ML20202D487

Forwards Nusco 151, Haddam Neck Plant Non-LOCA Transient Analysis. Reanalysis Submitted to Demonstrate That Northeast Util Svcs Co Has Expertise & Analytical Tools to Perform Transient Analyses to Support Core Reloads.Fee Paid

ML20202D487
Person / Time
Site:	Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png
Issue date:	06/30/1986
From:	Opeka J, Sears C CONNECTICUT YANKEE ATOMIC POWER CO.
To:	Charemagne Grimes Office of Nuclear Reactor Regulation
Shared Package
ML20202D493	List: ML20202D487 ML20202D507
References
TASK-04-02, TASK-15-02, TASK-15-07, TASK-15-17, TASK-15-2, TASK-15-7, TASK-4-2, TASK-RR B-12133, NUDOCS 8607140144
Download: ML20202D487 (5)

Text

s s CONNECTICUT YANKEE ATOMIC POWER COMPANY B E R L I N, CONNECTICUT P o. BOX 270 HARTFORD CONNECTICUT 06141-0270 TELEPHONE 203-665-5000 June 30,1986 Docket No. 50-213 B12133 Office of Nuclear Reactor Regulation Attn: Mr. Christopher I. Grimes, Director Integrated Safety Assessment Project Directorate Division of PWR Licensing - B U. S. Nuclear Regulatory Commission Washington, D. C. 20555 Gentlemen:

Haddam Neck Plant Reanalysis of Non-LOCA Design Basis Accidents The attached report describes the results of the reanalysis performed by Northeast Utilities Service Company (NUSCO), on behalf of the Connecticut Yankee Atomic Power Company (CYAPCO), of the non-LOCA design basis accidents for the Haddam Neck Plant. This reanalysis is being submitted for review for the following purposes:

1) To demonstrate that NUSCO has the expertise and analytical tools necessary to perform transient analyses to support core reloads and other plant changes.

2) To resolve several open issues related to the existing design basis analyses identified during the course of the Systematic Evaluation Program (SEP) review.

3) To resolve several open items related to the existing design basis analyses identified by the Connecticut Yankee Plant Design Change Task Group during the design change review mandated by the Order Modifying License issued to CYAPCO on December 13, 1984.(l)

Each of these objectives is discussed in more detail below.

The reanalysis has been performed to be consistent with the current plant design and configuration. It should be noted that while there is no requirement to meet the more recent guidelines of the Standard Review Plan (SRP), the reanalysis assumptions have been expanded to the extent practical. The analyses were performed to show conformance with a larger percentage of the SRP criteria and assumptions than the current design basis analysis. '

(1) 3. M. Taylor letter to W. G. Counsil, dated December 13,1984, Order Modifying License and Notice of Violation and Proposed Imposition of Civil /

Penalty, Docket No. 50-213, EA-84-Il5. v q3 wmmw e o P

e o.

Reload Analysis Capability NUSCO has established the objective of developing reload analyses for the Haddam Neck Plant, utilizing in-house resources. There are several factors which led to this initiative. For example, because of a number of unique design features (e.g., stainless steel clad fuel supplied by a vendor other than the NSSS supplier, early vintage plant,' etc.), the " generic" bounding analyses performed by the various vendors are often not applicable to Haddam Neck, and plant-specific "one of a kind" analyses for the life of the plant could result in significant costs.

Additionally, since these are plant-specific analyses, vendors are not always capable of performing this work on the required schedule and at reasonable cost.

To that end CYAPCO has previously submitted reports for non-LOCA and LOCA analyses (2d). This report represents the design basis analyses for the non-LOCA transients for Haddam Neck Plant based upon the generic methodology discussed in the reports transmitted by W. G. Counsil letter to D. M.

Crutchfield, dated August 1,1984.

SEP Open Issues As part of Systematic Evaluation. Program review (4), several questions were raised about the current design basis analysis. The questions raised were associated with the following transients:

o Uncontrolled Rod Withdrawal (IPSAR Section 4.15)

The review of SEP Topic IV-2, " Reactivity Control Systems, including Functional Design and Protection Against Single Failures," identified a number of single failures which could result in a reactivity addition through accidental withdrawal of control rods. The analysis of the uncontrolled rod

- withdrawal event was performed and is discussed in the attached report to address possible reactivity insertion events resulting from the withdrawal of any individual bank or the withdrawal of the two sequential control banks in controlled overlap.

o Excess Feedwater Flow (IPSAR Section 4.34)

The review of - SEP Topic XV-2, " Decrease in Feedwater Temperature, Increase in Feedwater Flow, Increase in Steam Flow, and Inadvertent (2) W. G. Counsil letter to J. R. Miller and D. M. Crutchfield, dated July 30, 1984 and enclosures entitled "NUSCO Thermal Hydraulic Model Qualification Volume I (RETRAN)," NUSCO 140-1, dated August 1,1984 and "NUSCO Thermal Hydraulic Model Qualification Volume II (VIPRE)",

NUSCO 140-2, dated August 1,1984.

(3) W. G. Counsil letter to W. A. Paulson, dated August 23, 1984 and enclosure entitled " Calculative Methods for the Northeast Utilities Small Break LOCA ECCS Evaluation Model," dated July,1984, Docket No. 50-213.

(4) Integrated Plant Safety Assessment Report, Systematic Evaluation Program, Haddam Neck Plant, NUREG-0826, dated June,1983.

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Opening of a Steam Generator Relief or Safety Valves," identified that for j'

the excess feedwater flow analysis in the FDSA, operator action was required in a very short time frame to prevent steam generator overfill.

CYAPCO committed to provide a reanalysis of this event to address the

SEP concern. The attached reanalysis confirms the FDSA results. The results have shown that for the limiting case (4-loop operation,40% power, I feedwater regulating valve fails open) a very short time frame is

available for operator action after the high-level alarms to trip the main feedwater pumps to prevent steam generator overfill.

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o Steam Generator Tube Rupture (IPSAR Section 4.39)

CYAPCO provided an analysis of a steam generator tube rupture event during the review of SEP Topic XV-17, Steam Generator Tube Rupture.

During a review of this analysis, the Staff identified several concerns regarding the analytical assumptions used, and thus could not complete its evaluation of the radiological consequences. CYAPCO committed to provide a reanalysis of the steam generator tube rupture event to address the Staff's concerns. As such, a reanalysis of this event is included in the attached report.

o Reactor Coolant Pump Rotor Seizure and Shaf t Break (IPSAR Sec. 4.36)

The safety objective of SEP Topic XV-7, " Loss of Forced Coolant Flow,

Reactor Coolant Pump Rotor Seizure, and Reactor Coolant Pump Shaf t Break," had not previously been reviewed for the Haddam Neck Plant, as the rotor seizure / shaft break event was not in the original design basis. To address the safety issue of this SEP topic, CYAPCO committed to perform an analysis of the rotor seizure and shaft break events. Loss of forced coolant flow had previously been analyzed and found acceptable by the i Staff during the SEP review, however, this event has also been reanalyzed and is discussed in the attached report. The single pump transients, such as i reactor coolant pump rotor seizure and shaft break, have been analyzed with credit for the low flow trip. The combination of a single pump failure and a failure of the single low flow instrum(ent in the same loop is beyond the design basis for the Haddam Neck Plant. 5)

CYPDCTG Open Issues As a result of the reactor cavity pool seal failure, CYAPCO established the Connecticut Yankee Plant Design Change Task Group (CYPDCTG) to review all plant changes approved during the period of January 1,1979 through December 31, 1984. The Task Group identified the following deficiencies related to the I

(5) This is not to be confused with the loss of coolant flow in one coolant loop which conservatively credits operation of the single low flow transmitter.

i The reanalysis of the loss of coolant flow transient has been performed to be consistent with the FD5A in terms of reactor trip protection.

Additional protection for the single flow instrument is provided by the

undervoltage trip and the pump circuit breaker trip.

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current design basis analyses (6)

I o Deficiency No. 20: Questionable Design Basis Analysis of the AFW system (Loss of Feedwater Analysis) o Deficiency No. 37: Incomplete review of acceptance limits for the RTDs (Uncontrolled Rod Withdrawal Analysis)

The analyses provided in the attached report resolve these deficiencies by upgrading the design basis analysis. ,

Findings From the Reanalysis Effort The reanalysis of the design basis analyses has lead to some changes in the requirements of some systems previously credited in the design basis analysis.

These changes are as follows:

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1) The PORV setpoint will be raised to a value that is greater than the reactor trip setpoint. This assures that a timely reactor trip will occur in the loss of load transient.

2) The analysis of the isolated loop startup transient has shown that credit is not needed for the valve interlock control on reactor coolant pump start

circuit.

3) The analysis of the dropped rod transient has shown that credit is not needed for the rod block or turbine runback features.

The new analyses have also identified two areas requiring further study. These l are as follows:

i 1) For reactivity transients, such as uncontrolled rod withdrawal, boron dilution and control rod ejection, initiated at low or intermediate power l levels, credit is taken for operator action to modify the overpower trip and monitor control rod position to ensure that adequate protection is maintained. Studies will be initiated to determine the feasiHlity and i desirability of providing automatic protection for these conditiom.

2) For the steam generator tube rupture, credit is taken for closure of the loop isolation valves. Studies have been initiated to determine the feasibility and desirability of increasing the steam dump capacity and to make the use of the loop isolation valves optional for a steam generator tube rupture.

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l (6) 3. F. Opeka letter to Dr. T. E. Murley, dated November 6,1985,

Subject:

l "Haddam Neck Plant Response to December 13, 1984 Order Modifying

License."

-S-CYAPCO intends to include these items in the Integrated Safety Assessment program (ISAP) for the Haddam Neck Plant. CYAPCO also intends to incorporate the analyses in this report into the upgrade of the Haddam Neck FSAR and to use the analyses in the evaluation of the Cycle 15 reload for Haddam Neck. To that end, we request a timely Staff review of the attached report, and we remain available to discuss it with the Staff at your convenience.

Once the Staff approves the attached reanalysis of the non-LOCA design basis events, it will represent the licensing basis for the Haddam Neck Plant. In the interim, we will continue to utilize the existing licensing basis.

In accordance with the provisions of 10CFR170.21, enclosed herewith is the required application fee of $150.00.

Very truly yours, CONNECTICUT YANKEE ATOMIC POWER COMPANY b..

J. F. Opeka '

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