[Federal Register Volume 59, Number 77 (Thursday, April 21, 1994)]
[Unknown Section]
[Page 0]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 94-9610]


[[Page Unknown]]

[Federal Register: April 21, 1994]


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NUCLEAR REGULATORY COMMISSION

[Docket No. 50-341]

 

Detroit Edison Co., FERMI 2; Environmental Assessment and Finding 
of No Significant Impact

    The U.S. Nuclear Regulatory Commission (the Commission) is 
considering issuance of an exemption from the requirements of 10 CFR 
part 50, appendix J, III.C, related to Type C local leak rate testing 
of containment isolation valves, to Detroit Edison Company (the 
licensee), for operation of the Fermi 2 Plant, located in Monroe 
County, Michigan.

Environmental Assessment

Identification of Proposed Action

    The proposed action would grant an exemption from the requirements 
of Appendix J. Paragraph III.C. of 10 CFR part 50 and approve 
alternative local leak rate testing of the containment isolation valves 
(CIVs) in the low pressure coolant injection (LPCI) lines of the 
residual heat removal (RHR) system. These lines are 24-inch injection 
lines whose primary containment penetrations are designated as X-13A 
and X-13B. Each contains an outboard-of-containment motor-operated gate 
valve in series with an inboard-of-containment check valve having a 1-
inch bypass line which contains a normally locked closed solenoid 
operated globe valve. The gate and globe valves are remotely operated 
from the control room. Under the provisions of General Design Criterion 
(GDC) 55 of Appendix A of 10 CFR part 50, these valves would be 
required to be designed in accordance with one of the listed 
configurations and designated as CIVs unless it can be demonstrated 
that the containment isolation provisions are acceptable on some other 
defined basis. The outside containment configuration contains a remote 
manual valve rather than an automatic or locked closed valve, but is 
acceptable because LPCI is required to operate for core cooling during 
an accident; therefore, no automatic containment isolation signal is 
used. The inboard valve configuration meets the explicit requirements 
of GDC 55.
    In a letter dated May 24, 1993, the licensee provided justification 
to consider differing from the explicit requirements of GDC 55 in 
accordance with guidance contained in the staff's Standard Review Plan 
(SRP), NUREG-0800, section 6.2.4 ``Containment Isolation System.'' 
Subsection II.6.e allows only a single CIV, outside containment, if the 
system is closed outside containment and certain other provisions are 
met. On this basis, the licensee proposed that the check valves and 1-
inch solenoid operated globe valves inside containment no longer be 
considered as containment isolation valves and; therefore, no longer 
subject to the requirements of Paragraph III.C of Appendix J to 10 CFR 
part 50. These valves would still be considered reactor coolant 
pressure isolation valves and subject to those leak testing provisions 
which are identified in Technical Specification 4.4.3.2.2.
    The licensee also requested an exemption from Paragraph III.C.2 
which requires that CIVs subject to Type C tests, unless pressurized 
with fluid from a seal system, shall be pressurized with air or 
nitrogen to a pressure of Pa, the calculated peak containment internal 
pressure during a design basis accident. The licensee proposed an 
alternative test to measure the external leakage of the CIVs (the motor 
operated gate valves outside containment) using water as a test medium 
with a limited allowable leakage.

The Need for the Proposed Action

    The proposed exemption is needed because compliance to Paragraph 
III.C.2 of 10 CFR part 50, Appendix J, would result in extended outage 
time and additional personnel radiation exposure while testing the CIVs 
described above, without additional safety benefit. The licensee's 
basis for proposing the alternate testing for the CIVs was 
demonstration of the existence of a water seal between the inboard and 
outboard containment valve configurations of these penetrations that 
would be maintained for at least 30 days following an accident, the 
consideration of the RHR system as a closed system outside of 
containment, and operation of the RHR pumps which assures that through-
seat leakage would be leakage in towards containment despite any single 
active failure. With through seat leakage not a concern, the only other 
containment isolation valve leakage of concern would be external to the 
valve (such as a stem or bonnet leak). The licensee's proposed 
alternative testing provides an acceptable basis to resolve this 
concern. Therefore, adequate containment integrity is demonstrated and 
the underlying purpose of the regulations is achieved.

Environmental Impact of the Proposed Action

    The proposed exemption would allow the substitution of an 
alternative testing for the required Type C leak rate testing for 
containment isolation valves. The staff has determined that the 
alternative testing would provide an acceptable basis for demonstrating 
containment integrity. The related proposed alternative basis for 
compliance with the provisions of GDC 55 does not require an exemption, 
but does formulate, in part, the basis for approval of the proposed 
exemption. The staff has determined that the justification provided by 
the licensee adequately meets the SRP section 6.2.4, II.6.e review 
criteria for an alternative basis to meet the requirements of GDC 55. 
The alternative design requirements provide adequate assurance of 
containment integrity. Although the inboard containment valve 
configurations will no longer be tested to the Type C integrated leak 
rate test criteria, these valves will continue to be leak tested to 
demonstrate their reactor coolant pressure isolation valve function. 
Therefore, post-accident radiological releases are not expected to 
exceed previously determined values as a result of the proposed action. 
Further, the exemption is not expected to have an impact on plant 
radiological effluent releases. The proposed action does have the 
potential to reduce occupational exposure by reducing the amount of 
time personnel are required to spend in a radiologically restricted 
area.
    With regard to potential non-radiological impacts, the proposed 
action and related change to the Technical Specifications involve a 
change in the surveillance requirements and will not affect non-
radiological plant effluents nor does it have any other environmental 
impact. Therefore, the Commission concludes that there are no 
significant non-radiological environmental impacts associated with the 
proposed exemption.

Alternative to the Proposed Action

    Since the Commission concluded that there are not significant 
environmental effects that would result from the proposed action, any 
alternatives with equal or greater environmental impacts need not be 
evaluated.
    The principal alternative would be to deny the requested exemption 
and amendment. This would not reduce environmental impacts of plant 
operation and would result in reduced operational flexibility and 
greater occupational exposure to plant personnel.

Alternative Use of Resources

    This action does not involve the use of resources not previously 
considered in connection with the Commission's Final Environmental 
Statement, dated August 1981, for Fermi 2.

Agencies and Persons Consulted

    The staff consulted with the State of Michigan regarding the 
environmental impact of the proposed action. The State had no comments.

Finding of No Significant Impact

    The Commission has determined not to prepare an environmental 
impact statement for the proposed exemption.
    Based upon the foregoing environmental assessment, the staff 
concludes that the proposed action will not have a significant effect 
on the quality of the human environment.
    For further details with respect to this proposed action, see the 
licensee's application and request for exemption dated May 24, 1993. 
This document is available for public inspection at the Commission's 
Public Document Room, 2120 L Street, NW., Washington, DC 20555, and at 
the local public document room located at the Monroe County Library 
System, 3700 South Custer Road, Monroe, Michigan 48161.

    Dated at Rockville, Maryland, this 14th day of April, 1994.

    For the Nuclear Regulatory Commission.
Ledyard B. Marsh,
Director, Project Directorate III-1, Division of Reactor Projects--III/
IV, Office of Nuclear Reactor Regulation.
[FR Doc. 94-9610 Filed 4-20-94; 8:45 am]
BILLING CODE 7590-01-M