ML19291C909
| ML19291C909 | |
| Person / Time | |
|---|---|
| Site: | Yankee Rowe |
| Issue date: | 11/21/1963 |
| From: | Bryan R US ATOMIC ENERGY COMMISSION (AEC) |
| To: | |
| Shared Package | |
| ML19291C906 | List: |
| References | |
| NUDOCS 8011140122 | |
| Download: ML19291C909 (4) | |
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UNITED STATES ATOMIC ENERGY COWISSION HAZARDS AN ALYSIS BY THE RESE ARCH AND POTER REACTOR S *FETY BRANCH DIVISION OF LICENSING AND REGULATION IN THE MATTER OF YANKEE ATOMIC ELECTRIC COMPANY DOCKET NO. 50-29 Introduction By Ictter dated July 17, 1963, Yankee Atomic Electric Company submitted Amendment No. 45 to its license application and requested authorization to increase the maximum steady state power IcVel of their reactor from 540 to 600 Me thermal for operation during the life of Core III.
As part of this request, Yankee furnished proposed superseding and additional pages which would be incorporated into the Yankee " Final Hazards Summary Report" if authorization to increase the power level is granted.
These pages describe the changes in the facility operations characteristics as a result of the proposed reactor power increase, and analyze the effects that the increased power level would have on safety.
Operation of the reactor with Core III installed is scheduled to begin Ic.
this year.
Core III is a multi-region core which will initially require t ne use of boric acid during power operation in the manner described in Proposed Thenge No. 36, which was authori:cd by the Commission on September 13, 1965.
(See our related hazards analysis.)
If the proposed increase in power level from 540 to 600 he thermal is authorized; Yanket would initially operate the reactor with Core 111 at a maximum power Ic eel of 540 Pw thermal in order to s crify the calculated values of hot channel facters through use of the reactor in-core inst rumentation before proceeding to a higher power level, giscussion 1lestinghouse and Stone and h'ebster were engaged by Yankee to review the de s i pt of the turbine and secondary plant at their facility in light of the proposed power increase, These organizations concluded that, with minor ordifications which Yankee has performed during their refueling operation,
_he turbine and secondary plant are adequate for operation at the increased owe r level.
These modifications consist urimarily of alterations in the niisture separators and feed water regulating system.
'Sc crincipal safety consideration regarding the proposed power increase ec:.ce rns the thermal properties of the core.
Yankee has recalculated the 00 1114 O/h l
l
^
-z.
DND ratios for 6001k thermal operation assuming that the power will not be increased above 540 Me until after that point in operation when the first five of the seven conttol rod groups are withdrawn from the core, By limit-ing the higher power to the ame when the core has few or no cont ral tods inserted, the reactor would operate at the higher power when the hot channel factors are of a considerably snailer s alue than they would ha at the be gin -
ning of cire life.
The calculated minimum steady state DNB ratio duiing the proposcai 600 Me thernc1 operation with Core III is 2.35.
This value was obtained by using the Westinghouse W-2 correlation.
Th e W-2 i s a "bes t fit" corIelatien, and the minimum steady state value corresponds to a ratio of 1 SS at a 951 con-fidence level.
(The minimum steady state value of the DNS ratio during Core Il life was estinated to be 2,2 (1,,"e at the 954 confidente level) prior to operation, and was calculated to be 2. 5 on the basis of in-core instrumenta-tion measurements made during Core II life,)
Using a conn rvative combination of pressure and temperature variations and assuming that the teactor actually scrans at 116% of full power instead of the maximum set point of 10Sh the
- -E overpower DNE ratio would be 1.74, or 1,39 at the 95% confidence level, The operating license for the Yankee reactor requires that the flux distribu.
w tion be measured at Icast once every 1000 equivalent full power hours.
Yankee has indicated that they will measure the flux distribution considerably more frequently than this during the times of Core III life when the DNB ratios are expected to be the smallest, The information obtained fIom these measure-ments would be used to insure that the steady state DNB ratio d:es not becore less than 2.35, Yankee would also insure that the exit tempe:ature of
-t e hottest channel does not exceed 611* F.
The saturation temocrature o f s (.
the operating pressure of 2000 psi is 636* F, Thus, although thc re cou; atbe sure local boiling, the:c wculd be no bu}k boiling in the core region if C.ese criteria ate followed.
Pane r would not be raised ebc.c f 4 % thermai enm me as uu nents hm.e con-fi rrsd that the DSD ratio alm not be less thEn 2,35. and the hot ch ann:J e xi t icancrature ill not be grcate r than 611' F dunng 600 % thermal ope > at ion.
Honever if the power were inneased and neasurements during subsequent ope ra-y tions indicated that these li tit s w ould be ex;c c ded, the power le vel would be 2 educed to insure conformance with these limits, Although operation at a DSD ratio as low as 2.35 has not been car: 1cd out in this acactor, that value is in excess of the authorited Cole II minimum talue and, we believe, adequately safe during steady state operation.
The accidents which would be affec. icd by the increase in power have been re, anc q :ed by Yankee.
Due prima:ily t o the lowe: hot channel factors, the fuel and cent rol rod temperatures reached during c Lcse accidents would not be sig-ni ficant'jy d _fferent from those p re neus]> estimat ed for 540 'k thermal opera.
I.cn.
Assuming that depsitu:e fica nucleate boiling occurs when the DND ratio he cc mes J ess than 1.25, the maximum clad tempe:ature f o r a lo5 s =of flow acci-s.nt was c alculated to be 14 S0' F, a value less than that at which damage to
3 the cladding would be expected to occur.
A loss-o f-coolant accident, in which it is assuned that the charging and safety injection pumps operate properly, is similarly calculated to result in a temperature below that at which failure of the fuel element cladding would be expected; and the melting point of the Ag-In-Cd control rods (1500* F) would not be approached in the event of such an a cci dent.
During the recent refueling, Yankee has installed contro] rods with hafnium absorber sections in 22 of the 24 control rods positions.
The use of hafnium, which has a much higher melting temperature than Ag-In-Cd. furthe r lessens the possibility of control rod melting during an accident.
Tne other two of the 24 positions would have control rods with redesigned Ag In-Cd absorber sc etions.
The Ag-In-Cd would be clad with 30 tils of Incenel instead f the 0.5-mil diffusion-bonded nickel plate used with Cores I and II 1his
- 1 adding is believed to be adeqcate to prevent the escape of rndicactive silver into the main coolant system as has occurred with control rods of the present desi gn.
The larger amount of stored energy in the primary system due to the proposed increase in average primary system temperature would not result in as high a vapor container pressure as was previously calculated in the event of an accidental rupture of the primary system.
This is due to the fact that con-servative assumptions regarding the water volume in the primary system were used in previous calculations; whereas, in the calculations for 600 Me opera-tions, the actual water volume of the installed system was used, The potential downwind radiation doses which could result from an accident which represents the upper lindt of public hazard incident to cperations at 600 ne have been analyzed. Yankee has concluded, and we have verifieo, +t at the potential consequences of such an accident are within guidelines set in Part 100 of the Commission's regalations.
3.dvisory Committee on Reactor Safeguards Rey,oy Anendment No. 45 was considered by the Advisory Committee on Reactor Safeguards at it s September 1963 meeting.
In its report to the Commission regarding the proposed incre: sed operating power level, the Coreuttee stated:
"The Committee believes that tbc proposed operation does not present an undue hazard to the health and safety of the public."
A copy of this letter to the Commission is attached.
Con clusion operation at the proposed higher power level of 600 Me will obviously result an a higher average power per fuel rod than that which occurred during Core I?
li fe.
This would be compensated for during Cote III operation by flux flat-
- ening, which will result from the use of the segmented core loading, and by not operating the reactor at the increased power level during the initial
4 period of core life when the hot channel factors are the highest.
On the
.:asi s of our review and that of the ACRS, we have concluded that operation of the reactor at 600 tta thermal with Cora III will not result in a sub-stential change in potential public hazard from that which was previously considered to be acceptable and that there is rearonabic asserence that the health and safety of the public wi21 not be endangered.
'Oririnni signed by I:%rt U " -:, n Robe rt H. B ryan, Ch i e f Research G Power Peactor Safety Branch Division of Licensin; 6 Pegulatien Date: f!OV 211953
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