ML19256A619
| ML19256A619 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 12/15/1978 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19256A606 | List: |
| References | |
| NUDOCS 7901090101 | |
| Download: ML19256A619 (10) | |
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UNITED STATES NUCLEAR REGULATORY COMMISSION c, A j
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SAFETY EVALUATICN 3Y THE CF: ICE CF NUCLEAD REACTCR QEGULATICN SUPPCRTING APENCPENT N0 26 TC FACILIT' CPE2ATING LICENSE NC. :P3-54 SACRAMENTO MUNICIPAL UTILITY DISTRICT RANCHO SECO NUCLEAR GENERATING STATION DCCKET NO. 50-312 1.0 Introduction By soolication for license amenanent dated Septemoer 13, 1978 (Reference 1) the Sacramento Munici pal Utili ty Di strict (tne licensee) recuested cnances in tne Tecnnical Scecifications appenced to Facili ty Operating License DPR-54 for the Rancno Seco Nuclear Generating Station (Rancno Seco). The prooosed cnanges relate to coerating limits for Cycle 3.
Addi tional information relating to tne requested changes were orovided in the licensee's letter of Novemoer 15,1978 (Reference 13).
The refueling of Rancno Seco for Cycle 3 operation will result in a core loading as follows: 56 new Satcn 5 Mark S 4 fuel assemolies wi th in ini tial enricnnent of 3.04 wtt U-235: 56 Satch 4 Mart 3 4 fuel assemolies wi th an initial enrichment of 3.19 wtt U-235; 60 Satcn 3 Marr. B-3 fuel assemolies wi tn an initial enricnment of 3.00 wt% U-235; and 5 Satcn 2 Mark S-3 fuel assemolies wi th an initial enricnment of 2.67 wtt U-235. The five twice burned Batch 2 assemolies will De reloaded into tne central portion of the core.
Batches 3 and 4 will be snuffled to new locations and Baten 5 will occupy the core perichery and eign interior locations.
The licensee has Dracosed changes to the follcwinc Technical Speci fi-cations (T/S):
1.
Core Protection Safety Limits (Figure 2.1-2) 2.
Core Protective Safety Bases (Figure 2.1-3) 3.
Protective System Maximum Allowaole Setooints (Figure 2.3-2) 4.
Limiting Concitions for Ooeration (LCO) on removal of recurdant eouiccent from service for maintenance (Section 3.3.3) 5.
Allewaole Quadrant Power Tilt and overall cnange in form of speci fication (Section 3.5.2.4)
[/[)()jL()bI C)(C3 (
. 6.
Rod Position Limits (Figures 3.5.2-1,
-2, -3.
4
-5, -6) 7.
Axi si Power Shacing Rod ( APSR) Dosition Limits (Figures 2.5.2-7, -8) 8.
Core Power Imoalance Enveloce (Figures 3. 5.2-9, -10, -11 )
The evaluation of Rancno Seco's crocosed modifications to tne T/S are presented in the foll: wing sections of tnis safety Evaluation Report (SER).
2.0 Evaluation 2.1 Fuel "ecnanical Desien The Saten 5 fresh fuel uses the Mart 3 4 fuel asseroly design re-viewed and accepted by us for use during Cycle 2.
Also. tnis tyce of fuel assemoly is currently being used in Oconee Nuclear Station, Units Nos. 2 and 3, and Arkansas Nuclear One, Unit No.1.
Tables 4-1 and 4-2 of Reference 2 summarize the design character-i stics of the Mark B-4 and Mart B-3 fuel types. The Batch 5 fuel assemoly design is based upon estaolished concepts and utilizes standard component materials. Therefore, on the bases of the anal-ysis presented and previous successful operations witn equivalent fuel, we conclude that the fuel mecnanical design of this fuel is acceptacle and does not decrease the safety marcin.
2.1.1 Cladding Creen Collaose Fuel rod cladding creec collaose analyses have been cerformed for the mos t limi ting (i.e., most highly exoosed) assemolies to ce in-cluded for Cycle 3.
Tne analyses were performed according to tne conservative metnods and assumotions descrioed in Reference 3.
These analyses show that tne time to rod cladding collaose will be in excess of 30,000 ef fective full power hours *(EFPH). Because no fuel assemoly will reach a total exposure as hign as 30,000 EFPH during Cycle 3 (Table 4-1 of Refere,ce 2), we conclude that cladding creep collapse will not occur duri.1g the cycle.
. 2.1.2 Cladcing Striin The fuel Cladding str3in 3nalysis das cerformed using a 9urcer Of conservative assunptions: maximum allowacle fuel cellet di ameter and density; and the lowest cermitted tolerance for tne cladding inner di ame te r.
Since the fuel cellet design based on tnis analysis is such that the cladding strain is less than it at 55,000 M'ad/'17U burnuo, and since tne max 1 mum excected tnree-cycle local pellet burnup in Cycle 3 will e less tnan 55,000 mwd /MTU the 1.0% limit on cladding plastic circumferential strain is not violated.
2.2 Fuel Thermal Design The thermai line3r neat r3te (LHR) limits nave been estaDli sned for tne Cycle 3 fuel usino tne TAFv coce (Reference 4) and issumed fuel densification to 96.5% of theoretical density. These limits sre stated in Taule 4-2 of Reference 2.
The tnermal LHR limi ts wnich ensure that fuel center melting does not occur are less re-strictive than the LOC A LHR limi ts.
Because the LOC A LHR limi ts will be met Dy operating witnin the limiting conditions for coera-tions contained in the Rancno Seco Tecnnical Speci fications, tne thermal LHR limits will also be met.
We conclude that the indicated thermal LHR limits are acceptable for preventing center melting of the Cycle 3 fuel.
2.3 Nuclear Design The design length for Cycle 3 is 310 effective full ;ower days.
Cycle 3 nuclear carameters including critical boron concentrations, control red wortns, Doooler coef ficients, moderstar coef ficients, xenon wortn and ef fective celayed neutron fractions nave been calcu-lated using the P0007 code. These are presented in T30le 5-1 of Reference 3 wnere tney are camcared to tne Cycle 2 values.
4-Shutdown margins have been calculated for Beginning of Cycle (3CC) and End of Cycle (EOC) (Tiole 5-2 of Deference 2).
The Calculated minimum shutdown margin during Cycle 3 is 2.25 ;K/K wnicn suostan-tially exceeds the required value of 10
</K.
During the last 20 days of Cycle 3, the APSR's will be totally wi tn-drawn from the core to gain additional reactivity wnicn will orovide a more ef ficient utilization of the fuel. The axial stabili ty of the core was verified using the accroved FLAPE code (Reference 5) and was found to be acceptacle wi tn all the APSR's fully wi thdrawn.
'ae conclude that the Cycle 3 nuclear design does not differ in a significant way from earlier cycles, tnat the nuclear parameters of Cycle 3 have seen calculatec ey acceptacle methods and that the nuclear design has resulted in an adeauate snutdown margin. The nuclear design for Rancno Seco Cycle 3 is, therefore, acceptaole.
2.4 Thermal Hydraulic Desicn The thermal-hydraulic design conditions for Rancho Seco Cycle 3 are included in Taole 6-1 of Reference 2.
Only the reference desian radial-loc 3l power ceaking factor and anticipated minimum decarture from nucleate boiling ratio di ffer from the Cycle 3 values. The first of these di f ferences is di scussed bel ow, and the second i s reasonaole and acceptacle in that it represents an increased margin to the safety limit Decarture fran Nucleate Boiline Ratio (DNBR).
2.4.1 Removal of Orifice Rod Assemolies The most significant di fference between the thermal hydraulic design for Cycle 3 and that for Cycle 2 is tne removal of the 52 orifice rod assemolies (ORAs). This will leave a. total of 106 fuel assem-olies out of a possible 108 wi thout orifice reds or other byoass ficw limi ting devices. Two fuel assemolies will contain neutron sources. The analysis was conservatively based on 108 vacant fuel asserolies and will result in an increase in eypass flew from 3.34%
for Cycle 2 to 10.4% for Cycle 3.
The increased byoass flow also causes a decreased ficw to fuel assemolies, and the licensee has re-evaluated the effect of this modification on the reactor core DNSR sa fe ty li mi t.
The re-evaluation indicated that a decrease in :ne reference design racial-local peaking factor (F g ) fran 1.78 to 1.71 3
compensates for the larger bypass ficw so :nat no cnange in the 2NER safety limit will be necessary. The DNBR safety limit was derived using the BAW-2 critical neat flux correlation (Reference 5).
Con-sicerir.g tne fact tnat tne maximum theoretical value of Fad is 1.54,
5 we 9 ave concluded tnat the orcoosed reduction in F,y to 1.71 i s 3ceouate to of fset tne increased cyoass ficw and still or0 vices an lit margin to the cesign value of 1.71, wnich i s mo re than acequate to cover calculational uncertainties.
2.4.2 Effect of Rod Bow on Thermal Desian The ef fect of fuel rod bow has been reviewed generically in Reference 7.
Based on the red Dow model approved by tne staff the licenses has soplied a rod bow DNBR penalty of 11.2% to all analyses that define plant operating limits and to design transients (Reference 2).
The 11.2* 7enalty wnicn has oeen applied includes 3 1% contribution,
associ ate. wi tn ci tch reduction due to f abrication tolerances and i ni tial red bow, and a 10.2% contribution from burr.uo decencent Dowing. The 11.2% penalty is valid for a maximum ournuo of 33,C00 mwd /MTU. Considering the actual Cycle 3 Durnuo, tne calculated red bow cenal ty i s 8%.
This is based, as for Cycle 2, on the maximum burnuo of the batch tnat contains the fuel assemoly wi th the maximum radi al-local ceaki ng factor. Several assenclies have burnuos greater that 33,000 mwd /MTU. The maximum Cycle 3 burnuo is stated by the licensee to be 37,680 mwd /MTU it, a Satch 3 fuel assenoly. The maximum radial-local peaking factor for a Batch 3 assemoly is 27% lower than the Batch 5 assemoly wnich deternined the calculated 8% red bow cenalty for this cycle. This large difference in ceaXing more tnan compensates for tne di f ference in rod bow cenalties.
Based on :ne use of an approved model and the results cf a boundina analysis, we conclude that the licensee has adequately taken iuel rod sowing into account for the thermal design of Rancho Seco Cycl e 3.
3.0 Evaluatien of Accidents and Transients As di scussed in Section 2.3 of this Report, tne nuclear parameters,
~
wnich comorise a portion of the inout to the accident and transient analyses, have been evaluated using acceptacle methods. Further, we concluce Dased on Tables 6-1 and 7-1 of Reference 2, tnat the Cycle 3 nuclear parameters are bounded by values assumed for accident analyses in the FSAR (Reference 8) and tne Rancho Seco Densification Report (Reference 9).
. J The apolicable LOCA analyses for Rancno Seco have been cresented in Deference 10 whicn nas been accepted oy ne staf f for generic accli-cation to B&W plants of the Rancno Seco class (177 T A Lowered Loco 31 ants). The fuel censification recort (Geference 9) cescribes tne ef fect of censification on LCCA analyses anc tae u3e of the TAFY coce (Reference 4) to calculate fuel rod internal pressure and cellet volu-metric average temperature. The latter parameters, wnicn are part of the LOC A' i nput, are also af fected by enhanced fission gas release, but the original TAFY calculations did not incluce tne ef fect. Cal cul a tions using the B&W code, TACO, (Reference ll) have shown that the internal pressures and average temperatures calculated using TAFY adequately ocund the ef fects of ennanced fission gas release for up to 42,000 mwd /MTU fuel rod burnup, a nigner turnuo than will te attained during Cycle 3 operation.
We concur that the refueling of Rancno Seco for Cycle 3 will not result in kinetics parameters outsice the bounds assumed for tne :SAR analysis, and that no change in the DNBR safety limit is recuired. Tu r *.ne rmo re,
the ef fects of fuel rod bowing, fuel densification, and ennanced fission gas release on safety limits and on all transients and acci-cents, including LOCA, have adeauately been taken into account.
Despi te the foregoing, in the light o, our Order for Modification of Licerse dated July 21,1978 (for details see the enclosed Exemotion),
we cannot conclude that oceraton of Rancho Seco in Cycle 3 as cresently configured would be wnolly in conformance wi th the requirements of 10 CFR 50.46 relative to the performance of the E.mercency Core Cooling System (ECCS). This is because until modi fications orcoosed oy the licensee and approved by the staff have been imolemented at tne facili ty, the licensee must rely upon granot operator action to assure acceptacle mitigation of the consequcnces of a small break loss of coolant accident
( LOC A). To adcress this concern in the interim, the licensee has de-fined certain operator actions to be canaleted wi thin a s:eci fied time frame and has orovided an acceptacle analysis demonstrating that if tnese actions are taken upon occurrence of a small creak LOCA, Onere is a very substantial safety margin relative to the acceptance criteria for sucn events. The licensee has also trained operating perso. nel to execute the required procedures and verfied that they are cacaDie c' canoletion wi thin the requi red time frame. In addi tion, the Canmission's s'fice of Inspection
, Ind EnforCenent nas ver1 fied 09a* the Orocedures "3ve 3een i*cl ?"e90ec it 3ancno Seco and tnat tr3ininq in tre or:cedures nas teen 00ccucted.
3ased on these consider 3tions, ina consiceritior af tre coolic interest.
.e tre grinting oursuant to lJ CFU 50.12, concurrent 41 On issuarce if inis imencne1t, an Exenotion from tae crevi sions Jf 11 ':03 50.2c sucn is to authorize coerition of 3ancho Seco in Cycle 1.
- c 1rintiac nis Exenotion, we tre 3dcina certain license conai tions ret i-'--
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vince of croce u ?s 'or oceritor act':n arc :
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,n itn eliminate reli ance in Or r:
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Based on the foregoing, and the fact tnat tre dose calculaticas of the FSAR assumed maximum ceakings and burnups wnicn Ocund One values present in Cycle 3, we further conclude that the consequences of transients or accidents during Cycle 3 will be no greater than pre-viously evaluated. There will be no increase in the probability of occurrence of any accident or transient, and no new type of accident or transient will be introduced as a result of the refueling.
We, therefore, conclude that operation of Rancho Seco in Cycle 3 within the limits set forth in the facility license, as revised ty ne accomoanying amendment and xemetion, is accectacle.
4.0 Physics Startuo Test Procran The physics startup test program for larcho Seco Cycle 3 as crocosed in Section 9 of 3AW-1499 'was reviewed. The anysics startuo test program consists of low power neasurement of critical boron concen-tretion, temperature reactivi ty coef ficients, control red croup reactivi ty worth and ejected control rod reactivi ty worth. Dower escalation tests consist of cower di stribution veri fication at acoroximately 40%, 75t and 100% power and temoerature reactivi ty coef ficient and power Doppler reactivi ty measurements at sooroximately 100t cower. For each test tne acceotance criteria and tne actions to be taken if the acceptance criteria are not net were sucolied.
Addi tional informatian about the ejected rod reactivi ty worth test, the control rca grouc reactivi ty worth reasurements and One recorting of resalts was requested. This infornation was suoplied in Reference 13. 01 scussionswi th the licensee on Novemoer 23 and Movencer 30, 1978, further clari fied One actions to be taken if tne measured reactivi ty worth of rod grouos 5-7 is more tnan 10% less than the credicted worth.
The licensee has agreed that addi tional rod wortn neasurenents will be cerformed i f tne measured worth of groups 5-7 is less than the predicted worth minus 10%. The acor0eri ate uncertainties will e acclied to these measurements before tne measured values 3re used to veri fy adecuate shutdown margi q.
The pnysics startuo test cronran includ ng tsst condi tions, i
icceptance cri teri l and actions to ce taken if the ic:40tinc?
- ' teri i tre *ot *e.1as oeen reviewea ina f un ic:er lo's.
. 5.0 Evaluation of Tecnnical Soecification Cnances The arcoosed cnances to the are li sted in Section 1.0 of nis 2
SED.
ine licensee nas states, in Re'erence 2, tnat tne :/S limi ts cased on Ceoarture from Nucleate 3 oiling Ratio (DNBR) and Linear Heat late (LHR) include aDercpriate allowances for projected rod tow penalties and a statistical comoination of tne nuclear uncertainty factor, engineering hot cnannel factor, and rod oow ceaking cenalty was used 4 1 evaluating LHR criteria, as accroved in Reference 12.
Tne develecmeat of the revi sed safety limi ts, trip settings, core imoalance limits, and control rod and aPSR insertion limits aas done usinq the approved FLAPE Code (Reference 5) wi th the use of Cycle 3 oeaking factors including a ceakinc f actor of 1.0736 to account for tne I/S change to 4.92% maximum steady state cuadrant power til t.
The licensee has verified tnat the consecuences of all accidents are acceptable considering a s* aady state quadrant til limit of 4.92% and that this condi tion caunds one otner allewable cuadrant tilts at lower power levels as permitted by tne I/S.
The change to the Core protection Safety Bases (Figure 2.1-3) is merely a cnange in the units used to ex:ress primary coolant flow rate and has no safety significance.
The change in the form of the technical s:ecificacion governing quadrant tilt (Section 3.5.2.t, together with the associated changes in the basis for One specification and the revised defini-tion of quadrant tilt (Section 1.6) clarifies the present potentially amoiguous specification and brings it into closer agreement with the forn of scecification governing quadrant til: currently issued for Babcock & Wilcox reactors. We nave reviewed both the cresent and the proposed forms of the technical scecification and note that both provide accroximately the same pcwer penalty for the same degree of excess quadrant tilt, ano nave a:ner provisions which are basically similar. Accordingly, we conclude that the change in the form of the specification does not have a significant adverse effect on the safety of operations and may, through its greater clarity, enhance safety. We therefo re find this change in form acceptable.
It is noted that the staff suggested the change in the definition of quadrant tilt and minor wording changes in the s:ecification as procosed by the licensee.
These were discussed with and agreed to by the licensee.
.g-Based on the foregoing considerations, ae 'ind the changes to T/S Sections 2.1, 2.3 and 3.5.2 to be acceptacle.
With regard to Specification 3.3.3, this s;ecification presently requires verification of :ne ocerability of one of a cair of engineereo safety feature components before maintenance can be performed on the other component. The licensee has requested that this special test be waived provided tne comscnent wnicn would otherwise be tested is within the interval covered by its most recent surveillance.
In support of this position, tne licensee cites the Standard Technical Specification for S&W reactors, which does not include this provision.
We nave advised the licensee that the Standard Technical Specifications are based on a number of interrelated provisions and that if he wisnes to adopt this particular Drovision, he should propose to accot the related provisions as well. He has not procnsed to do so, but has stated that if we co not concur 4ith his judgment that tne procesec change is in the interest of safety, ne will witharaw his recuest for this change. While ne recognize some terit in the Oroposed change, we do not concur tnat taken alone this change would improve safety. Accordingly, we shall consider this request to have been withdrawn by the licensee.
6.0 Environmental Considerations We have determined that the amencment does not authori:e a change in ef fluent types or total amounts nor an increase in power level and will not result in any significant environmental impact. H avi ng mace tnis deternination, *e have furtner concluded trat the amendment involves an action wnich is insignificant fran tne ;;ancooint of envirarmental impact and pursuant to 10 CFR Sl.5(d)(4), tnat an environmental impact statement or negative declaration and environ-mental imoact aopraisal need not be prepared in connection ai tn the issuance of tnis amendment.
7.0 Conclusion We have concluded, based on the considerations di scussed above, nat:
(1) because the amendment does not involve a significant increase in the probability or consequences of accidents previously considered and does not involve a significant cecrease in a safety margin, the amendment does not involve a significant hazards consideration, (2) there is reasonaole assurance nat the nealtn and safe y of tne public will not be endangered by coeration in the crocosed manner, and (3) except where Exemption is expressly granted, sucn activities will be conducted in concliance with the Commission's regulations and the issuance of tnis amendment will not be inimicai Oc :ne common defense and security or to the health and safety of tne puolic.
Cated: Decemoer 15, 1978
- 2eferences 1.
Letter from SMUD (J. J. Mattimoe) to NRC (R. w. Reid) da ted Sec tecoe r 13, 1973.
2.
Rancho Seco Nuclear Generating Station, Unit 1 Cycle 3 Reload Recort, 3 AW-1499 dated September 1978.
(Enclosure to Reference 1, above) 3.
Program to Oetermine In-Reactor Performance of 33W Fuels -
Cladding Creep Collapse BAW-10084 Rev. I dated Novemoer 1976.
4 TAFY-Fuel Pin Temperature and Gas Pressure Analysis, BAW-10044 da ted May 1972.
5.
FLAME - Three Dimensional Nadal Code for Calculating Reactivi ty 3nd Power Di stribution, 3AW-10124A ca ted August 19 76.
6.
Correlation of Critical Heat Flux in a Sundle Cooled by Pressuri:ed Water, 3AW-10000A dated May 1976.
7.
Memo to D. 3. Vassallo (NRC) from D. F. Ross (NRC), Revised Interim Safety Evaluation Recort on the Effects of Fuel Rod Bowing on Thermal Margin Calculations for Light Water Reactors, Feo rua ry 16, 1977.
8.
Rancho Seco Nuclear Station, Unit 1 - Final Safety Analysis Report, Occket No. 50-312.
9.
Rantno Seco Unit 1 - Fuel Densification Recort, 3AW-1393, Babcock &
Wilcox, Lyncnourg, Vi rgi nia, June 1973.
10.
R. C. Jones,
. R. Siller, and 3. M. Dunn, ECCS Analysis of 34W's 177-FA Lowered Loco NSS, 3AW-10103A, Rev. 3, 3accock & Wilcox, Lynchourg, VA.
- 11. TACO - Fuel Pin Performance \\nalysis, BAW-10087.
12.
S. A. Varga (USNRC) to J. H. Taylor (38W), Letter, " Comments on B&W's Subni ttal on Combination of Deaki ng Factors," May 13, 1977.
- 13. Letter from SMUD (J. J. Mattimoe) to NRC (R. W. Reid) dated Novemoer 15, 1978.