ML20212N446

Safety Evaluation Supporting Amend 95 to License NPF-3

ML20212N446
Person / Time
Site:	Davis Besse
Issue date:	08/20/1986
From:	Office of Nuclear Reactor Regulation
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Shared Package
ML20212N443	List: ML20212N439 ML20212N446
References
NUDOCS 8608280233
Download: ML20212N446 (5)

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/[* -) Tif. 'n UNITED STATES NUCLEAR REGULATORY COMMISSION g%Q W ASHINGTON, D. C. 20555 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT N0. 95 TO FACILITY OPERATING LICENSE NO. NPF-3 TOLED0 EDIS0N COMPANY AND THE CLEVELAND ELECTRIC ILLUMINATING COMPANY DAVIS-BESSE NUCLEAR POWER STATION, UNIT NO. 1 DOCKET NO. 50-346 INTRODUCTION By letter dated June 6, 1986, Toledo Edison Company (TED or the licensee) requested amendment to the Technical. Specifications (TSs) appended to Facility Operating License No. NPF-3 for the Davis-Besse Nuclear Power Station, Unit No. 1.

The proposed amendment would modify Surveilla ce Requirement 4.4.10.1.b to extend the surveillance period of inspection and operability testing of the Reactor Vessel Internals Vent Valves (RVVVs) to coincide with the next reactor vessel head removal but not later than the refueling outage following Cycle E operation which is expected to occur no later than March 1988.

Following review of the initial application, a telephone conference was held with the NRC staff and representatives of Toledo Edison Company and Babcock and Wilcox Company (B&W) on July 24, 1986 to obtain clarification and to request additional information. The licensee submitted additional information regarding mechanical design, materials of construction, comparison of Three MileIslandNuclearStation,UnitNo.1(TMI-1)andDavis-Bessedesigns,and chemistry of the Reactor Coolant System (RCS) by letter dated July 30,'1986 (No. 1292).

BACKGROUND In 1975, the NRC revised 10 CFR 50.55a to require an " Inservice Inspection" of various safety related components, including pumps and valves, to be performed in accordance with the ASME Boiler and Pressure Vessel Code,Section XI, "to the extent practical within the limitations of design, geometry, and materials of construction." The Davis-Besse TS Section 4.4.10.1.b requires that the RVVVs be demonstrated operable at least once per 18 months with a provision that an extension of 25% (4.5 months) may be granted for the 18-month period.

The RVVVs were last tested in accordance with the surveillance requirement on October 12, 1984 during the refueling outage for Cycle 5.

The refueling outage was completed in December 1984 and the facility was restarted in early January 1985. The core design for Cycle 5 allowed for 390 effective 86oe280233 e60820 PDR ADOCK 05000346 P

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. full power days and, therefore, even with a capacity factor as low as 60%, the surveillance requirement could be met when the reactor vessel head was removed to refuel. However, in early June 1985, a loss of feedwater event occurred and the facility has not operated since then.

In order to meet the TS requirement, valve operability wnuld need to be demonstrated not later than August 27, 1986, unless the requested extension is granted.

The Davis-Besse facility is expected to be restarted about November 1986.

Approximately 12-13 months additional operating life remains in the Cycle 5 core. Therefore, refueling would not be expected to be required until about March 1988. Thus, the interval from the last surveillance test could be as long as about 42 months.

Toledo Edison provided a justification for the requested extension by reference to paragraph IWA-2400(c) of Section XI, ASME Boiler and Pressure Vessel Code, addendum of Summer 1978 which states:

...For power units that are out of service continuously for 6 months or more, the inspection interval during which the outage occurred may be extended for a period equivalent to the outage."

Toledo Edison noted that while the plant has been in the prolonged outage, the internals vent valves have not been exposed to the turbulent flow environment that exists when the plant is operating. The flow through the core while cooling is provided by the Decay Heat Removal System is less than 1% of the flow with four reactor coolant pumps operating. Under these conditions, the internals vent valves do not experience the large differential pressure that exists when the plant is operating. Without this large differential pressure, the vent valves are being held closed only by gravity in a quiescent environment.

At TMI-1, where the RVVVs were held in a similar environment for over three years, no adverse effects were found upon inspection and exercising of the RVVVs. Becaure of the similarities in valve design, construction and water chemistry between TN 1 and Davis-Besse, Toledo Edison Company asserts that the favorable performance of the RVVVs at TMI-1 would. indicate the acceptability of extending the surveillance interval at Davis-Besse for a comparable period.

A review of the results obtained from RVVV surveillance tests performed at eight operating B&W plants was performed by B&W. The data repnrted about 400 inspections and exercises over the past ten years without a failure. These data demonstrate that the RVVVs have exhibited a high degree of reliability with no observable degradation in valve operability with reactor age.

DISCUSSION AND EVALUATION The intervals vent valves are installed in the core support shield to prevent a pressure unbalance which might delay or interfere with emergency core cooling following a postulated inlet pipe rupture. The arrangement consists of four 14-inch inside diameter vent valve assemblies installed in the cylindricci wall of the internal core support shield. The internals vent valves provide a direct path to vent steam in the upper plenum through the break following a postulated cold-leg rupture. The vent valves are required i

because the arrangement of the RCS could delay.the venting of steam generated in the core after the system is depressurized, if significant quantities of coolant remair. in the reactor inlet piping at the end of the blowdown period.

Without venting of the steam, the pressure in and above the core region could be g,reater than the pressure in the reactor vessel inlet annulus where emergency coolant is injected. This pressure differential could retard flow into the core. The vent valves provide a flowpath from the region above the core directly to the pipe rupture location. This flowpath allows the pressures to equalize and permits emergency coolant water to reflood the core rapidly.

In its evaluation of the licensee's request to extend the RVVV surveillance interval, the NRC staff considered the results from the approximately 400 inspections and exercises which have been performed at all operating B&W facilities.

The typical interval t'etween RVVV inspection and exercise was 12 to 18 months with a maximum test interval (with the exception of TMI-1) of about two years.

In the case of TMI-1, the longest interval between tests was 37 months. These inspections indicate a high degree of reliability for the RVVVs sir.ce no degradation in valve operability has been observed.

The NRC staff also evaluated infomation pertaining to the RCS chemistry, material compatability and corrosion resistan.ce, and the reactor coolant environment. The chemistry of the RCS water is controlled to minimize corrosion and material activations and to assure the reliability of reactor and steam generator equipment. The licensee reported that the RCS water has not been out of specification at any time since the last surveillance test of the RVVVs. Comparison of the critical elements of water chemistry such as the concentrations of boron, fluoride, oxygen', and chloride, as well as the pH for the Davis-Besse and TMI-1 plants, indicates that the water environments in both plants are similar. The RVVVs at TMI-1 were not exercised for about 37 months -- slightly less than the maximum interval to be expected in the case of the RVVVs at Davis-Besse. The staff believes that the favorable TMI-1 results are indicative of what would be expected at Davis-Besse.

Corrosion, which could have an effect on the operation of RVVVs, has been considered. The parts vulnerable to corrosion are the shaft, bushing, and the body. These components are constructed of Type 431 martensitic stainless steel, Stellite No. 6, and Type 304 austenitic stainless steel respectively.

Available data for the RCS hot operating conditions indicate that the general corrosion rates of these materials, as reported by the licensee, are in the range of 0.05 mils / year or less. This information has been verified independently by the NRC staff in the scientific literature (Reference 3).

Since the accumulation of the corrosion deposit is about three times the corrosion rate, the expected thickness of the deposited material would be 0.15 mils per year. The minimum cold clearance gap dimensions vary from 3 to 60 mils, therefore the gap would not close and hinder the operation of the valve during the period of time until the next test of the RVVVs.

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1 ENVIRONMENTAL CONSIDERATION This amendment involves changes in the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20.

We have determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that this amendment involves no significant hazards consideration and there has been no public comment on such finding. Accordingly, this amendment meets the eligibility criteria for categorical exclusion set forthin10CFR51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of this amendment.

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

(1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (2) such activities will be conducted in compliance with the Commission's regulations and the issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public.

Dated: August 20, 1986 Principal Contributors:

R. Lipinski G. Hammer

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References:

1.

Letter from Joe Williams, dr., Toledo Edison, to John F. Stolz, NRC, dated June 6, 1986.

2.

Letter from Jae Williams, Jr., Toledo Edison, to John F. Stolz, 1

NRC, dated July 30, 1986.

s 3.

Uhlig, Herbert, H., " Corrosion and Corrosion Control," John Wiley and Sons Inc., 2nd Edition, 1971.

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