ML19208A916

Forwards Corrected Version of loss-of-feedwater Flow Transient Analysis,Originally Submitted on 740329

ML19208A916
Person / Time
Site:	Yankee Rowe
Issue date:	09/12/1979
From:	Vandenburgh D YANKEE ATOMIC ELECTRIC CO.
To:	Ziemann D Office of Nuclear Reactor Regulation
References
TASK-15-05, TASK-15-5, TASK-RR WYR-79-104, NUDOCS 7909180268
Download: ML19208A916 (12)

Text

Telephone 6!7 366-9011 rwx 710-390-0739 YANKEE ATOMIC ELECTRIC COMPANY B.3.2.1 WYR 79-104 Ya$$

20 Turnpike Road Westborough, Massachusetts 01581 mmes

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September 12, 1979 United States Nuclear Regulatory Cccmission Washington, D. C.

20555 Attention: Office of Nuclear Reactor Regulation Mr. Dennis L. Ziemann, Chief Operating Reactors Eranch #2 Division of Operating Reactors

References:

(1) License No. DPR-3 (Docket No.

(2) YAEC letter R. E. Groce to USNRC, A. Burger dated April 23, 1979 (3)

" Yankee Nuclear Power Station Core XIX Performance Analysis", YAEC-ll62, September 1978 (4) YAEC letter to USNRC dated March 29, 1974 Proposed Change No. 115 (5) YAEC letter to L'SNRC dated September 8,1978 (WYR 78-79) Proposed Change No. 163, Supplement No. 1 (6) USNRC letter to YAEC dated December 6, 1978 Amendment No. 54 to License No. DPR-3

Subject:

Yankee Rcwe Loss of Feedwater Analysis As discussed in Reference 2, during the 7eview of the safety analyses that support the operation of Yankee Rowe, tso errors were discovered in the loss of feedvater analysis. These errors, an error in assumption concerning reactor coolant pump operation following reactor trip and an error in the version of the code used in the analysis, were described in detail in Reference 2.

The purpose of this letter is to submit the corrected loss of feedwater analysis.

To recall, the loss of feedwater analysis in error was submitted to NRC as part of the Core XI reload submittal via Reference 4.

Subsequent cores, including the present core (Core XIV), were licensed via reference to this Core XI analysis because the refueling changes either 1) did not affect this transient, or 2) the Core XI analysis assuuptions bounF.ed the applicable latter core characteristics. The evaluation for the current cycle, Core XIV, is contained in Reference 3, which was submitted to NRC via Reference 5 and subsequently approved via Reference 6.

The results of the corrected analysis, as provided in tne attachment, show acceptable consequences for this event and continue to show that the loss of feeduater event is a non-limiting transient for Yankee Rowe.

3G0005 7 9 091802(a p

United States Nuclear Regulatory Commission September 12, 1979 Attention: Office of Nuclear Reactor Regulation Page Two Therefore, the safety of the plant is shown by the re-analysis to be assured in the event of a total loss of feedwater event.

If you have any questions concerning this letter, please call Mr. Jamee R. Chapman at our Engineering Office, 25 Research Drive, Westboro, Massachusetts 01581, (617) 366-9011, extension 218.

Very truly yours, YANKEE ATOMIC ELECTRIC COMPANY D. E. Vandenburgh Senior Vice President COMMONWEALTH OF MASSACHUSETTS)

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COUNTY OF WORCESTER

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Then personally appeared before te, D. E. Vandenburgh, who, being duly sworn did state that he is Senior Vice President of Yankee Atomic Electric Company, that he is duly authori cd to execute and file the foregoing information in the name and on the behalf of Yankee Atomic Electric Company, and that the statements the 2n are true to the best of his knowledge and belief.

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. Loss of Feedwater Flow Transient Analysis General The feedwater system o' Yankee Rowe is designed to provide a continuous flow of water to the four steam gencrators during operation of the plant. A rapid and large decrease in feedwater flow when operating at pos ar without a corresponding reduction in steam flow would lead to a decrease in water inventory of the steam generators. The main feedwater system is designed to prevent a total loss of feedwater flow. Three electric motor driven parallel pumps with a common suctL.1 and discharge header provide normal feedwater flow. The pumps take suction from the condenser via three parallel condensate pumps.

Flow is controlled by four parate control valves designed to fail in position.

In the event of a total loss of main feedwater, an auxiliary feedwater system (AFWS) is available to provide water to the steam generators and the reactor protection system responds automatically to trip the plant.

The auxiliary feedwater system contains one positive displacement steam driven pump (capacity greater than 80 gpa) ahich feeds the main feedwater system downstream of the boiler feed pump discharge header but upstream of the feed control valses. The auxiliary feed pump takes steam f rom the main steam header and suction f rom the desineralized water storage tank. A backup source to supply water to the steam generators in the event of failures in the AFWS is the plant's three (3) charging pumps with a total capacity of approximately 100 gpm (33 gpc/ pump). The system is connected permanently by a spool piece that connects to the main feed line header.

The water supply to the charging pumps is the 135,000 gallon Primary Water Storage Tank. The High Pressure Safety Injection and Low Preesure Safety Injection pumps provide another uackup source to supply water to the AFWS

-through the permanently connected spool piece as used f or the charging pump path. The flow available from the combination of a single high pressure safety injection (HPSI) pump and low pressure safety injection (LPSI) pump is greater than 100 gpm, and there is a total of three (3) HPSI pumps and LPSI pumps.

The reactor protection system protects the plant through a series of reactor trips.

Specifically for the loss of feedwater transient, a low steam generator water level provides the signal to trip the plant.

The analysis presented assumes a total loss of feedwater event with makeup by the auxiliary feedwater system. The major assumptions are as follows:

1) thin feedwater flow is instantaneously reduced to sero at cime zero.

2)

The steam rypass system is assumed to function nor;.:lly after turbine trip.

This will accentuate primary and secondary WOOO7 cooldown by releasing more energy through the bypass, thus minimizing the steam generator liquid inventory.

3)

The moderator temperature coefficient is significantly 1.ms negative than expected, including uncertainties (+0.5x10-43pfo ),

7 4)

The fuel temperature coefficient is that for beginning-of-life fuel for Core XIV (-1.42x10-5 gyop),

5)

The pressurizer is assumed to be in the manual operation mode.

Thus, no credit is taken for the effect of the spray, letdown, and charging systems.

6)

The nuclear steam supply system parameters are:

1) reactor power is 600 MWt per Technical Specifications plus a 3% uncertainty, thus yielding 618 MWt,

11) core inlet temperature is 515 F per Technical Specifications plus a 4 F uncertainty yielding 519 F, iii) reactor coolant system pressure is a nominal 2000 psig, iv) steam generator pressure is the raxinum expected for normal operations; 600 psia, and v) reactor coolant flow is 35 x 106 lbs/hr which is the Technical Specification minimum allowable core flow.

7)

Low steam generator water level signal to trip the reactor occurs 18 seconds after the loss of main feedwater and includes a 10% uncertainty.

8)

Auxiliary feedwater is restarted 15 minutes after the loss of main feedwater and provides flow at 80 gpm.

9)

Restarting of the two reactor coolant pumps which were powered from the turbine-generator and tripped 60 seconds after reactor trip, occurs at 10 minutes af ter the loss of main feedwater.

10)

For the pressure and temperature transients, the product of tha heat transfer coefficient and heat transfer area (UA) was alloued to vary with stean generator liquid level.

For the inventory calculation, UA was held constant at its full power steady-state value as calculated by GEMINI-II.

SG0008 METHOD OF ANALYSIS The analysis of the loss of feedwater flow transient was perforned using two basic procedures:

1) the GEMINI-II cocputer code, and 2) manual calculations which considered the total conservation of mass and energy of both the primary system and the secondary system.

The GEMINI-II computer code (Reference 1) uas used to determine the initial transient response of the primary and secondary system and to provide a comparison to the previous analysis, Core XI, which was submitted to NRC via Reference 2.

The manual calculations consider the total mass ar.d energy of the primary and secondary systens and supplement the GEMINI analyses by accounting for the following parameters not modeled in GEMINI-II:

1) reactor coolant pump energy input to primary system, 2) steam generator secondary side structural heat capacity, and 3)

ANS 5.1 decay heat (Reference 3).

Consideration of these paraneters yields a more accurate calculation of the steam generator liquid inventory.

Parameter 1, reactor coolant pump energy input to primary systen, results n a nore rapid decrease in the steaa generator liquid inventory.

Parameter 2, s tean generator secondary side structural heat capacity, reduces the rate of steam generator liqui inventory decrease.

Paraceter 3, ANS 5.1 decay heat, was used because it is the best decay heat model currently available.

Primary systen structural heat capacity is included in the GEMINI-II codel and hence does not need to be separately included in the tanual calculations.

RESULTS The primary coolant pressure versus time for the loss of feedwater transient is shown in Figure 1.

The primarv pressure attained a peak of 217d psia at 22 seconds with a rapid decay thereaf ter finally leveling af ter 60 seconds to approxinately 1860 psia.

The peak pressure of 2178 psia is significantly less than the opening pressure of the pressurizer power operated relief valve, 2400 psia, and hence the valve is not actuated.

Figure 2 illustrates the secondary pressure versus time for the transient.

The secondary pressure surges to approximately 890 psia upon tripping of the turbine throttle valves (20 seconds into transient) and in the long tern levels out at approximately the steam bypass systen set point.

The secondary pressure is uaintained at this level by neans of t..e

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steam bypass system, approx.cately 760 psis.

The core average coolant temperature versus time is shown in Figure 3.

After the initial upswing before reactor trip, the average temperature levels at approxiuately 533 F and t.en gradually decays to a teuperature consistent with the secondary steam by pass systen set point.

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Steam generator liquid inventory as a function cf tice is shown in Figures 4 and 5.

Figure 4 is the inventory as calculated by the GEMINI-II code. The inventory cass 13ss was maximized per assumption 101 so as not to account for heat transfer degradation with steam generator water level.

The inventory begins to increase af ter approximately 57 minutes with the minimum inventory corrasponding to 26 percent of the initial inventory.

The results or the manual calculations are provided in Figure 5, which shows the steam generator liquid inventory history considering the effects of pump heat, steam generator secondary side structural heat capacity, and ANS 5.1 decay heat, as previously discussed. The results show inventory recovery at 74 minutes with the minimum steam generator liquid inventcry being 14 percent of the initial inventory. This calculation remains conservative since heat losses to the ambient from both the prinary and secondary systems were not included.

CONCLUSIONS The combination of the reactor protection system and the auxiliary feedwater system assure the integrity of the core, and pricary and secondary system pressure boundaries by 1) reactor trip on low steam generator water level, and 2) auxiliary feedwater flow sufficient to assure adequate stean generator liquid inventory for pritary system cooldown, decay heat remo va l, and reactoc coolant pucp heat removal for the entire course of the event.

360010 REFERENCES g

1)

YAEC-1068, " GEMINI-II, A Modified Version of the GEMINI Lomputer Program", Thomas R. Hencey, April 1974.

2)

Proposed Change No. 115, March 29, 1974.

3)

Proposed ANS Standard 5.1, " DECAY HEAT POWER IN LIGilT WATER REACTORS",

American Nuclear Society, September 1978.

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