B11660, Provides Informational Update of Analysis of Plant Auxiliary Feedwater Sys.Increased Primary Sys Pressure Predicted for Loss of Main Feedwater Event Has No Unacceptable Safety Impact

Provides Informational Update of Analysis of Plant Auxiliary Feedwater Sys.Increased Primary Sys Pressure Predicted for Loss of Main Feedwater Event Has No Unacceptable Safety Impact

ML20134M620
Person / Time
Site:	Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png
Issue date:	08/28/1985
From:	Opeka J, Sears C CONNECTICUT YANKEE ATOMIC POWER CO.
To:	Zwolinski J Office of Nuclear Reactor Regulation
References
B11660, NUDOCS 8509040196
Download: ML20134M620 (3)

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CONNECTICUT YANKEE ATOMIC POWER COMPANY B E R L I N. CONNECTICUT P o Box 270 H ARTFoRD. CONNECTICUT 06141-0270 TELEPHONE 203-665-5000 August 28, 1985 Docket No. 50-213 Bil660 Director of Nuclear Reactor Regulation Attn: Mr. John A. Zwolinski, Chief Operating Reactors Branch No. 5 U.S. Nuclear Regulatory Commission Washington, DC 20555

Reference:

W.G. Counsil letter to D.M. Crutchfield, dated May 19, 1980;

Subject:

Haddam Neck Plant, Auxiliary Feedwater Systems.

Gentlemen:

Haddam Neck Plant Loss of Main Feedwater Analysis The purpose of this letter is to provide the NRC Staff with an informational update of the analysis presented in Reference 1 on the Haddam Neck Auxiliary Feedwater (AFW) system. The referenced letter provided AFW system information applicable to plant design basis transients and accident conditions.

This information was used to assess the performance of the AFW system and demonstrate the adequacy of the minimum available feedwater flow during loss of heat sink events. Adequacy was defined as the ability of the available flow to remove primary side heat to a degree that plant acceptance criteria for the events were not violated. The referenced letter assumed single failures and defined various conservatisms. The following criteria were used to determine if the minimum AFW flow rate was acceptable:

1. The pressurizer pressure corresponding to the PORV setpoint (2285 psia) is not reached as a result of low steam generator inventory. This criterion was self-imposed.

2. No DNB condition is experienced at the clad surface of any fuel rod in the Core.

3. Sufficient steam generator level remains ( 10% of the height of the U-tubes) to remove the primary side heat generated.

The bounding case to assess the adequacy of minimum AFW flow is the Loss of Main Feedwater Event. This event was analyzed by Connecticut Yankee Atomic Power Company (CYAPCO) using the RETRAN $1 MOD 2 computer code. Cases were evaluated from an initial condition of 102 % power assuming an 8509040196 050020 \

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instantaneous loss of all main feedwater. The reactor was assumed to be tripped due to the coincidence of low steam generator level and steam flow / feed flow 1 mismatch. Only one of the two AFW pumps was assumed to operate. Both manual and automatic initiation of AFW were considered.

The results of these analyses were that the above acceptance criteria were satisfied for either manual initiation of AFW at 10 minutes or automatic initiation of AFW when the wide range water level in two steam generators dropped to 45E Subsequent to the referenced submittal, additional in-house Loss of Main Feed Water analyses have been performed to evaluate the effects of:

1. The automatic trip cf the reactor coolant pumps in loops 1 and 3 following reactor trip,

2. three loop operation, and

3. a reduction in the AFW flow rate that was necessary to prevent Terry turbine overspeed.

These analyses, which were also performed with RETRAN $1 MOD 2, predict higher primary system pressure than those transmitted to the Staff, but the results still satisfy the acceptance criteria listed above.

Another impact on these analyses occurred as a result of the recent installation of RETRAN $2 MOD 3 on our computer. One of our software QA requirements of installation of computer codes is to determine the effect of the new code version on analyses performed using older versions of the computer code. Performing the analysis with RETRAN $2 MOD 3 results in the predicted primary system pressure exceeding the PORV setpoint for this event initiated from both 4 loop and 3 loop operatior. Even though the PORVs were not assumed to open in the analyses, the predicted system pressure of 2342 psia for three loop operation and 2337 psia for four loop operation did not exceed the safety valve setpoint of 2500 psia. Although the new results exceed one of the criteria established by CYAPCO, the predicted pressure did not exceed the SRP criterion of 110% of design pressure.

This result is believed to be caused by the difference in the way water properties are modeled in the two codes. RETRAN $1 used tabular values which may result in noncontinuous derivatives of water properties. RETRAN 62 uses functional fits for both properties and their derivatives. The pressurizer model requires the property derivatives. Therefore, the difference in derivatives results in differences in calculated pressure.

The reason for CYAPCO's more restrictive self-imposed maximum pressure criterion was the desire to limit the frequency of challenges to the PORVs.

However, challenges to the PORVs has not been a problem at the Haddam Neck Plant. Only two actuations of the PORVs have been experienced at the Haddam Neck Plant since 1981. In addition, the operator has ample indication of flow through the PORVs, including annunciators on PORV open setpoint pressure,

, pressurizer relief tank high temperature, high level, and high pressure, an

! acoustic monitoring system, and control board indication of pressurizer relief tank pressure, temperature, and level indication. Given these indications, the

1 . .e operator can stop the flow through the PORVs by closing the PORY block valves.

Emergency Operating Procedures include specific actions required for a stuck-open PORV. and operators have received extensive training in this area.

Therefore, the increased primary system pressure predicted for the Loss of Main Feedwater event has no unacceptable safety impact.

! No Staff action is required or requested by this letter; it is being provided for information only.-

. Very truly yours, CONNECTICUT YANKEE ATOMIC POWER COMPANY 3.F. Opeka bM '

Senior Vice President bt By: C.F. Sears Vice President 4

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