ML20213D948
| ML20213D948 | |
| Person / Time | |
|---|---|
| Site: | Columbia  |
| Issue date: | 12/01/1981 |
| From: | Rubenstein L Office of Nuclear Reactor Regulation |
| To: | Tedesco R Office of Nuclear Reactor Regulation |
| References | |
| CON-WNP-0438, CON-WNP-438 NUDOCS 8112280013 | |
| Download: ML20213D948 (10) | |
Text
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DEC 1 mai
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f HEMORA!!DUM FOR:
R. L. Tedesco, Assistant Director for Licensir.g. DL FR0ft:
L. S. Rubenstein, Assistant Director
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for Core and Plant Systems, DSI
SUBJECT:
SER INPUT FOR THERMAL AflD HYDRAULIC DESIGN OF THE CORE FOR WNP-2 POWER PLANT 3
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1 Plant Nane:
WNP-2 i
Docket flumber:
50-397 Q. l l' / ',s j
NSSS Supplier:
General Electric OL9,
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Licensing Stage:
OL-FL 4
-3 Responsible Branch:
SSPB
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Project Manager:
R. Auluck k
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3' Review Status:
Complete (with open item) s.'-
l Requested Co..pletion Date: October 12, 1981
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- y Enclosed is the draft SER for the U!!P-2 themal and hydraulic design which was described by the applicant in Section 4.4 of the FSAR. The scope of the review included the design criteria, implementation of design criteria as presented by the final core design, and the analyses of core themal-hydraulic perfomance.
We found that the applicant's themal-hydraulic analyses were performed using approved methods and correlations. He conclude that the proposed themal-hydraulic desion for UNP-2 is acceptable. liowever, the operating license should be conditioned to the following restrictions:
1.
single loop operation is not pemitted unless supporting analyses are provided and approved; 2.
operation beyond Cycle 1 is not pemitted until stability analysis is provided and approved; 3.
operation in a natural circulation mode is not pemitted while we continue our generic evaluation of themal-hydraulic stability for BWRs; and l
l 4.
the core flow should be checked at least once per day and the average power range r:onitor flow biased scram calibrated at least l
_once per month to account for possible effects of crud deposition.
M B112280013 811201 j
l~M ADOCK 05000397 W
1 OFFlCE >
._o DATE) mc rom 3u o>sm,scu c:o OFFICIAL RECORD COPY t.m e-mw l
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, The above restrictions should be incorporated into the proposed Tech-nical Specifications by the LGB except for Item 2 which should be incorporated as a license condition.
In addition, the following open item should be resolved prior to issuance of the operating license:
- The operating limit !!CPR as calculated by including the ODYN methods must be provided for review and approval.
d Prior to release of the SER, the applicant should provide a written comitment to evaluate the Loose Parts Monitoring System (LPMS) in accordance with the Regulatory Guide 1.133, Revision 1 (fiay 1981) on a schedule specified by the applicant. The confomance evaluation report should emphasize the procramatic aspects such as establishing the alert level, the operator training in the purpose and implementation of the LPf1S, and diagnostic procedures used to confim the presence of a loose part.
%fgYnal n'?dD L. S. raba:u '
L. S. Rubenstein, Assistant Director for Core and Plant Systems, DSI
Enclosure:
As stated cc:
R. Mattson DISTRIBUTION D. Eisenhut Docket Files J. fiiller CPB r/f D. Skovholt Plant r/f R. Auluck S. Sun R. Bottimore L. Phillips P. Triplett C. Berlinger D. Fieno L. Rubenstein R. Meyer OFFICE >
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s DEC 1 1987 THERMAL HYDRAULICS SECTION - CORE PERFORMANCE BRANCH SER INPUT WNP-2 POWER PLANT 4.4 Thermal Hydraulic Design Evaluation The thermal-hydraulic safety design bases for WNP-2 can be summarized as follow:
1.
No fuel damage occurs as a result of moderate frequency transient events.
Specifically, the minimum critical power ratio (MCPR) operating limit is specified such that at least 99.9 percent of the fuel rods in the~ core are not expected to experience boiling tran-sition during the mos't severe moderate frequency transient events.
2.
fhe tore and fuel des ign h isis for sindy-s tate cpera tinn, i.e.,
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.5 fined to pru/ide rxgin %t',cen i: ly s a te c;m ra t i..g cor.d i t i vas
.i any P;el a we t iditicn to 1
acce - date ':ncar.iinties and to as%re l'ut J.o fuel t'
.ge reculls evan during the Worst nodarate frequency ccansicnt ccodition at anytiice in life.
l 3.
No undmnped oscillations or other hydraulic instabilities should occur for normal operaticn nor for the most severe ocderate frcquency transient event.
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2-l A summary of the thermal-hydraulic parameters for WNP-2 is given in Table 4.4.1.
A comparison with the parameters for Hatch-2 core design is given for reference.
The Hatch-2 core design has been previously approved in the Safety Evaluation Report issued in June 1978 and is an operating reactor.
The primary difference in core design between WNP-2 and Hatch-2 is size.
Hatch-2isaBWR/4coreandWifP-2isaBWR/5 core.
Both use the improved 8 x 8R fuel assemblies.
General Electric Thermal Analysis Basis (GETAB) is used for WNP-2. The figure of merit chosen for, reactor design and operation is the critical power ratio of the critical bundle power to the cperating bundle power.
This mathod has been previously approved and found acccptable.
The applicant inlicates that i:.cipient center.elLing of the ur: nit.m dioxide pellet occurs in the range of 19 to 21 kilcratts per foot; this is higher than the peak linear hcat generation rate during any abnorcal operating transient.
The operating limit for peak linear heat generation rate, 13.4 kilowatts per foot, results in an acceptable maximum linear heat generation rate of less than 17 kilowatts per foot during transients.
The MCPR limit originally proposed was based upon calculations using the REDY model described in NED0-IC302.
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TABLE 4.4-1 l
Thermal-Hydraulic Design Parameters l
I Hatch-2 WNP-2 (218-560)
(218-624) i l
Design thermal output - MWt 2,436 3,323 i
Final feedwater temperature (FFWT) - F 420 #
420 Steam flow rate at FFWT - 106 lbs/hr 10.47 14.30 Core coolant flow rate - 106 lbs/hr 77.0 108.5 Feedwater flow rate - 106 lb/hr 10.44 14.2 j
Steam pressure, nominal in steam down - psia
. 1,020 1,020 Steam pressure, nominal core design - psia 1,035 1,035
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Average pcwer density - kw/ liter 49.15 49.15 Max ~imum linear thermal output - kw/ft 13.4 13.4 Average linear thennal output
- kw/ft 5.38 5.4 Core total heat transfer area - ft2 54,879 74,871
-Fuel type 8x8 8x8 j
Water rrds per bandle 2
2 Core inlet er.thalpy at FfWT - Stu/lb 526.9 527.6 i
Core m?.xin':.n exit void within assa:.blies - %
76.3 76.0 Core average void, active.ccolant - %
42.2 41.8 2
15.82 15.824 Active coolant fini area per asscuSly, in Core average inlet velocity, ft/sec 6.6 6.88 Total core pressure drop - psia 73.9 24.74 Core support plate pressure drop - psia 19.46 70.32 Average orifice pec3sure drop - psia Central region 8.0 6.03 Peripheral region 16.52 16.54 Nur.ber of fuel rods per bundle 62 62 Rod outside diameter - in Fuel rod 0.483 0.483 Water rod 0.591 0.591 Active fuel length - in 150 150 Rod pitch - in 0.640 0.640 I
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The results from the tests performed at Peach Bottom-2 BWR revealed that in certain cases the results predicted by REDY are nonconservative.
Therefore, we require the applicant to use ODYN methods to analyze the a
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following transients:
(1) feedwater controller failure-maximum demand
)
with and without bypass, (2) generator load rejection, and (3) turbine
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trip with and without bypass. We will review and verify that the applicant has properly calculated the steady-state gperating limit for the minimum critical power ratio.
Recent BWR fuel design changes which effect stability include decreasing the rod size and increasing the gap conductance because of pre-pressurization.
l As a consequence, the maximum decay ratio for most BWRs increases and beccies larcer than 0.5, rhich is the original C_ aral Electric design i
criteria for F'lR stability.
Therefore, GE nnt prgosas a 6. cay ratio of 1.0 for their celieria.
The staf f has not agreed that tha proposed criterion of a 1.0 decay f
ratio calculated by using FABLE code is ecceptable.
t To further evaluate this criterion and other stability criteria, we are performing a generic study of the hydrodynamic stability characteristics i
of light water reactors under normal operation, anticipated transients, j
and accident conditions. The results of this study will be applied i
to nur revicw and acceptance of stability analyses and analytical i
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'J.I in "e.e by t he rcactor
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!urs. 1 Le ':.:P-2 s f.:ib i lii.y cnilysis i.*ul; d in a t wi; i decay i ' io of 0.7 f'r i'c cod of-tii s t cycle.
TLe staff has approved fer opc-catinn previers core designs h4ving i
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. calculated maximum decay ratio values as high as 0.7 for the initial cycl e.
Therefore, we conclude that WNP-2 core design stability is acceptable for Cycle 1.
However, in order to provide additional margin to stability limits, na,tural circulation operation of WNP-2 will be prohibited until our review of these conditions are completed. Any
. action resulting from our study will be applied to WNP-2.
We will condition the operating license to r,equire that a new stability analysis be, submitted and approved prior to second cycle operation.
Also, since no analysis has been presented for minimum critical power ratio limits or stability characteristics for single loop operation, we will require by Technical Specifications. ti at single loop operation will not be parmittcd until supporting rnalyses are provided and appreved.
Critical pcuer tests have bcen run on prc,totypical 8 x 3 fuel bandles with tuo uater reds.
Test data for cosine axial heat flux shapes indicate that the water rods do not affect the GEXL capability of predicting the bundle critical pcwer performance for bundle radial peaking patterns typical of 8 x 8 retrofit fuel. We have previously found that the GEXL data base, which includes top and bottem peaked axial heat flux distributions, conbined with the two watcr rod data decrostrate the adequacy of the GEXL correlat.icn to predict critical pr.wr in both 8 x 3 and 8 x 8 rc trofit badles.
L'e hwe pct..*i wsly
- c...n:{!cd t hat the CrvL rorreli.t hn is ci o picble for
- Sh 3 x 3 cnd d x 8 retro fi t f.ml c ppl ii a u. n.
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I Crud deposition causes gradual flow reduction in some light water reactor cores. However, measurement of core flow by jet pump pressure drop and core plate pressure drop will provide adequate indication of such flow reduction, if it should occur.
Technical Specifications will be modified to require that the core flow be checked at least once every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and that the average power range monitor flpw biased scram be recalibrated at least once per month. This frequency is sufficient to detect crud deposition effects.For pressure deop considerations in design analyses, it is assumed that a conservative amount of crud is deposited on the fuel rods, and the fuel rod spacers.
This is reflected in a decreased flow area, increased friction factors, and increased spacer loss ccafficients.
The effect of this crud dcrositicn is to increase i.he core pressure drop by apprcximately 1.7 psi.
'le ccnclude that the assu:.ptions regarding crud deposition used in <*.s::;n :nalyses in ccnjunction with the required flow canitoring are acceptable.
'!."P-2 will have a loose parts monitoring system (LPMS) operai.ional at the time of initial reactor startup testing.
This system provides the capability to detect, alarm and record accoustic signals generated when i
loose parts within the reactor coolant system impact other reactor coolant system components. The systcia has been installed to meet the i
operability requirements of Regulatory Guide 1.133.
The applicant olso cc : ail.s to evaluate the system to addr ess confor:nnce with the'
cylatuy Cuida 1.133, Pcvisica 1 (:'ay 1531).
The e'nfoi..mte c.aluation
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,-iasize the follen:ing arms:
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1.
A description and evaluation of diagnostic procedures used to I
confirm the presence of a loose part.
t 2.
A description of how the operators will be. trained in the purpose and implementation of the system.
/
We will review the applicants conformance.qvaluation report when it becomes available, consistent with our plans,for review of the operating 4
plants. Anz action resulting from our review will be applied at that time. On this basis, we'-fi,nd the LPMS acceptable for an operating license.
? :ial
, !.e ': tuff has revir. cd the t' ernal hydreiulic design of the core as J:scrib:.d in Section 4.4 of the EcAR for ZP-2.
The scope of the review included the design criteria, imploa:ntai.ico of the design criteria as presented by the final core design, and the steady-state analysis of the core thermal-hydraulic performance.
The epplicant's thermal-hydraulic analyses were performed using approved iaethods and correlations and found acceptable.
iiowever, the operating license should be restricted to the following conditions:
t 1.
single lcop operaticn is not peraitied voless suppori.ing caalyres era provided and appiow d; i
e.
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7-2.
operation beyond Cycle 1 is not permitted until stability analysis is provided and approved for the additional cycles of operation; 3.
the natur'al circulation operating mode is not permitted; and 4.
the core flow should be checked at least once per day and the average power range, monitor flow biased scram calibrated at least a
once per month to account for possible effects of crud deposition.
l 1
The above restrictions should be incorporated into the proposed Tech-7 nical Specifications, except for Item 2 whic.h should be incorporated as a license condition.
In addition, the following open item should be resolved prior to issuance of the operating liranse:
3
-- the operating limit.iCPR calcul;tcd by mcluding the ODY,'; cethods 2
j r.ust be pra ided,'or iaview m d n;;,rcval.
- le conclude t%t, with the exceptions noted above, Uie thcnnal hydraulic design of the core conforms to the requirements of General Dasign l
Criterion 10 of 10 CFR Part 50, Regulatory Guides 1.68 and 1.133, and Section 4.4 of the Standard Review Plan and is, therefore, acceptable.
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