[Federal Register Volume 59, Number 84 (Tuesday, May 3, 1994)]
[Unknown Section]
[Page 0]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: 94-10530]


[[Page Unknown]]

[Federal Register: May 3, 1994]


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NUCLEAR REGULATORY COMMISSION
[Docket No. 50-341]

 

Detroit Edison Co. (Fermi 2); Exemption

I

    Detroit Edison Company (the licensee) is the holder of Facility 
Operating License No. NPF-43 which authorizes operation of the Fermi 2 
Nuclear Plant at steady-state reactor power levels not in excess of 
3430 megawatts thermal. The Fermi 2 facility is a boiling water reactor 
located at the licensee's site in Monroe County, Michigan. The license 
provides, among other things, that it is subject to all rules, 
regulations, and Orders of the Nuclear Regulatory Commission (the 
Commission) now or hereafter in effect.

II

    Paragraph III.C of appendix J to 10 CFR part 50 requires, in part, 
that valves, unless pressurized with fluid (e.g., water, nitrogen) from 
a seal system, shall be tested by pressurizing with air or nitrogen at 
a test pressure of Pa (56.5 psig), the calculated peak containment 
internal pressure as a result of the design basis accident. Further, 
the combined leakage rate of all penetrations and valves subject to 
Type B and C testing shall be less than 0.60 La (La is the maximum 
allowable leakage rate at Pa). Leakage from containment isolation 
valves that are sealed with fluid from a seal system may be excluded, 
provided the leakage rates for these valves do not exceed the Technical 
Specification leakage requirements and the seal system fluid inventory 
is sufficient to ensure the sealing function for 30 days following an 
accident at a pressure of 1.10 Pa.
    Pursuant to 10 CFR 50.12(a), the NRC may grant exemptions from the 
requirements of the regulations (1) which are authorized by law, will 
not present an undue risk to the public health and safety, and are 
consistent with the common defense and security; and (2) where special 
circumstances are present.

III

    By letter dated May 24, 1993, the licensee requested an exemption 
from the requirements of 10 CFR part 50, appendix J, III.C for 
performing Type C integrated leak rate tests of the containment 
isolation valves in the Low pressure coolant injection (LPCI) lines of 
the residual heat removal (RHR) system. The valves in question are the 
loop A and B LPCI isolation valves, which are motor-operated gate 
valves, outboard of containment and which are operable from the control 
room and are designated as E11-F015 A and B. The licensee proposed, as 
alternative testing, an external leakage test, with water as the test 
medium, at a pressure of 1045 psig with an allowable leakage value of 5 
milliliters per minute (ml/min). The license also provided 
justification in its May 24, 1993, letter to reclassify the inboard 
containment LPCI valve configuration (which consists of a reverse flow 
swing check valve with a 1-inch, locked closed, solenoid-operated 
bypass valve) as other than containment isolation valves and thus no 
longer subject to Type C testing.
    The staff evaluated the licensee's proposal for reclassification of 
the inboard containment valves and concluded that the licensee's 
proposal met the guidance in the Standard Review Plan (SRP), NUREG-
0800, Section 6.2.4 for differing from the explicit requirements of 
General Design Criterion 55 in 10 CFR part 50, appendix A for 
containment isolation valves (CIVs). Subsection II.6.e allows only a 
single CIV outside containment, if the system is closed outside 
containment and certain other criteria are met. Details concerning the 
staff's review are contained in the staff's safety evaluation dated 
April 22, 1994.
    Two aspects of the RHR system form the basis for the proposed 
exemption. It is a closed system outside of containment, and the 
containment penetrations will be water sealed during a loss of coolant 
accident (LOCA). The licensee's analyses showed that a water seal. 
pressurized to greater than 1.1 Pa (62.15 psig), would exist outboard 
of the LPCI CIV for at least 30 days following the design basis LOCA 
despite the most limiting single active failure. However, if one or 
both of the LPCI CIVs is shut, that water seal might not prevent 
external valve leakage (valve stem or bonnet leakage). The licensee 
also showed that a water seal, would exist inboard of the LPCI CIVs 
following a design basis LOCA. The licensee's analyses demonstrated 
that the volume of the water seal is sufficient to last for greater 
than 30 days assuming the leakage limit proposed in their alternative 
testing acceptance criteria. The licensee's analyses also showed that 
through seat leakage of the LPCI CIVs would be in toward containment 
and would not deplete the water seal.
    Although the external leakage water seal would prevent atmosphere 
from leading out of containment, it does not satisfy the requirements 
for a water seal contained in Appendix J. Appendix J allows water 
sealed valves to be excepted from the normal Type C testing with air, 
but it requires that the water seal be pressurized to at least 1.1 Pa 
during an accident. The water leg inboard of the LPCI CIVs does not 
meet this requirement. Nevertheless, the staff has determined that the 
licensee's analyses of the water seals provide sufficient assurance 
that containment atmosphere leakage out of containment will be 
prevented during an accident to justify granting the requested 
exemption from Type C testing of the LPCI CIVs with air as the test 
medium. The licensee's alternative test will measure external valve 
leakage with a limit of 5 ml/min using water as the test medium at a 
pressure of 1045 psig.

IV

    Accordingly, the Commission concluded that the licensee's proposed 
alternative testing plan and analyses provide sufficient assurance that 
the containment atmosphere would not leak out of containment through 
the LPCI CIVs during a design basis accident and that containment 
integrity will be maintained by granting the proposed exemption.
    The special circumstances for granting this exemption pursuant to 
10 CFR 50.12 have also been identified. As stated in part in 10 CFR 
50.12(a)(2)(ii), special circumstances are present when application of 
the regulation in the particular circumstances is not necessary to 
achieve the underlying purpose of the rule. The purpose of Section 
III.C of Appendix J is to measure containment isolation valve leakage 
rates. This leakage, when summed with the allowable Type A and Type B 
leakage is limited to a value which ensures overall containment 
integrity in preventing the uncontrolled release of radioactivity to 
the environment. The licensee has demonstrated through analyses and by 
proposing alternative testing criteria, that containment integrity will 
be maintained. Consequently, the Commission concludes that the special 
circumstances of 10 CFR 50.12 (a)(2)(ii) exist in that application of 
the regulation in these particular circumstances is not necessary to 
achieve the underlying purpose of the rule.

V

    Accordingly, the Commission has determined that, pursuant to CFR 
50.12 this exemption as described in Section III above is authorized by 
law, will not present an undue risk to the public health and safety, 
and is consistent with the common defense and security. The Commission 
further determines that special circumstances as provided in 10 CFR 
50.12(a)(2)(ii) are present justifying the exemption.
    Therefore, the Commission hereby grants an exemption as described 
in Section III above the requirements in 10 CFR 50.12, Appendix J, 
III.C. for performing Type C containment integrated leak rate tests of 
the CIVs in the LPCI lines of the RHR system and approves the 
licensee's alternative testing plan.
    Pursuant to CFR 51.32, the Commission has determined that the 
granting of this exemption will have no significant impact on the 
environment (59 FR 19028).
    This exemption is effective upon issuance.

    Dated at Rockville, Maryland, this 22nd day of April 1994.

    For the Nuclear Regulatory Commission.
Elinor G. Adensam,
Acting Director; Division of Reactor Projects--III/IV; Office of 
Nuclear Reactor Regulation.
[FR Doc. 94-10530 Filed 5-2-94; 8:45 am]
BILLING CODE 7590-01-M