ML20039G508
| ML20039G508 | |
| Person / Time | |
|---|---|
| Site: | 05000514, 05000515 |
| Issue date: | 08/07/1975 |
| From: | Stello V NRC OFFICE OF INSPECTION & ENFORCEMENT (IE) |
| To: | Moore V Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML111090060 | List:
|
| References | |
| FOIA-80-515, FOIA-80-555 NUDOCS 8201180338 | |
| Download: ML20039G508 (21) | |
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Directcr for Light l'ater ite ctors, Group 2, RL SER II;FUT TO PEBBLE SPRIllGS NUCLEAR PLANT Pl:nti: t:
Pebble Springs !!uclear Plant D.ci.t'. i;.;- Sir s :
50-514 and 50-515 Licensing str p.:
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Core Perim m: French
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1,up'.t t i, H H Ct '.. id Enclosed is the Core Performance Branch input for tha fuels Section (4.2.1 and 4.4), iluclear Design Section (4.3) cud Accident Analysis (Chcpter 15) for the Pebble Springs fluclear Plant.
These are in accordance uith the latest version of the Standard Review Plan.
- f. ' / _
Victor Stello, Jr., Assistant Director for Reactor Safety Division of Technical Review Office of Nuclear Reactor Regulation
Enclosure:
As Stated cc:
S. Hanauer F. Schroeder A. Giaduusso C. Stable W. I'cDonald L. S. Rubenstein D. F. Ross V. Stello l
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4.2.1 and 4.4 The Pebble Spri,ng: Nuclcar Plant fu:1 r: 2, designed cnd fabri-cated by Babcock cnd Wilcox Company, consist of cylindrical uranium dioxide fuel pellets placed in zircaloy-4 tubes (referred to as the cledding).
A radial gap renains between the fuel pellet and the cicdding inner diancter to allow for in:re.:rce in pellet dimencions "S:n operation at por:r.
Thesc incrceter crc c'v 2 to irrndiction
- - lli:.1; c;.2 :~nc: z_1 t., _:a.icn of thc.
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....lu;.: e.lro e::icts
. 11.c. f u.1 12." - : : 1..
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- cer ffericn products uhich will Lc relecsed upon irr... ; on.
The fuel rod is scaled by velding end caps onto the zircaloy tubing.
Before the final veld, the fuel rod is pre-pressurized -ith helium fill gas to a high pressure.
This pressure is high enough to retard the inward creep rate of the cladding onto the fuel pellets but low
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enough so that the internal pressure of the fuel rod never exceeds the external coolant pressure on the fuel rod so that the cladding is always in compression.
,The fuel rods in the Pebble Springs Nuclear Plant are arranged.
in a 17x17 lattice array and are supported laterally along their length by eight spacer grids.
These grids maintain the lateral spacing between the rods throughout the design lifetime of the i
assenblics.
These spacer grids are constructed from strips of Inconel 718 which are slotted and fitted together in an " egg crate" fashion.
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r.:.ch fuel rod asser.bly contcine guide 4.nd' inrtrunn.: thitbles.
The guide thi.Lles are structural ner.Lcrc~vhich c13: pr:: :c. thcnnels for tha neutron absorber rods, burnabic poison rods or neutron cource ascenblies. They are fabricated from zircaloy-4 tubing.
w Some of the dimensions and design parameters for the Pebble Springs INelct: ricnt are given in T:bic I.
A1ro chorn for cc 7.cr!ren cre th e re -- r rrmtere fcr the Orence Dit 1 INc2 ccr T1rn:. 10cs c fuel is
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G: * ;;; i:. whic.. 2c. c.: ci:;&tcr
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e et n Ac.:cr :.n.icun lincar heat genut. tion rc.tc rhich uili rc;. alt fn 1crer fuel temperatures. This in turn will result in less fission gas re.
Icase from the fuel during the life of the fuel rod. Also, the lower heat flux results in a larger therra1 margin to DNB than in the 15x15 design.
The Oconee Unit 1 Nuclear Plant has accumulated one cycle of -
burnup and has experienced no significant fuel related pr'oblems.
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Table I ~
Co parison of Ecbcock and Wilcox 15x15 and 17x17 Designs Pebble, Springs Oconee-1 Fuel Rod.
17x17 15x15 l'urber 4
c rfoe Ticeter. in
.379
.430
-.' t. t.1 Dr.l., f.
.C^S
.007 Ult 6 ni ^.=:cs, in
.0235
.C*C5 Citd Heterial I.ir cc.loy -4 lirc:.ity-4 i
luel 1:aterial, Density, % TD Sintered UO, 94.
Sintcred UO, 93.5*.
2 2
Fuel Pellet Diameter, in
.324
.370 I
Maximum Thermal Output, kw/ft 12 17.6
~ 412,800.
534,440 Maxitum Heat Flux, BTU /hrft Maximum Centerline Fuel Tempcrature at 100% Power, *F
- 3,000 4,250-e 4
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1 T.ecent expcrier.cc with uranium dicxide fur:
r. rhown that it will dcnsify upon irr: diction te 2 dentity hf cher thnp that C
to chich it was manufactured.
This type of densification differs fros that which occurs due to ther=al effects in which a radial der. !ty gradient is fer cd in the fuel as the result of material Ligratica up the ther=;l trtdicnt.
The nc.0 y if tntified dcnsification
' f rm dir ' ' rn e c rf f3 ce tien U.'
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c urc6 ty t he 1:: dir.tica c,f C
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of the fuc'l pellet with corresponding changes in the pellet rL61us and length.
There are three principal effects associated with fuel densification:
(a)
A decrease in the pellet length will c,ause the linear heat cencration rate to increase by an atount in direct proportion to the percentage decrease in pellet length..
(b)
A decrease in the pellet length can Icad to generation of axial caps within the fuel column, resulting in increased local neutron flux and the cencration of a local power spike.
(c) A decrease in the pellet radius increases the radial clearance gap between the fuel pellet and fuel rod cladding causing a decrease in the gap ther=al conductance, and consequently in the capability to transfer heat across the radial cap.
This decrease in heat transfer.
capability will cause the stored energy in the fuel pellet to increase.
A decrease in radial gap conductance also will degrade the heat transfer capability of the fuel rod during various transient and accident conditions.
In summary, the effects of fuel densification cause the fuel rod to contain more stored energy, increase the linear heat generation rate of the pellet, decrease the heat transfer capability of, the fuel rod and crcate the potential for a local power spike in any fuel rod.
To assess the safety implicaticns of fuel densi,fication, all of these effects must be evaluated for cach reactor under all codes of reactor operation.
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3 Babcock and Wilcox has proposed to the staff a densification redel to consider the kinetic densification of B&W fuels (1). This codel is now under review by the staff.
Babcock and Wilcox uses the TAFY fuel rod thermal perfor=ance co=puter program to calculate both fuel tc=peratures end fuc1 rod interr:1 pres, cures.
This codel
- c - - the crur tien of f rcrzr*r eour <%-
'ficction to a fuel dentity 1
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of the tit orcti e: 1 dcr. it;. ci U3.
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- .L the cci'q sc -
of the fuel rod cladding into the axial gcps"uhich usy be formed.,
This collapse is caused by the inward creep of the cladding due to the combined effects of fast neutron flux, te=perature and external coolant pressure.
The creep increases the ovality o5 the cladding to the point where the cladding shape becomes unstable and the cladding collapses.
The applicant is required to perform a calcu-lation in order to predict when the cladding for the Pebble Springs Nuc1 car Plant fuel could be expected to collapse.
The staff has reviewed the Babcock.and Wilcox calculational method.
A discus.sion of this review is given in Referenc~e (2).
A generic surveillance program is part of the safety evaluation of the new Babcock and Wilcox 17x17 fuel design.
Babcock and Wilcox has committed to placing two precharacterized 17x17 assemblics in Oconce 2, Cycle 2 ( 3).
The precharacterization will consist of the following:
1.
Length, diameter and cass of each fuel pellet (8 rods)
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Fuel stoichiometry, enrichment and resintering characteristi s
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Tuel cercLographic exanination' and thenical cnalysis 4.
Clad spiral and ifnear profilometry
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5.
pellet stack length, overa11 rod length, rod backfill pressure 6.
T.od radiography, fuel assembly length
. In addition, chcrc:terictics of cctricted astctblic vill be tr - - ' - 'er tc c -" 3 ccdf nr urin-rl+ rcr.e er
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- cnt. ::rdcn citel i. Jintion. U tt c:2nts to et L2Cc ;;4 1u-
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Irret.
In c6dition to this progrcm, the sLcif rcquircc, cs the finci stage of the verification process for the new fuel design, that 100%
of the peripheral rods on 100% of the fuel assemblies be visually examined in the first two Babcock and Wilcox reactors to use this fuel.( }
Pebble Springs is one of those reactors which are required to participate in this surveillance program. Depending on the order in which B&W plants with the new 17x17 fuel design come 'on-line, Pebble Springs might not be required to perform the detailed in-spection.
It must, however, have the capability to do so.
Evaluation Findings The analytical models employed by the applicant have been shown to be acceptable by comparison with measurements on fuel rods I
which have been subjected to reactor operating conditions.
These models, described in topical reports, are, based on data for fuel.
similar to that proposed for use in Pebble Springs.
These analytical models, which have been reviewed in detail by the staff, provide accept.,bic assessments of the anticipated fuel rod behavior.
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On the basis of our revicv of the " proposed analytical codels and the confirmatory results from tests on ' irradiated fuel rods, we have concluded that, -(1) the fuel rod mechanical design vill' provide acceptable engineering safety margins for normal operation, and.(2) the effects of densification will be acceptably accounted for in the fuel design.
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t References 1
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"Ecbcock and* k*ilcox 1:odel. for Fredic It.;, in-ncactor Dencification,
t/.U 10083 P. April 1975.
2.
Letter from V. Stello, Assistant Director for Reactor Safety to V. }!oore, Assistant Director for Light k'ater Reactors, Group 2L "A Generic Review of the B&W Cladding Creep Analysis Topical Report L.u.'-10084 (TAR-970)," nu;;uct 9,1974 3.
Lu.tcr frc: ::r. J. T. :: alley, ::nn:ger, licencing I:.5 cock & Wilcox to 1:r. V. Stell'o,.'.;;iste.nt D'r c eter f t Z uct r f f ty, USNRC..
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t 4.3 Euclear Design Our review of the nuclear design of the Pebble Springs Nuclear Plant vas based on information supplied by the applicant in the PSAR and amendments thereto (vir. rerponses to rteff cuestionr), d're::rsions with the reactor
( u.i r r.c r. ?;tecck & Uilcer (El"), v rieur t c 't:1 rcrorts cupplied 1
- c, L Lnrtup reper:r for ept: ti:c.
Lit..:.
".he I c htic "tfch t'rcr 'brk C
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Washington Public Power Supply System Nuclear Froject No.1 (WPPS-1) reactor.
The reactor will be operated in a canner si=ilar to that of the Rancho Seco Unit I rcactor, a 177 fuel asse=bly plant designed by B&W.
4.3 1 Desinn Bases The design bases presented for the nuclear design of the fuel and reactivity contr'ol systems are satisfactory and comply with all applicable
. General Design Criteria of 10,CFR'50, Appendir A.
f 432 pescription 4 3 2.1 nuclear Desien Description 4
Descriptions of the fuel assembly enrich ents, physics of the fuel f
burnout process, burnable poison distribution, soluble boron concentrations, i
delayed. neutron fractions, and neutron lifetimes have been provided.,
The values presented for these parameters neet the design bases and satisfy the applicable sections of the General Design Criteria.
9 4 3 2.2 Pcuer Dirtribution Ue have reviewed the cetheds used by B&U to calculate power distributions for both steady state and transient conditions.
The : Joe ce=putation tool it iLG-7, a diffusion theory code with indo:.trf-wide 'usare.
Ue have reviewed
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une in thcre
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.'.r 71.r r e prc.ce:'ert - ' c ci-O r -
'n :th:rs e,;; in the industry.
.... ci.tr.u; rc; ort for L;.;..;
.cc u.... A ;.tncc. e;.4c.risons between cileul:ted cnd sec ured pouer distributic:.s for a reactor a B&E reactor.
These cc parisons showed that total peaking facters were predicted to within
--7.05.
This value is within the 7 5% nuclear. uncertainty factor which is applied to calculated peaking factors.
In addition, 'our consultant (BNL) has performed an independent audit calculation of heat generation rates for. the BOL (equilibrium xenon) power distribution for Ra,ncho Seco 1.
The rese:ts for peak'ing fac' tors agreed with these calculated by B&W within ^'3 5%.
- 0) the basis of our review, we conclude that the applicant has made suitable predictions of core power distributions.
4323 Reactivity coerricients Comparisons of calculated and ceasured reactivity coefficients are precented in the Rancho Seco Unit 1 startup report Isothermal temperature coefficient and moderator coefficient were censured at zero power and power Doppler and Moderator coefficients at various power levels.
Calculation d
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. and ceasurcment agreed to within
'l x 10-$ ek/k/ F at =cro poucr.for both coefficients.
The measured moderator coefficient at near full power (91.4%) was within -0.5 x 10~I ok/k/ F of the calculated value. The ceasured power Doppler coefficient at near f ull power agreed to within --
-5 4 x 10 ok/k/7FP with the calcq}ated value.
On the beris of this Good cgrtc: ctt tot-
.<;rurt.cnt and calculttio T:r en c:::-ti:11; timil:.
r cretcc, vs c:n:1ude that t he applicant har rede cuittb1c Tredictions of the rc:ctivit: c c:fficicntr for the Tchble f rir c ruelcs-Plent.
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4.3 2.4 control Eaouirements To allow for changes in reactivity due to reactor heatup, load following, i
and fuel burnup with consequent fission product buildup, a significant amount of excess reactivity is built into the core.
The applicant has presented infor=ation on first cycle reactivity control distribution for Pebble Springs, which is operated in the " feed and bleed" mode.
Soluble boron is used to control reactivity changes due to:
. coderator deficit from ambient to operating temperature
. equilibrium xenon and sa=arium buildup
. fuel depletion and fission product buildup throughout cycle life (that part not controlled by lu ped burnable poison)
. transient xenon resulting from load following.
Regulating rods are used to control reactivity changes due to:
. moderator deficit 'from HZP to HFP
. power 1cvel changes (Doppler) y
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t-Lucped burn ble poison rods cre used for radial flux shaping and to
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centrol_ part of the reactivity change due to fuel burnup and fission product tuildup.
Ptrt length control rcd: trc
- d to.;intain cn cxially a l:
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.dequ-tc contr:1 croists "i-
- r. 7 2 ict:.t L:; ;r esenttd C;tt to : a
.u a n.. :.., ;..<. L
... ;. s.. _. 5... ; u.; n..._.. ; _... :a 1 crin rol to Trovide a chutdcun ;:
'1 0.99 during the initial and' equilibrium fuel cycles 77 with the most reactive rod stuck out of the cere.
Cc parisons between calculated and ceasured rod worths have been presented for the similar I
Rancho Seco Unit I reactor in the startup report for that unit.
These comparisons showed a deviation of ~ 6% between measured and calculated values with the ceasured values being smaller.
The stuck rod worth was censur'ed to be about 11% lower than the calculated value.
The rod worth available for shutdown was -' 3 5% lower than predicted - well within the 10% uncertainty allowed for this value in safety analyses.
The soluble boron system is pable of shutting down the system and
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of maintaining it in the cold shutdown condition at any time during core life. This condition satisfies the requirements of General Design Criterion 26.
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On the basis of cur review, which has included the comparisoft between cr.lculated and measured rod worths, we conclude that the applicant's a secccc,, of reactivity control requirements is suitably conservative End thst adequate negative reactivity worth has been provided by the
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ne r-O rst'vity P~'i k
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L-s.cics a. J.; der; into twc crea;,s -
centrcl rods and chutde:n (or :cfcty) rods.
Lotd changes will be made with the control rods and/or the soluble boron system.
P.od insertion will be controlled by power-dependent. insertion limits given in the Technical Specifications.
These limits ensure that:
- 1) There is sufficient negative reactivity available to permit the rapid shutdown of the reactor with ample margin.
- 2) The worth of control rods that might be ejected in the very unlikely event of f.ailure of a pressure barrier in a control rod drive techanism will be no greater than that which has been shown to have acceptable consequences in the safety analysis.
- 3) The overall peaking factor does not exceed the limiting value used in the accident analysis.
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k'o have reviewed the calculated rod worths and the ccthods used by B&W to obtain the worths.
Our consultant, BNL, has performed independent calculations of rod worths for the similar Rancho Seco Unit I reactor.
The E!?L criculation agreed with the B&W values to within 2% for the re; ;1c;ints groups.
L'c havc Liso rcvicwcc the c:;parin: L:tr.::. cd culated
- rd rt r cd vclues (cce Section b.3 2.2 cle"=).
On the basis of our review, we have concludcc that the rod grcupings prcposed for Fetble Springs satisfy the requirc:ents for safe shutdown and power distribution control and that values for ejected rod worth and stuck rod worth are sufficiently conservative.
4.3.2.6 Stability The stability of the reactor to xenon-induced power oscillations and the control of such transients have been discussed by the, applicant.
The reactor is always stable to azimuthal oscillations, but under certain conditions (BOL, full power, equilibrius xenon) may experience non-damped axial oscillations.
The stability of 177 fuel assembly B&W plants was investigated during startup tests for the Oconee Unit 1 reactor (2)
A diaSonal (combination s
of axial and azimuthal) oscillation w's inducej at 75% FP and the reactor a
response was monitored for ^ 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
The aztmuthal component of the oscillation was damped but the axial component' was divergent.
At s iO hours s
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4 into the transient, the part length rods were used to suppress the axial imbalances which was reduced to near cro whcre it was kept.
No significant diffcrcncc: in scrforstnec a:'c expecid f ar 2^! f;:1 r. :c bly corcs.
On t F s M rir of thi r c.--
r.rtrEtic o ( f ti.c s~r _ tr.:.1 stcbility of a
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i o rupprccc c::ici
._ ilcr < ::. n';d tM :.b..: : + ; cf t '-
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cerclude t hrt the re teter vill not experience uncontrolled cecillatien.
4.3 3 Analvtical ifethods k'e have reviewed the analytical tethods used by E&W to perform core design. The major design tool _ is PDQ-7 as diffusion theory code with industry-wide usage.
Cross-sections for use with this code are prepared in a canner similar to that used by others in the industry
. Comparisons between calculated and measured design parameters have been made during startup tests on six reactors designed by B&W.
In all cases, the comparisons l
have been satisfactory.
On the basis of. our review, we conclude that the.
analytical methods used for the design of Pebble Springs are acceptable.
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-B-REFERE? ICES 1.
Sacramento Municipal Utility District, Rancho Seco !!uelcar Generating Station, Startup Report, March 1975 (Docket 50-312).
2.
Duke Power Cc pany, Oconce !!uclear Station Unit 1, Startup Report,
!!cyc.ber 16, 1973 (Decket 50-269).
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15.0 ACCIDENT ANALYSIS 15.1 Uncontrolled Centrol Pod Group k'ithdrawal L*c have reviewed the analysis of the rod withdrawal transient at low power (startup accident) and at full power.
We have reviewed the rcnge of parameters ass Imed in these analyses and the results of the calculitacns and we conclude that the annlyses are Latisfactory.
The i::: ricnte are terrin ted by the nerrti.*e Dopplcr ccefficient, th: hich pressure trip, or the nuclear everpouer trip.
The desi n everptwer ccndition is not reached in any transient and the peak pressure ne('er execeds allowable limits.
15.2 Centroi Bod Misoceration Control rods cay be misaligned from their group average by as much as nine inches without appreciable effect on power peaking factors.
If a rod is cisaligned by more than nine inches, it is defined to be a dropped rod. This definition covers both the action of dropping a rod and of sticking a rod.while moving a group.
The maximum worht of a rod which may be dropped while operating at full power is 0.50% a k/k.
L'e have reviewed the analysis of the dropped rod accident.
The most severe accident occurs at EOL when the moderator and Doppler l
1 temperature coefficients have their most negative values.
The initial
. power decrease is followed by a return to full power as the reactivity decrease due to the dropped rod is offset by the temperature decrease.
Startup tests at Rancho-Seco Unit 1 ha e shown that heat generation rate ani DNBR limits are not exceeded at full power when the cost reactive rod is drcpped into the core.
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On the basis of our revicw, we conclude that the di cursion of the control rod miscperatien transient is adequate.
15 3 ~ Bod Eiection Accident-This de:icn basis accident is assumed to be caused by the physical
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nc'Epplicant's propored Technical Specification: to -1. Z o h/h at zero pcuer and 0.65% sk/k at full power.
Uc have reviewed the rod-ejection analysis presented in the PSAR.
The requircaents of Regulatory Guide 1.77 are met.
Analyses were performed for zero and full power conditions at BOL and EOL.
The limiting case is that a full power at BOL.
The environmental consequences,of the postulated.
accident are shown to be acceptable.
The applicant has not described.
the consequences of ejecting a rod.of 1.0% 4 k/k worth at zero power.
We will require that this analysis be submitted at the operating license t
stage. Based on the results for similar reactors, we expect that the consequences will be less severe than these for a 0.655 a k/k rod at full power.
g Three-dimensional effects are treated in the analysis by assuming a t
larger than normal radial peaking factor (to account for the effect o'r j
the ejected rod) with the design axial peaking factor.
Birkhofer, et al(1) l I
have reported the need for three-dimensional time-dependent calculations to I
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predict correctly peak flux and tesperature distributions for super-prompt-critical reactivity excursions.
B&W has submitted consents on this article, observing that the Birkhofer article addressed a BWR which has several core -
features that tend to make three-dimensional effects more important than in a FUR.
imong thcsc are the magnitudes of the feedbtek eccfficients, the rec;etrical derign cf the control roc, and the reactivity ec:arol echcne.
HU furthcr cbrerycs t'.. t the red =rth f;r the carc : :: - ; ;,: 1.67% 4 h/k, rhich is nore than a fector of two 1crcer than the Technical Erecification limit for B&W reactors.
It is to be expected that three-dimensional effects would increase in i=portance rapidly as a function of rod worth.
- Also, the two-dicensional problem analyzed by Birkhofer had an axial peaking factor of unity as opposed to a value 1.7 used in B&W analyses.
Application of this factor to the two-dimensional results will bring them in line with those of the three-dinensional calculations.
We agree with the comments of B&W regarding the weaknesses of the Birkhofer article. Unti.1 full three-dimensional time-dependent calculations
- are performed, however, we are unable to ascertain the magnitude of the uncertainty invovled in the synthetic three-dimensional treatment performed by B&W. Meanwhile, we accept the analysis of the rod ejection accident o'n the basi's that the computed maximum enthalpy of the hottest fuel rod is of the order of 170 cal /go, which is far below our acceptance criterion of 280 cal /cm.
It is very unlikely that residual three-dimensional i
effects could cause the 280 cal /gu value to be exceeded.
r EEFERFHCES 1.
A. Birkhofer, A. Schmidt, and W. Werner,' thwicar Technolosrv, 24, pp. 7-12, October 1974.
2.
Letter, J. liallay to V. Stello, dated Fcbruary 5,1975.
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