ML19317G792
| ML19317G792 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 10/07/1977 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19317G757 | List: |
| References | |
| NUDOCS 8004010621 | |
| Download: ML19317G792 (6) | |
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NUCLEAR HEGULATORY COMMISSION
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1.0 Introduction By applicatinns for license amendment dated June 2a,1977, and August 5,1977, the Sacremento itunicipal Utility District (the licensee) requested changes in the Technical Specifications to Facility Operating License DPR-54 for the Rancho Seco Nuclear Generating Station (Rancho Seco).
The proposed changes relate to the Cycle 2 core reload and the required qualifications of the Chemical-Radiation Supervisor. The qualifications of the Chemical-Radiation Supervisor were stated in our letter of March 16, 1977.
The reload consists of replacing 50 batch 1 assemblies with an equal number of fresh batch 4 assemblies and reshuffling the batch 2 and 3 assemblies to new locations.
The batch 4 assemblies, containing 3.18 wtG uranium -235 enriched fuel, will be placed in the periphery of the core and in eight interior locations.
ine licensee has proposed the following chances to the Technical Specifications (Reference 1):
(a) Change of pressure-temperature limit curve.
(b)
Chance of reactor power-power imbalance limit curve.
(c) Change of power level limit for one reactor coolant pump operatinn in each loop.
(d) Chan::e of protective system max.inum allowable setpoints for rsctor pressure-temperature and core imbalance.
(e) Change of the quadrant power allowable limit.
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L (f) Change of position limits for regulating control rods.
(g)
Establishment of withdrawal / insertion limits for axial power shaping control rods.
-(h)
Change power-imbalance envelope.
These changes resulted from the following r 3difications of the operating procedures and computational methods proposed in the Cycle 2 reload submittal:
(a) Use of 104.4% of the first core design flow in,the Cycle 2 analyses.
(b) U w of the FLAME computer code in setting the Technical Specification limits.
(c) Use of the BAW-2 CHF correlation in the Derarture from Nucleate Boiling Ratio (DNBR) analysis.
(d) Modification of accounting for fuel densification in the DNBR analysis.
(e)
Decoupling regulating rod banks 6 and 7.
The analyses presented by the licensee in support of the proposed Technical Specification changes (Reference 2) were reviewed by us cnd found to be acceptable for all transients. The bases for the acceptability are discussed in this Safety Evaluation (SE).
2.0 Evaluation O
2.1 Fuel System The batch 4 fuel assemblies incorporated only minor design modifications to the spacer grid corner cells and to the end fittings.
These modifications improved seismic capability of the assemblies, but did not alter their mechanical interchangeability.
Slightly lower hydetulic resistance exhibited by the batch 4 assemblies re:ulted in a minor increase in flori. This change has been considered in the Cycle 2 core fica distribution analysis.
However, no credit has been' taken for the increase in system flow that resulted from the reduction total core pressure d, rop.
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' 1 Cladding collapse' analyses have indicated that the fuel in batches i
2 and 3 is more limiting than the fuel in batch 4.
However, it was demonstrated that the exposure time for the fuel in Cycle 2-will not er.ceed the time at which cladding collapse is predicted (35,000 EFPH).
It was also shown that the fuel will not reach the burnup levels at which plastic cladding strain exceeds 1 percent.
Some modificationt were introduced in calculating power spikes due to fuel densification.
The changes consisted of changing-the values of Fg (fraction of rods having a gap in fuel stack) and Fk (fraction of gaps within gap size interval k) parameters in the equation for the probability of gap' occurrence.
These changes have been previously
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reviewed and approved by the NRC staff.
The fuel temperature analysis performed by the licensee has indicated that the linear heat generation capability for Cycle 2 fuel is
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greater than the capability determined for Cycle 1 fuel.
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2.2 Nuclear Characteristics There were no.,ignificant changes in core design between Cycles 1 and 2.
In both cases all the important core physics parameters were calculated using the same calculational methods.
Slight differences in the core physics parameters were due mainly to the fact-that' the core has not yet reached an equ. :brium cycle.
Due to changes in isoto;)ics and the radial flux dic. ibution, the Beginnina of Cycle' (B0C) hot full power control roc worths are generally lower for Cycle 2 than f or Cycle 1.
However, the Cycle 2 control rod worths are sufficient to maintain the required shutdown margin of 1.0f, ak/k.
The contral rod shutdown margin calculations include two conservative assumptions:
(a) 10" uncertainty on net rod worth and (b) flux redistribution cenalty due to the use of a two dimensional model in the shutdo'.;n analysis.
The worth of e.iected and stuck rods were calculated to be less in Cycle 2 than in Cycle 1 for all pcwer levels.
This inf orration nrov""; ccervative input to the control rod accident analy c 3 The Doppler and rroderator coefficients and xenon worths are similar in both cyclcs.
However, the coefficients are more negative than tne values given in the Final Safety Analysis Report (FSAR) (Reference 3).
This also introduces an additional degree of conservatism in the accident analyses.
Similarly, lower boron worth at the begint ing of Cycle 2 ontributes further to the conservatism of the analyse:..
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2.3 Thegn,1-J!ydrauli s, f4 The inajor change in thermal hydraulic design evaluation consisted of' using the BA'!-2 CI'F correlation instead of the W-3 correlatica esployed. in the previously reported evaluations.
The BAW-2 correia-tion has been approved for use in the DilBR analysis (Reference 4) and its use is justified.
The Cycle ? cualysis was based on a flow rate of 104.4% of the design flow used in Reference S.
This change is conservative since the actual teasured flou significantly exceeds this value.
In addition, 2.3#; bypass flow was used in calculating core flow to account for orifico rod removal.
This assumption introduced additional conser-vatisa into the enalyses.
In determining ninimum D::CR, a densification power spike was not considered.
}lc. crer, the fuel densification ef fect was implicitly included through the reduction of active fuel length (1.56% for batches 2 and 3 and 1.33l for batch 4).
The DilBR calculation also included a rcd bow penalty that is burnup deoendent.
For all analyses that define plant operating limits cnd for all analy7ed tranaients, except for flux / flow trip setpoint affalysis, an ll.?S rod bow penalty was used.
This penalty corresponds to a burnup of 33,000 MHD/fifU and has been partially offset as discussed in Reference 9.
In the flux /flou trip setpoint analysis the rod how penalty of
- 7. 9'.; was assund.
This penelty was associated with the naximum burnup in the hatch 4 fuel nast:::blies which will exhibit the anxi /.a radial-local pcak (uring Cycic 2 and which were limiting for Din
anrlysis.
In cc:1p> ting the flux /ficw trip set points, flow measure-ment error and noise us.re f cctored in.
In addition, the analysis was extended in time to inclv:!e the power decrease due to reactor trip.
This acccunted for the delay time between control rod release and the effective insertion of the negative reactivity into the core, and provided additional conservatism to the analysis.
2.4 Accidtat Antivr.is Tha licensce ht.s evalur ted the effects of the Cycle 2 reload on the accid at; and i m ients cac,1yred in FSN!.
It was shown that tha results o' Cych 7 a wlyse; wm e more conservative than the cases analy:.cd in ti.t roterence. cycle.
In all case the initial nucicer pare.eters in Cycle 2 ucre within the hounds ossumed in the ref er'ince cycle cnalyses.
!!e concur that the accident analyses in the Cycle
? relord.ubmitt al have boun performed conservatively Wp:lts
-of th: analyser are acceptable.
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3.0~ Physics Startup Testing The ph?/ sics startup -test prngram, as proposcu by the licensee (References 2 and 10) has becn reviewed.
~ A discussion with the licensee resulted in further clarification of the critical boron concentration measurerrents and the Control Rod Assembly (CRA) Group worth measurements.
A critical-boron concentration iaeasurement with Grcups 7, 6 and 5 inserted will be nnde and compared to the predicted value with an; acceptance criteria of 100 ppm.
If a rod drop reactivity measurement is verify that the necessary shutdown needed, it will be u
n margin is availab1r licensce also agreed to submit a physics startup report with vs of criticality instead of incorporati".9 the results in the Ah, irt.
We find the physics startup test progri a as sub:1.iti.ed in twa rences 2 and 10, plus the above-racutioned clari f ication, to be acceptchle.
4.0 Technical Specification Chennes In Reference 1 the licensee has proposed to amend.the follouing sections of the Technical Specifications for Rancho Seco:
2.1 Safoty Limits, Reactor Core '
2.3 Liniting Safety System Settings, protective Instrumentaticn 2.5.2 Control P.od Group and Power Distribution Limit The pri. pond chcn< '. are listed in the intrcduction section of this Sr.
i n sr. ttin th lechnical Specifica:.icn limits, the FL/ME cuivu er code '(i.eierences 6 cud 7) was used.
This code has been revi: <cd a: d ti'pcoved by the !!RC (Reference 8).
Af ter revirsing iba bases Ond the supporting analys.cs submitt'd by the liu risce in p.cfcrpnces 1 and 2, we conclude that the propwed a'nensent confcr:.v. to the criteria specified in 10 CFR 50 and is accep ::ble for operating narcho Seco for Cycle 2.
He also revic-red th" littnv.n's p',oposed change ubich states that the Chemical-hadiation Supe:vit.or r. hell' amet or encecd the tinima requirement of Regulator;/
Gui/c 1.0, and ue find this to be acceptable.
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m 6-b.0 -Environmnntal Considerations 110 have determined that the auendment does not authorize a change in effluent types or total amunts nor an increase in power level and uill not result in any sir:nificant environmental impact.
!!aving made this. determination, we have further concluded that the amenc' ment involves an action which is insianificant from the stendpoint of envircom ntti impact and pursuani. to 10 CFR E51.5(d)(4), that an eiivirorc Atal impact statement or negative dcclaration and environmental impact-nppiaisal need not be prepared in connection.with the issuance of this cuendr;?nt.
G. 0 Conclusion
!!e have concluded, bcsed on the considerations discussed above, that:
(1) br ccur.e the amndment does not involve a significant incre'ese' in the pr ehability or consequences of accidents previously conshr ed and does not involve a significant decrea';e in a safety margin, th?-
tuendment does not involve a significent hazards consideration, (2) thorn is tcasonv51e tssurence that the health and safety of the public will not he er.dbogered by operation in the proposed manrer, and (3) such activitics will be conducted in ctwpliance with the Cten:ission's regulations and the issuance of this amendment uill not be inimic.al to the coa:an defence and security or"to the health!
and safety of the public.
Date: October 7,1977 9
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References
- 1. ' Letter to R. el. Reid (iiRC) from J. J.11attimoe (Sacrc-'nto Manicipal Utility District), dated June-24, 1977, t ns.nitting technical specification changes.
2.
BAW-1460, " Rancho' Seco l'uclear Generating Station, Unit 1, Cycle'2 Reload Report", dated June 1977.
3.
FSAR for Rancho Seco Nuclear Generating Station, Unit 1, (Docket No. 50-312), Volume 1.
4.
10 CFR 50, Appendix K. I.C.4.
5.
DAW-1393, " Rancho Seco Unit 1 Fuel Densification Report,"
dated June,1973.
6.
BAW-10124A, "FLAPE-Three-Dimensional flodal Code for Calculating Reactivity and Power Distribution," dated August 1976.
7.
BAW-10125, " Verification of Three-Dimensior.al, FLAf1E Code,"
dated A" gust 1976.
8.
Letter to K. E. Suhrke (B&W} from J. F. Stolz (I4RC), ddted liay 20, 1976, transmitting, " Review of Topical Reports BAW-10124 and BAW-10125P."
9.
Memorandum for D. B. Vassallo from D. R. Ross, dated February 16,1977, " Interim Safety Evaluation Report on the Ef fe:ts cf Fuel Rod Bowing on Thermal Margin Calculations for Light u ter Reactors, Revision 1."
10.
Letter from William C. Walbridge (Sacramento Municipal Utility District) to R. W. Reid (?!RC), dated August 29, 1977, transmitting additicaal information reiative to zero power physics tests and power tests.
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