ML19282C214
| ML19282C214 | |
| Person / Time | |
|---|---|
| Site: | Quad Cities  |
| Issue date: | 02/23/1979 |
| From: | Ippolito T Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19282C212 | List: |
| References | |
| NUDOCS 7903220230 | |
| Download: ML19282C214 (11) | |
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UNITED STATES r
4-NUCLEAR REGULATORY COMMissiCN 5
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WASHINGTON, D. C. 20555
'5 SAFETY EVALUAT!ON SY THE OFFICE OF NUCLEAR REACTCR REGULATIO1 SUPPORTING AMENDMENT NO. 50 TO FACILITY LICENSE NO. DPR-29 COMMONWEALTH EDISON COMPANY OVAD CITIES NUCLEAR POWER STATION UNIT NO.1 DOCKET NO. 50-254 1.0 Introduction By letter dated Novemoer 20,1978 (Reference 1) and supplemented by letters dated December 15,1978 (Reference 2), and February 14, 1979 (Reference 3) Commonwealth Edison Company (the licensee) requested amendment to the Technical Specifications appenced to Operating License DPR-29 for Quad-Cities Nuclear Power Station, Unit 1 (QC-1). The oroposed changes relate to the fouhh refueling of OC-1, wnich involve tne replacement of 192 exposed fuel assemblies with a like number of fresh, two water rod, retrofit 8x8 (8x8R) fuel assemblies. Four of the 192 fresh fuel assemblies will be barrier lead test assemolies which are designed to investigate potential fixes for pellet-cladding interaction fuel failure mechanisms (Section 4.0).
In support of tnis reload appli-cation, the licensee has submitted a supplemental reload licensing docu-ment (Reference 4) prepared by GE, and proposed Technical Specification changes in Reference 1.
This reload is the first for QC-1 to incorporate General Electric's (GE) 8x8R fuel design on a batch basis. The description of the nuclear and mechanical design of the 8x8R fuel and the exposed fuel designs is con-tained in GE's generic licensing topical report for SWR reloads (Refer-ence 5).
Reference 5 also contains a. complete set of references to GE's topical reports which describe GE's SWR reload analysis methocs for the nuclear, mechanical, thermal-hydraulic, transient and accident calcula-tions, together witn information on the. applicability of these methocs to cores containing a mixture of different fuel designs. Portions of tne plant-specific data, such as operating concitions and design para-meters which are used in transient and accicent calcul,ations, have also been incluced in the topical report.
Cu safety evaluaticn (Reference 5) of GE's generic reload licensing
- ical recor: ccnclucec tna :ne nuclear anc mecnanical cesign of :ne 3x5R fuel and SE's analytical oe:nocs for nuclear, ne-nal-nycraulic,
- ansient and accicen calcula:icns, as accliec :: c:res containing micures' cf 7x7, 5x5, ann 5x5R fuel, are acceo sole. Our acceptance cf :ne nuclear and mecnanical cesign of :ne stancarc 5x3 fuel was ex-ressed in :ne staff's evaluation (Reference 7) of :ne infor ation in
- eference 8.
79 03220 2,3 o c
As part of our evaluation (Reference 6) of Reference 5 we fcund the cycle-independent input data for the reload transient and acci-cent analyses to be acceptable. The supplementary cycle-depencent information and input cata are previoed in Reference 4, whien follows tne format and content of Appencix A of Reference 5.
As a result of the staff's generic evaluaticn (Reference 6) of a substantial numoer of safety c:nsiderations related to use of 8x8R fuel in mixed core loadings with 8x8 and 7x7 fuel, only a limited numcer of additional review items are incluced in this evaluation.
These include the plant and cycle-specific input cata anc results presented in Reference 4, the LOCA-ECCS analysis results for the reload fuel cesign, and those items icentified in our safety evalua-tion as requiring special attention during reload reviews.
2.0 Evaluation 2.1 Nuclear Characteristics For Cycle 5,192 fresh 8x8R fuel bundles, with a bundle average en-richment of 2.65 wt/% U-235 will be loaded into the core. These will replace a like number of exposed fuel assemblies. The remainder of the 724 fuel assembly reload core will consist of the irradiated 7x7 and 8x8 fuel assemolies exposed curing previ::us cycles.
The information provided in Section 6 of Reference 4 incicates that the fuel temperature and void cependent reactivity response of the reconstituted core is not significantly different from that of pre-vicus cycles.
Additionally, scram effectiveness, Figures 2a and 20 -
of Reference 4, is also similar to earlier cycles. The 2.0% ak/k calculated shutdown margin for the reconstituted core meets the Tech-nical Specification core subcriticality requirerent in the most re-active operating state with the single most reactive control rod fully withdrawn and all other rods fully ~' inserted. Finally', Reference 4 indicates tnat a boron concentration of 600 ppm in tne mocerator bas been calculated to make tne reactor subcritical oy at least 4'.5% ak at 20*C, and xenon free concitions. Therefore, tne alternate shutccwn requirement of the General Design Criteria can be achievec ey the Stancby Liquic Control System. We nave reviewed tnese analyses and on :ne bases as stated aoove find tne results to :e acceptacie.
I 2.2 Inemal-Hveraulics 2.2.1 Fuel Claddina Integrity Safety Limit MCPR As stated in Reference 6. for SWR cores unicn relcac with GE's retrofit Sx5R fuel, the allo.<aole minimum critical power ratio (MCPR), resulting f rcm either core-wide or locali:ed abnormal cperational transients, is equal to 1.07. With this MCPR safety limit, at least 99.9% of the fuel rocs in the core are expected to avoid boiling transition curing these transients.
The 1.07 safety limit minimum critical pcwer ratio (SLMCPR) proposed b'y the licensee represents a.01 increase from the previcus 1.06 SLMCPR.
The basis for the revised safety limit is adcressed in Reference 5.
This enange continues to meet the reccmmendations of Standarc Review Plan 4.4 and on that basis has been found acceptacle in Reference 6.
Modifications to the Tect.nical Specification have been incorporatec per this finding.
2.2.2 Coerating Limit MCPR Various transient events will reduce the MCPR from its normal operating value. To assure that the fuel cladding integrity safety limit MCPR will not be violated during any abnomal cperational transient, the most limiting transients have been reanalyzed by the licensee to deter-mine which event results in the largest reduction in critical pcwer rati o.
Each of the events has been conservatively analyzec for fuel types and for the full range of exposure through the cycle.
The calculational methods, which include cycle-incepencent initial conditions and transient input parameters, are cescrioed in Reference 5.
Our acceptance of the values used and relatec transient analysis methocs appear in Reference 6.
Supplemental cycle-cecendent initial conditions and transient input parameters used in the analysis aopear in the table in Section 6 and 7 of Reference 4.
Our evaluation of the methocs usec to cevelop these su;plementary transient input values have alreacy oeen accressed and appear in Reference 6.
The overall transient metnodology, including cycle-incependent transient analysis inputs, provices an ace-cuately conservative basis for the detemination of transient J.CPRs.
Tne transient events analy:ed were loac rejec. ion witncut cy; ass, tur:ine ri: wi ncut ryoass, feecwater :0ntrolier failure, less of 100*: feec-water nea-in; anc c:n.r:1 roc witncrawai error.
r
4 Based on cur review of the licensee's suomittals, tne most limiting acnormal operational transient for all fuel types and exposure in-tervals is the load rejection without typass. Therefore, tne licensee wfil be recuired to meet the following operating limit MCPRs:
Fuel Tyce Oceratino Limit MCPR 7x7 1.23 8x8 1.29 8x8R 1.29 Thus, when tne reactor is operated in accordance with the above operating limit MCPRs the 1.07 SLMCPR will not be violated in the event of the most severe abnormal operational transient. This is acceptable to the staff per the finding of the previous section.
On this basis, operating limit MCPR Technical Specifications nave been established.
In the, analysis of the rod withdrawal error (RWE), flow biased upscale rod block monitor (RBM) setpoints are established to assure that the safety limit MCPR is satisfied. Therefore, this setpoint is specified in the Technical Specifications. On the basis of the acceptance of RWE analysis methods in Reference 6 we find the calculatec CPR and RBM setpoint for the RWE acceptable.
'2.3 Accident Analysis 2.3.1 ECCS Accendix K Analysis The licensee has reevaluated the ECCS performance in Enclosure IY to Reference 1.
This reevaluation provides the bases for relaxation of Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) limi ts. The relaxation is primarily ' cue to the effects of crillea lover tie plates in the 8x8R reload fuel. This reevaluation is casea on NRC acces ed GE ECCS models and input (Reference 9).
The lead plan-for this reevaluation is Duane Arnold Energy Center (Reference 10). The justification for the use of tnis lead plant.
analysis is the same as for Pilgrim (References 11 anc 12) wnicn '
was found accectacle in Reference 13. Upon cur request, tne licensee nas proviced further cocumentation on the use of Duane Arnold as a leaa pian.
In our review we find that OC-1 nas similar ECCS injec-ti:n logic as Juane Arnold.
We nave also creviously conciuced tnat tne use of ne :uane Arnold leac clan analysis f:r pian s of similar t
. power censity (Reference 13) ano size (Reference 14) is acceptaDie.
This conclusion is based en the fact that a BWR 4 witn plugged one inch diameter holes in the core support plate and with drilled lower tie plates'(Duane Arnolc) is hycraulicly similar to a BWR 3, which never had tne one inch core supcort plate nole, witn crilled lower tie plates (Quad-cities 1).
Therefore, the use of tne Duane Arnold analysis as a lead plant is acceptable.
In the NRC required confirmatory Dreak spectrum analysis, a longer time period for hot node uncovery was predicted for a 34% of cesign basis accicent (DSA) size break tnan for the DBA. However, the 34% of DBA size break coes not result in higher peak claccing temperature than the DBA because boiling transition and core uncovery for the smaller break occur at a later time af ter event initiation than for the DBA.
Thus, lower decay heat generation is present at tne critical time in
- ne calculation. This is the same explanation as the previous QC-1 ECCS performance analysis (Reference 15) which has previously been accepted (Reference 16). The licensee has documentec that confirma-tory analysis has shown that the DBA is limiting (Reference 3). There-fore, based on the referenced lead plant applicability and the creak spectrum verification that the limiting break is the DBA, we find that all requirements of 10 CFR 50.46 and Appencix K to 10 CFR 50.46 will ce met wnen the reactor is operated in accorcance with MAPLHGR versus aver-age planar exposure values of Taoles 4A thrpugh 4H in Enclosure lY to Reference 1 wnich have been incorporated in the revised Tecnnical Speci fications.
2.3.2 Control Rod Droo Accident The analysis of the control rod crop accident (CRDA) has oeen performed on a generic (bounding analysis) basis,. In our safety evaluation (Reference 6) of GE's generic reload methocs (Reference 5) we concluded tnat tne bounding. analysis basis is acceptable with the provision that tne key input parameter for a plant specific reload are conservatively counc oy ne analysis assumptions.
In tne plant scecific reload appli-cation (Reference 4) the licensee has snewn that tne maximum incremental centrol rod worth is conservatively represented in tne councing analysis.
This is acceotao.le en tne previcusly centienec basis.
~
. 2.2.3 ruel Loacine Error Tne licensee nas also considered the effect cf a possicle fuel loacing error en bundle CPR. An analysis of tne most severe misoriented fuel loacing error using GE's new methocology (References 17 and 18), which as codifie'd, nas been acproved (Reference 19) cy the staff, shows tnat
- ne worst possible rotation of a fuel bundle will no: cause a violation of the 1.07 safety limit MCPR. Additionally, an analysis of the most severe misiccated fuel bundle with GE's new, approved methocology shows that the worst potential mislocation will not viciate the MCPR safety l imi t.
We find tne results of these analysis acceptacle.
2.4 Overcressure Analysis The overpressure analysis for the MSIV closure with high flux scram, wnich is the limiting overpressure event, has been perfomed in accordance with the requirements of Reference 6.
As specified in P.eference 6, the sensitivity of peak vessel pressure to failure of one safety valve has also Deen evaluated. We agree that there is sufficient margin between tne peak calculatec vessel pressure and the cesign limit pressure to allow for the failure of at least one valve.
In the analysis the licensee has assumed relief valve setpoints with conservative bias wnich accounts for measurement uncertainty as specified in the revised Technical Specifications.
(The only change from previcus analyses is a reduction of 10 psig in ene safety / relief valve setpoint.) Therefore, the limiting c-verpressure event as ana-lyzed by the licensee is considered acceptacle on the bases outlinec in Reference 6.
2.5 Thermal-Hydraulic Stability A thermal-hydraulic stability analysis was performed with the methods cescribed in Reference 5.
The results show tna :ne cnannel hycro-cynamic and reactor core cecay ratios.at the least stable operating state (correspondi~ng to the intersection of :ne natural circulation curve and 105% roc line on tne power-flow mac) are celow the 1.0 J1 imate Performance Limit cecay' ratio pr:posec Oy GE.
The staff has expressed generic concerns regarcing reactor core ~
.nermal-hycraulic stacility at the least stasie reac:cr concition.
~ is c:ncition ::ulc :e reached curing an c erational transient n
fr:m hign pcwer if ne plant were to sustain a trip of coth re:ircu-ati:n cumps wi ncu a reactor trio. Tne ::n: erns are motiva:ec :y increasing ce:ay ra-ics as ecuilicrium fuel cycles are a:Orcacne:
anc as reicac fuei :esigns :nange. The staff c:ncerns relate to
- ctn :ne consecuencas cf c:erating at a :ecay ra ic cf 1.0 ano ne :1:acility f :ne analytical me:nces :: a::gra eiy predi::
ce:ay ratics.
7-The General Electric Ccmpany is accressing these staff concerns througn meetings, tcpical reports anc a stacility test program.
Althougn a final test report has not as yet been received cy tne staff for review, it is expected that the test results will aid considerably in resolving the staff concerns.
For the previcus operating cycle, the staff, as an interim measure, added a requirement' to the Technical Specifications wnich restricted planned operation in the natural circulation moce. Ccntinuation of this restriction will also provide a significant increase in tne reactor core stability operating margins for tne current cycle so tnat the decay ratio is <l.0 in all cperating modes. On the basis of the foregoing, the staff considers the plant thermal-hyoraulic stability characteristics to be acceptable.
3.0 Physics Startuo Testino The licensee will perform a series of pnysics startup tests anc procecutes to provide assurance that One conditions assumed for the transient and accident analysis calculations will be met. The tests will eneck that the core is loaded as intenced, that the incore monitoring system is functioning as expected, and that tne process computer has been reprogrammed to properly reflect changes associated with the reload.
The test program is consistent with that previously found acceptable.
4.0 Barrier Lead Test Assemblies Four of these 192 fresh fuel assemblies are barrier lead test assem-blies (BLTA) which are designed to investigate potential fixes for pellet-cladding interaction (PCI) fuel rod f ailure mecnanism. Detailec cescriptions and analyses of the BLTAs are given in Enclosure III to Reference 1.
The SLTAs have the same 8x8 lattice configuration as the 8xSR fuel assembli.es. They differ.in that the SLTA's fuel rocs consist of two segmented rocs, their inside fuel claccing surface is linec with :irconium or cooper, and tneir fuel rocs are prepressurizec to nree atmospheres. The BLTA have been evaluated witn specific at entien to these differences and the evaluation results snew that all cesiin recuirements are satisfiec.
Safety analyses incicate tnat tne 3LTAs will have an insignificant effect on core cnaracteristics. On :nese bases,,e finc tne use of tne SLTAs to ce a:ceotacie.
f
. 5.3 ECC*pewer Ceasceown In Reference 1, the licensee has proposed EOC pcwer coastcown opera-tion which is justified on the basis of our evaluation (Reference 6) of GE's reload topical (Reference 5). In our evaluation, we did not specifically consider ECC power coastcewn operatien. We, there-fore, do not consider the subject to have completed a generic review and cannot find operation in this mode acceptable on the referenced ~
basis.
In response to our request for additional information (Reference 3),
tne licensee has referenced previous coastdown made analysis (Refer-ences 20 and 21) and has presented an argument of the acceptacility of coastdown operations. The referenced analyses are for specific reactor cycles and are, therefore, not directly applicable to this core. The analyses show that the safety margins increase for CpR and overpressuri:aticn. These increased safety margins are due to the dominant effect of decreasing total power level curing c' oast-The analyses assume a linear power decrease with exposure.
down.
This assumption is conservative because actual reactor power will decrease exponentially.
In the referenced analyses, the void co-efficient becomes less negative during coastdown operation and the scram reactivity becomes less effective as a shut cwn mechanism.
The impact on aCPR is a decrease for the font.er and an increase for the later change. The referenced analyses show that the overall effect is, as p eviously stated, increased pressure and ther=al safety margin (CPR).
As previcusly stated, the referenced analyses are not specifically applicable to this plant and cycle. However, we do agree with the licensees argument that the overall trend will be the same.
This agreement is restricted to a terminal power level of about JO%. We are confident that at 70% power the scram reactivity in-sertion will not be degraded sufficiently to result in a transient more severe than that at EOC.
For lower power coastdown operations we have requested cycle specific transient analyses or appropriate justification.
Currently, the licensee has indicated that they plan to submit the requested analyses or information and our review of operation at powers lower than 70% of rated is pen, ding on this submittal. On the above bases, we find the coastdown operation as restricted in the license condition to be acce table.
9 5.0 Linear Heat Genera-ion Rate, 8x8R Fuel Linear heat generation rate (LH3R) Technical Specifications enanges for 8x8R fuel have been mace in accorcance with the preposals of Reference 5.
The design LHGR anc maximum power spiking :enalty have Deen previcusly reviewed and found acceptacle in Reference 5.
On tnis basis we find the changes acceptaole.
In order to assure compliance with LHGR design limits for the roc withdrawal error, Limiting Total Peaking Factors (LTPF) are estao-lished for use in the APRM scram trip and rod block trip setooints.
An LTPF of 3.0 has been calculated using the methods outlinec in Reference 5.
We have considered nis method in cur generic Refer-ence 6 review and, thereby, have found it acceptable. On this basis, the specification for this LTPF on 8x8R fuel assedolies is also acceptable.
7.0 Conclusions Based on our evaluation of the reload application and available information, we conclude that it is acceptable for the licensee to proceed with Cycle 4 operation of Quad Cities Unit No.1 in the manner proposed.
We have reviewed the proposed changes to tne Technical Sn'ecifications and find them acceptable.
We have determined that the amendnent does not authorize a chance in effluent types or total amounts nor an increase in powe-level and will not result in any significant environmental inpact. Having nade this determination, we have further concluded that the 69endment involves an action which is insignificant from the standooint of environnental impact and pursuant to 10 CFR 151.5(d)(4) that an environnental imoact statement or negative declaration and environnental incact aporaisal need not be prepared in connection with the issuance of nis amend 9 nt.
We have con:1uded, based on the considera-ions discussed above, tha't-(1) Decause the anendment does not involve a si:nificant inc sase i-ne pr0banility or consecuences of accidents orevi0usly :onsice ad a--
coes n:t inv'tive a sicnificant cecrease in a sa'etv 9ar:'9, t e :Na :n does not inv:lve a sicnificant ha:ards nsideratic, f:1 tae e 's reasonanle assurarce that the hea'.t and saf ety of tne :;,;1i:
~
-n i n:.
De endanger =d by 00eration in e 'rocosed manner, an< '2) su:n 1:." :-
ties will be concucted in connliance viitr the ~:--issi: 's renul tti---
and -he i'ssuance of tris amenc, eat will not ce ini-ical :
tae ::- : -
defanse a-c secu-itv ~ cr :: the n=aith a : safe y of : e -; 'i:.
Dated: February 23, 1979
. Refere.ces:
1.
Cc==0nwealtn Edison (CE) letter to Director of Nuclear Reacter Regulation, USNRC, dated Novemoer 20, 1978.
2.
CE letter (Turcak) to Director of Nuclear Reactor Regulation, USNRC, catec Decem:er 15, 1978.
3.
Cd letter (Turbak) to Director of Nuclear Reactor Regulation, USNRC, dated February 14, 1979.
4.
" Supplemental Reload Licensing Suomittal for Quad-Cities Nuclear Power Station Unit 1 Reload 4," NEDO-24145, Septemoer 1978.
5.
" Generic Reload Fuel Application, General Electric Report,"
NEDE-24011-P-3, cated March 1978.
6.
USNRC letter (Eisenhut) to General Electric'(Gridley) cated May 12, 1978, transmitting " Safety Evaluation for the General Electric Tepical Report, ' Generic Reload Fuel Application,' (NE0E-24011-P)."
7.
" Status Report on the Licensing Topical Report, General Electric Boiling Water Reactor Generic Reload Application for 8x3 Fuel,"
NEDO-20360, Revision 1 and Supplement i by the Division of Technical Review, Office of Nuclear Reactor Regulation, USNRC, April 1975.
8.
" General Electric Boiling Water Reactor Generic Reload Application for 8x3 Fuel," NEDO-20360 Revision 1, Supplement 4, April 1,1976.
9.
" Safety Evaluation for General Electric ECCS Evaluation Mocel Modifications," letter from USNRC (Goller) to GE (Sherwood),
dated April 12, 1977.
- 11. " Pilgrim Nuclear Pcwer Station Unit 1 Emergency Cere Ccoling Sy. stems Reevaluation," August 17, 1917.
- 12. Acditional Information on Reload 3 for Pilgrim Unit 1,- August 1,1977.
- 13. NRC letter (Cavis) to 3 cst:n Edison Ccc:any ( An:0gnini), ca ac
- ct::er
.7, '977.
'a.
NRC le ter (.: polit:) to TVA (Mugnes), catec Ncvem:er '3, '973.
f
. 15.
" Loss-Of-Coolant Ac:icent Analysis Report for Drescen Units 2 anc 3 and Quad-Cities Units 1 and 2 Nuclear Power Stations (Lead Plant),"
NED0-24046 Class I, August 1977.
- 16. NRC Memorancum frca Saer to Golier, " Evaluation cf Crescen Unit 2 Reload for Cycle 6 Operation (TACS 57084)," Decemoer 2,1978.
- 17. GE letter (Engle) to NRC (Eisenhut), " Fuel Assemoly Loacing Error" cated June 1,1977.
- 13. GE letter (Engle) to NRC (Eisenhut) dated Novemoer 30, 1977.
~
- 19. NRC letter (Eisenhut) to GE (Engle) dated May 8,1978.
20.
R. L. Bolger (CECO) letter to B. C. Ruscne (NRC), " Quad-C'ities Station Unit 2 Proposed Amencment to Facility License No. DPR-30, Docket No. 50-265," cated June 11, 1976.
o 21.
R. L. Bolger (CECO) letter to E. G. Case (NRC), "Dresden Station Unit 2 Preposed Amencment to Facility Operating License No. DPR-19 to Permit Power Coastdcwn from 70% Power to 40% P0wer, NRC Docket No. 50-237," cated June 6,1977.
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