ _. _ -. _, _ _. _. _.. _, _. _.
. - - ~ - - - _ - -
If the Dnpler reactivity coef ficient, arc equal to or more negative tha" thoze given 'as solid lines in Figure 5, then onother one of the threc ccnditions needed to conscrvatit.uly ensure RDA per.k ruel enthalpy 200 cal /g:a is satisfied.
Using the BOL Doppler reactivity coef ficient will be conservative since the Doppler coef ficient always beco:res more negative with increasing exposure.
This etfeet is typically demonstrated by the exposed core data shcan as dotted lines on Figure 5, and is due primarily to the pu-240 buildup and contribution as a function of exposure.
3.
Scecn ReactiQty Feedback The scram reactivity feedback function is unique in that the total scram' feedback is not required to terminate the accident and limit cook fuel enthalpy in the time scale of interest.
The cc~5 ired J
Doppler and.01ak scram uill be : cra than tufficient to terminate the accident anc bring-the reactor cara wbcr;tical for centiel rod ucrths cf interest.
This is not :ccent to i:rply that total scran is not required for complete shutdown but rather to emphasi:c the fact that partial scram bank insertion would be suf ficient to linit the resultant RDA peak fuel enthalpy to 280 cal /gm in the time scale of interest.
Therefore, up to.01Ak, the actual plant scram reactivity feedback function Must be equal to or greater than the data presented in figuros G and 7 fer the cold and hot startup operating states respectively in order to satisfy the third of the three conditions needed to conservatively ensure RDA peak fuel enthalpy 6 280 cal /gm.
A typical exarple derived from operating plant data is also plotted cn tF s.c 'igures as dctted lines to ?c"enstrate that the coHition b
is met in actual scram perfoncance.
Additional available data was not plotted to avoid graphic cenfusion, but is sut m ized with total scram worths in Table I.
N L_
. - - -. _. = -. - -...
.t i
/ nlica tion of +he 100 gl/p Bo_undary D.
T in su:cury, all three condi tions 1, 2, and 3, an stated alove, just be satisfied in order to conservatively s tay within the 200 cal /gm daign limit boundary.
I f any of the,cenditicas are not met then a more detailed evaluation would have to be performed to demonstrate ccmpliance with the dasign limit, Likewise, given a particular set of conditions, a baximkna rod worth could be determined which could show compliance with a Technical Specification based on keeping RDA conscquences belcw the peak fuel enthalpy design. limit of 2C0 cal /gm.
It is inportant to recognize that there is no practical way to analyze all possible conditions or parametric values as they may occur during the cycle at a particular plant or plants.
However, semo evaluatiens havc been perforced to obtain typical values as shown in this document and judgetent cca be exercised to ebtain ucrst cases or perceive the effects of variations.
On this basis, it uculd be reasonable to pick som worst case values of the key parcaeters in the RDA based on the approaches used in this document and derive a rod worth for Technical specification application that could be widely used without recourse to lengthy repetitive analyses for each reactor and each fuel cycle.
Such a prccess was conducted in the course of preparing thn document, with the following results:
1.
Scras reactivity condition:
While there could be significant variation in the shape and total worth of the scraal reactivity curve, c'.tul cperation in the f uture 1: not li! aly to d:gr+ dem to the point where the net ef fect on a PDA uould be any less than that repretented by the 200 cal./gm curves of Figures 6 and 7.
- P
2.
P. ppler reactivity con F ticn:
lh leas t ef f t ttivc- (?! ) Dappler l
l fecoack has been assun.ad in the 2C3 cal /gu boundar;. cnes atated f ar this dot.u cent and it uuuid be si:'ples t to uaintain this n'sut ption in deriving a cc7rchen2ive Technical Specification applicaticn.
This censervatism would also serw to conpensate for any concern in other areas t.here variations beyond the 260 cal /gm boundary might be pos tulated in extrece situatiot s.
3.
Accident reactivity charactenstic condition:.,Lf.it is assumed that the 200 cal /gn buuadary conditions established in 1. & 2.
above represent wors t case values that no operating plants are likely to exceed, then selcction of a reccmended comprehensive Technical Specification on maximum alle.iable rod worth reduces to a consideration of the parameters associated with the accident reactivi ty characteristics discussed in C.l. above.
There are four parameters considered for this 280 cal /gm bounJary condition and it was csteblished in C.) that the closest apprcach of actual plan t operat'ng parcLeters to this 200 cal /c Lcundary was represented b.
Figure 4 It..a:- also c m bilah:d that t..
er u~.e p a ram ;ers, the acciden t rectivit / shape functicn and bcla, derived f rc".m actual plant cperating data, generally could not reach these used in calculating the 2E0 cal /gm boundary shoun in Figure 4.
Titus,
the maximo allov:able rod worth can be derived by determining the maximum P in the hot startup condition and using the corresponding solid curve.
As stated in C.1, a P above 1.30 would not be L
expected a s any plant and a mimun olic.:able red.. orth woula, there fora, be 1. 34.
This value is recc= ended for compreh_nsive Technical Specification coplication on a " worst case" basis in the absence of specific datailed ana'/ sis on each operating plant.
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.i f 4_ _c.t o_.n_._ _' c.c i_d :_n.t__E v.a_l u a_t i_o_n_
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'.n order to cu tablish ao ccn:ervative upper bound on th; nwber of fu21 rods that could fail as c r : ult of a ponulatcJ centr'el rod drop accident, it'tas assu:r.ed that a peo). fuel energy content of 200 tal/gni ucs attained.
From this analysis it yas datermined that conservative upper limit of C00 fuel reds would reach a fuel energy content of 170 cal /gni for 7 x 7 fuel.
For E x 8 fuel, this nu~ber is 850 rods, however, the total quantity of fission products rel. cased by this event is about ti.e scme sinc; the C x 3 fuel rods operate at a lo.ter pcuer.
The 1mit of 170 cal /gm for eventual fuel cladding perforation is based upon a survey of experii.: ental data and has bean tradi tionally used in Safety Analysis R2 ports.
Safety Analysis Reports written prior to tha reanalysis of the control red drop accident based on the new models and technique rcported that less th.n 330 fuel rod (7 >: 7 ). auld exrcrience a fuel en:rgy content of 170 cal /g'a.
Cased on the differen:c bet..aen th bcundary approach and the previcu; ana~jses, t.
pr.. lur iy r4 or tec o f f;i te doses are dcubled when the bcue,n rj cr en.ach is uscd.
P ce'.ar, ny q uith this increase, the offsite doses cre still well teicu the guideline values set forth in 10 CFR 100.
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TAllLE I TY PIC A L lt!'.LOAI) Ol)MitATING COllMS Ml'C I l',A ll I)ATA l
A.
In-Fruttenim Cont rgt 1tod Worth 1)i A NT_
CONDITION 1)OINT IN MAX.O.kg CYCI.M A
Cold SU 110 0 0.007 D
Cold SU UCC 0.011 D
Cold SU EOC 0.003 C
Cold SU DOC 0.005 B
llot SU '
BOC 0.003 C
!!ot SU DOC 0.005 B.
Ec turn Think Worth
- Pl. ANT Coill)lTION
. l'OINT IN
'l OTA L NEG.
CYCLE M #f A
Cold SU BOC 0.071 4
B Cobl SU DOC 0.0 19 D
Cold SU EOC 0.05i A
!!ct SU DOC 0.131 l-l B
Ifot SU DOC 0.125 B
~
llot SU EOC 0.I21 D
llot SU HOC 0.I47 D
liot Sil MOC 0.143 D
llot SU EOC 0.I41
- . linus the d roppin;; rod in tho IlDA 1 -.
..._~...
- = - - = - - -
(Centinued)
TA 111,y
- l'Yl'IC A l. IIMI.OAD O P1
- llATiljG CG111:S Niit:1.1-: Alt 1),\\TA _.
4 C.
D@we<1 Neutron P raction (p)
POINT IN 111;TA PI. ANT CONDITION CYCLE l
A Ilot SU DOC 0.0050
~A Itot SU I.:OC 0.0054 Ilot SU BOC 0.0059 D
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TO:
ORIG CC OIRER SENT AEC PDR XXX S T 1RAI, NR m
JJ. O' Leary 3 signed l
Clf.S S UNCLASS PROP 11?r0 IUPUT NO CYS REC'D DOCIIT !;0:
XXX 37 50-263 l
i DESCRIPT10;i:
ENCLOSURES:
Ler te their 9-22-73 ltr...furn add'1 info Exhibit A concerning suppl #1 to the Request for chanpc Tech Basin f or Changes to the Tech Spcen.
in Tech Specs #3.....trans the following.....
rigures 1 thru 8 (37 cys ca enc 1 rec'd) jf f s s J r s i
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FLANT HA!IE!
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