ML20062A422

Safety Evaluation Rept Supporting Amend 43 to Lic DPR-33 Covering Pump Seizure Accident,Inadvertent Cold Loop Startup Transient & Monitoring of Safety Margins for 1 Loop Oper

ML20062A422
Person / Time
Site:	Browns Ferry
Issue date:	09/29/1978
From:	Office of Nuclear Reactor Regulation
To:
Shared Package
ML20062A414	List: ML20062A406 ML20062A417 ML20062A422 ML20062A431
References
NUDOCS 7810130205
Download: ML20062A422 (4)

Text

-

?

/ >

UNITED STATES j

? ,j NUCLEAR REGULATORY CCMMISSION wAswiwaTom, c. c.2asas  !

I

%, %s f SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMEN 0 MENT NO. 43 FACILITY LICENSE NO. DPR-33 TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT, UNIT NO. 1 DOCKET NO. 50-259 Introduction On Septemcer 19, 1978, we issued Amendment No. 41 to Facility Operating License No. DPR-33 which permitted TVA to operate Browns Ferry Unit No.1 (BF-1) with one recirculation loop out of service for the remainder of the current fuel cycle (cycle 2) but with the restriction that power level be limited to a maximum of 50% of licensed power. Power was limited to 50% of full licensed pcwer by the new Technical Specifications due to lack of an acceptable analysis of the locked pump rotor accident for one Icop operation. Our SER issued with the Technical Specification changes gave the bases for acceptability of the 50% power limit. That SER also discussed the bases for a restriction that pcwer c:uld not exceed eB% of full licensed pcwer until further infor nation regarding the inadventent pump startup transient analysis was provided.

Su: mary We have reviewed the TVA submittal of September 28, 1978, requesting deletion of the above described power restrictions. We find it acceptable to remove both of the pcwer restrictions on the bases discussed belew, provided that the maximum power level not exceed 82% of full licensed power for the reasons discussed belcw.

Evaluation  !

One Pum: Sei:ure Accident TVA has sucmitted results of calculations for the pumo sei:ure accident assuming one loop operation (submittal of September 23,1978). These calculations utili:ed tne standard, conservative, plant scecific inputs that were used for the transients and accidents nat were analy:ed -

for :he lates: 3F-1 relcad (Amencment 35 issued January 10,1978).

fpg These calculations utilized the General Electric REDY c:de to mcdel the  ;

transient behavior of the reactor core. No model chances were made to the '

75/OlWeede for this use. The code is currently used for reload transient analyses perfor :ed for SF-1 and cther SWRs. . ump sei:ure was simulated in the same PCR t10eCA c f* .2 59 P 15o 9pf

acceptable manner as for previously perfomed pump seizure analyses with two pump operation (the method results in the pump stopping completely in less than one revolution). Appropriate input changes were made to, reflect operation with one recirculation loop. Although a review is currently in progress concerning continued acceptability of the REDY code for analysis of pressurization transients, this transient is characterized as a loss-of-flow event. REDY code results are currently accepted for such analyses.

REDY results were input into the General Electric SCAT 02 code which cal-culates themal margin (Minimum Critical power Ratio, MCpR). This code was not modified in any way for this purpose. SCAT 02 is currently used and accepted for BF-1 and other plant transient analyses that are per-formed for reload submittals.

Two sets of initial conditions were analy:ed (power-ficw initial conditions equal to 75% power - 58% flow and 82% power - 56% ficw). These conditions represent, respectively: (1) the maximum flow possible through the core (allowing for backflow through the idle jet pumps) due to one pump cpera-tion with power corresponding to that flow frem the highest (105%) power-ficw load line; and (2) an even higher power for the same one pump opera-tion (core flow is 2% lower due to increased flow resistance in the core due to added voiding at the higher power). We find these assumed initial conditions acceptable since they conservatively bound the highest power operation possible under one loop operation conditions.

The analyses indicate that the following sequence of events occurs: the operating pump seizes and external recirculation loop flow goes very rapidly to :ero; core flow continues due to momentum effects and buoyancy effects; as momentum is dissipated, core flow will decrease; as core flow decreases, voiding in the core will increase; increased voiding will reduce core thermal pcwer due to the feedback effect frem the negative void coefficient; and finally, the core will reach a new equilibrium power and flow detemined by the natural re-circulation characteristics of the core, wi.. no ficw in the external recirculation loop (flow will be up through the core, driven by buoyant force as water is heated and boiled in the core, then dcwn through the internal jet pumps, then repeat). Steam produced will continue to go cut the main steam line to the turbine / condenser system, then back to the reactor througn the feedwater system. Water level in the core will first increase due to the increased voiding, then will decrease as core pcwer decreases and stored hea; is removed, but the level fluctua-ions are not severe enough to trio (scram) ne reactor on either hign or icw water level. Neutr:n flux decreases cue to tne void feedback so no high flux scram occurs. pressure does not reach the high pressure scram point (in fact, pressure decreases) -

r ~ :e no isolation occurs and steam continues ficwing o ne taroine wnile steam production decreases as gewer decreases. Thus, no reactor scram is anticipated frem any signal.

. .- ~._ - . - - . . . . _ - - - . . - . __

, I 3

i 4

During the above sequence of events, the MCPR does not go below the Safety Limit MCPR, which is set to insure lack of any significant fuel damage.

Therefore, we conclude that no significant fuel damage will occur, and therefore the acceptance criteria is met for this accident -(f.e. , that i

releases remain below a small fraction of the requirments stated in 10  :

! CFR Part 100). '

i On the basis of the above described plant specific analyses which have 3 been provided, we conclude that the 50% power limit previously imposed on SF-1 for single loop operation for the remainder of the present cycle can be increased to 82% pcwer, the maximum power level for which the L consequences of the pump seizere accident have been demonstrated to be 4

acceptable.

I Inadvertent Cold Loco Startuo Transient -

TVA has closed the suction valve on the disabled loop and has locked out and tagge power to that valve and to the disabled loop's recirculation i pump. In addition, TVA has stated that "the brushes have been lifted from the recirculation pump M-G set and will not be reinstated until the next refueling outage. Replacement of the brushes takes several hours and could not be performed without authorization." Therefore, we consider unplanned startup of the idle loop to be incredible during the remainder of the present cycle. Furthermore, the discharge valve (whose internal failure and partial closure caused the loop's shutdown) cannot be repaired until the unit is.

i shut down at end-of-cycle. Therefore, planned loop startup will not occur during the remainder of the present cycle, and the attendant slight possibility of utilizing improper procedures during such a planned startup, and inadvertently introducing cold water suddenly into the primary system, will be avoided.

On the above stated bases we conclude that the occurrence of this transient is not credible during the remainder of the present cycle, and we find it-acceptable to delete the 68% maximum power restriction for one loop oper-4 ation stated in our previous SER.

Monitorina of Safety Marcins for One Loco Ocerstion I

Our orevicus SER, issued with Amendment No. 41, stated the bases for a ceptability of the 3F-1 MAPLHGR limits that will be observed during the remainder of this cycle for one loco operation. Those limits

conservatively account for the unavilability of one of the olants recirculatien icoas by assuming no credit for ccastdewn ficw frcm i

the unavailable loco folicwing a LCCA. We therefore cenclude that oceration with cne loco out of service, with thereduced limits, dces not result in any decrease in overall plant safety margins, i.e., that -

the reduced limits acceotably account for the reducticn in plant

souipment tnat is availaole.

?

1

~ '

-4 The plant is required to monitor ccmpliance with the MAPUGR limits every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. The process computer prints out a quantity called r MAPRAT, which is the ratio of the current (measured) MAPUGR to the current MAPUGR calculated limit. This is done every 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for the, 6 locations in the core that are closest to the limit. Values greater than 1.0 thus indicate a violation. The process computer has been, programed to assume the reduced limitsthat have been approved for one loop operation when calculating MAPRAT. We find the above described method of monitoring cc:npliance with the MAPUGR limits to be easy to monitor and acceptable.

We do not believe that overall plant safety margins have been decreased as stated in the first paragraph above. We note that we have previously approved operation of BF-1 at higher pcwer levels with the same method and frequency of monitoring for compliance with the MAPUGR limits.

Since safety margins has not been degraded by operation with a single loco, we find the method of monitoring equally to be acceptable for one loop operation.

Environmental Consideration We have determined that this amendment does not authorize a change in effluent types or total amounts nor an increase in power level authori:ed by the license and will not result in any significant environmental impact.

Having made this determination, we have further concluded that this amendment involves an action which is insignificant frem the standpoint of environmental impact and pursuant to 10 CFR 151.5(d)(4) that an environmental imcact statement, negative declaration, or environmental impact appraisal need not be prepared in connection with the issuance of this amendment.

Conclusion We have concluded, based on the considerations discussed above, that:

(1) because the amendment does not involve a significant increase in the probability of consequences of accidents previously considered and does not involve a significant decrease in a safety margin, the amendment

does not involve a significant ha

ards consideration, (2) there is reason-able assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (3) such activities ,

will be concucted in ceccliance with the Ccrmission's regulations and the issuance of these amendments will net be inimical to the comon cefense i and security or to the health and safety of the public.

Dated: Septamcer 29, 1978 l

l

_-. - .