B12501, Proposes Rev to Bases Section 2.4, Max Safety Settings - Protective Instrumentation. Rev Increases PORV Setpoint Above High Pressure Trip Setpoint

Proposes Rev to Bases Section 2.4, Max Safety Settings - Protective Instrumentation. Rev Increases PORV Setpoint Above High Pressure Trip Setpoint

ML20215B350
Person / Time
Site:	Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png
Issue date:	06/09/1987
From:	Mroczka E CONNECTICUT YANKEE ATOMIC POWER CO.
To:	NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
References
B12501, NUDOCS 8706170298
Download: ML20215B350 (4)

Text

3 CONNECTICUT YANKEE ATOMIC POWER COMPANY

' B E R L I N, CONN ECTICU T P o. BOX 270

• HARTFORD CONNECTICUT 06141-0270 TEL EPHONE 2caus som June 9,1987 Docket No. 50-213 B12501 Re: 10CFR50.36

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U.S. Nuclear Regulatory Commission  ;

Attn: Document Control Desk l Washington, D.C. 20535 l

Reference:

(1) 3. F. Opeka letter to Dr. T. E. Murley, " Power-Operated Relief Valve (PORV) Setpoint, Maintaining the Integrity of the Design Basis," dated August 29,1986.

Gentlemen: l 1

Haddam Neck Plant Maximum Safety Settings Connecticut Yankee Atomic Power Company hereby proposes to change Bases Section 2.4, Maximum Safety Settings - Protective Instrumentation, of -the Haddam Neck Plant Technical Specifications. As stated in 10CFR50.36, the

" Bases" are not a part of the technical specifications. The " Bases" describe aspects of the plant design bases in order to provide information about the context of associated Technical Specifications. This letter is therefore being forwarded as an informational letter to the staff and the change is requested to be included as part of the next approved amendment request.

Reference (1) established the framework for the proposed change, which would delete the sentence in the bases for the pressurizer pressure trip setpoint that -

references the Power Operated Relief Valve (PORV) set presure of 2270 psig.

The pressurizer pressure allowable of 2300 psig (which is unchanged) is no longer dependent on the PORY set pressure. Instead, the PORV set pressure will be increased to 2325 - 2350 psig in order to ensure an early reactor trip during a loss of load event. The proposed change will establish consistency between the bases of Technical Specificatica 2.4 and the design basis safety analysis.

The sentence to be removed credits the PORY for reducing the challenges to the high pressurizer pressure trip. Recent experience has shown that using the PORV to reduce reactor trips is not recommended. Af ter the TMI accident, it became evident that challenges that could lead to leakage or failure to rescat were more significant than additional reactor trips. Thus, the stated reason for the PORV setpoint is no longer desirable and this function of the PORV is no longer a basis for the trip setpoint.

8706170298Of0 13 PDR ADOCK O PDR P l l

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1 U. S. Nuclear Regulatory Commission B12501/Page 2 June 9,1987 In performing the re-analysis of the design basis accidents, it was found that  !

given the right set of conditions, the current PORV configuration could have an )

impact on the loss of load response. While the current PORV capacity and '

previous setpoint would not have caused safety limits to be exceeded during any design basis events, it may have led to an undesirable primary response in that the pressurizer could have been filled with water during a loss of load transient.

The major cause of this primary system response was the change in PORY capacity due to the replacement of the PORVs in 1977. This situation will be ,

corrected by increasing the PORY setpoint above the high pressure trip setpoint. j l

As stated above and in Reference (1), the change is reflected only in the Bases i Section of the Technical Specifications. Therefore, NRC approval prior to implementing the PORV setpoint change is not necessary. Accordingly, issuance of a license amendment reflecting the attached change is not needed prior to implementation of this change. We request the NRC Staff issue the attached change with the next license amendment.

If you should have any questions concerning this proposal, please let us know.

Very truly yours, CONNECTICUT YANKEE ATOMIC POWER COMPANY 1

% W E. I Kroczka (/

Senior Vice President cc: W. T. Russell, Region I Administrator F. M. Akstulewicz, Project Manager, Haddam Neck Plant P. D. Swetland, Resident inspector, Haddam Neck Plant Kevin McCarthy Director, Radiation Control Unit Department of Environmental Protection Hartford, Connecticut 06116

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Docket No. 50-213 B12501 Attachment 1 1

Haddam Neck Plant

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Maximum Safety Settings l 1

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1 June,1987

(1) Pressurizer Pressure. The high pressure reactor trip is established to maintain the DNB-ratio (see Section 2.2) above 1.30 following a large load mismatch between the reactor and turbine generator, such as a turbine trip without reactor trip. Setpoint and instrument errors of 30 psi are ircluded in the determination of this' setting. (The actual setpoint and instrument error totalis expected to be considerably less than 30 psi.)

(2) Pressurizer Level. The pressurizer high level reactor trip setting (86% of the compensated level range)is chosen to prevent water relief through the pressurizer code safety valves. This trip is particularly effective with respect to slowly i

developing load mismatches between the turbine generator and reactor. -The signal may be defeated with the reactor at least 1.5% Ak subcritical to allow cold (ambient temperature) reactor testing such as rod drop testing. The margin to criticality is chosen so as to maintain the reactor subcritical even if the highest worth rod were to be removed.

(3) Variable Low Pressure. The variable low pressure and nuclear overpower trips provide the basic protection for the reactor coolant system and the core. The manner in which protection is provided by these trips, over the full range of reactivity incidents, is discussed in detail in Section 7.2 of the Facility Description and Safety Analysis. As explained in that section, the purpose of the variable low pressure reactor trip is to protect the core against DNB or excessive core exit quality (see Section 2.2) resulting from those uncontrolled slow reactivity insertions which cause reactor coolant system temperature and pressure to increase unduly before an overpower reactor trip can occur. The for:aula for the trip setpoint, which is based on reactor core temperature rise

( AT) and reactor coolant system average  !

temperature, defines a minimum allowable f pressure for operation which is continually f compared to pressurizer pressure. The trip point d for a particular combination of variables is calculated by a simple analog conaputer. A reactor ,

trip occurs when tLe minimum all>wable pressure j rises above pressurizer pressure. The entire  ;

calculation, including both temperature and j pressure instrumentation, has a maximem possible error of 82 psi. The signal is defeated below 10% of rated power to allow unimpeded plant startup 2-6

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