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Information Notice No. 91-56: Potential Radioactive Leakage to Tank Vented to Atmosphere
UNITED STATES NUCLEAR REGULATORY COMMISSION OFFICE OF NUCLEAR REACTOR REGULATION WASHINGTON, D.C. 20555 September 19, 1991 NRC INFORMATION NOTICE 91-56: POTENTIAL RADIOACTIVE LEAKAGE TO TANK VENTED TO ATMOSPHERE Addressees All holders of operating licenses or construction permits for nuclear power reactors. Purpose The U.S. Nuclear Regulatory Commission (NRC) is issuing this information notice to alert addressees to potential problems resulting from the leakage of isolation valves in emergency core cooling system (ECCS) recirculation lines to the safety injection water storage tank, which may be vented to atmosphere. It is expected that recipients will review the information for applicability to their facilities and consider actions, as appropriate, to avoid similar problems. However, suggestions contained in this information notice are not NRC requirements; therefore, no specific action or written response is required. Description of Circumstances In September 1990, the Consumers Power Company, the licensee for the Palisades Plant, performed a test to measure the leakage through the suction valves from the safety injection and refueling water tank (SIRWT; see Figure 1). The licensee concluded that leakage through the suction valves was not a concern, but that radioactivity could be released through two other leakage paths: through the ECCS recirculation (minimum flow) header to the SIRWT and through the ECCS test header to the SIRWT. The concern arises because the SIRWT is vented to the atmosphere. During the initial phase of recovery from a loss-of-coolant accident (LOCA), the ECCS recirculation header collects water from the safety injection pump minimum flow lines for return to the SIRWT through the minimum flow recirculation control valves (See Figure 1). During the later recirculation phase of recovery, this header is isolated from the SIRWT by these two independent control valves, which are in series. The test header collects water from the shutdown cooling heat exchangers during ECCS surveillance testing and during SIRWT mixing operations for return to the SIRWT through a manual test valve (See Figure 1). During reactor operation, the manual test valve is locked closed, isolating the test header from the SIRWT. The leak tightness of these three isolation valves has never been previously verified at the Palisades Plant. The licensee's previous calculations of 9109130099 . IN 91-56 September 19, 1991 Page 2 of 3 control room and offsite doses from a maximum hypothetical accident did not include such valve leakage. Preliminary estimates indicate that a leak rate of 0.1 gpm, together with other assumed sources, could cause a dose to control room personnel exceeding the limits in General Design Criterion (GDC) 19, "Control Room," 10 CFR Part 50, Appendix A. These limits include a dose of 5 rem to the whole body or an equivalent dose to any part of the body. A leak rate of 1 gpm, together with other assumed sources, could cause a dose to people at the site boundary equal to the limits in 10 CFR Part 100. These limits include a dose of 25 rem to the whole body or 300 rem to the thyroid from iodine exposure. The licensee does not know the rate of leakage and thus can not determine if prior analyses underestimate the possible accident consequences. The licensee determined that excessive valve leakage could cause consequences to be larger than those from previous assessments and could cause the radiological consequences at the site boundary and to control room operators to exceed the regulatory limits. Therefore, the licensee's safety review committee reviewed this issue and determined it to be an unreviewed safety question. The licensee has submitted a license amendment to the NRC for review. The licensee's justification for continued operation addresses the following factors: � Current confidence in valve performance and leak tightness � Conservatisms in valve leak rate evaluations � System design considerations and their mitigating effects � The low probability of a LOCA with fuel damage and a large fission product release to the containment building � Conservatisms in the present maximum hypothetical accident The licensee is resolving this concern by eliminating conservative assumptions in leak rate evaluations and current maximum hypothetical accident analyses, implementing procedures to lessen the consequences of valve leakage, evaluating a modification to the manual valve, and including leak rate testing of the minimum flow recirculation control valves as part of the inservice testing program, following modifications scheduled for February 1992. Discussion Plants other than Palisades may be vulnerable to an unmonitored release from the safety injection water storage tank during and following a postulated design basis accident. On August 19, 1991, the Omaha Public Power District, the licensee for the Fort Calhoun Nuclear Power Plant, reported a similar concern with respect to the SIRWT suction valves. In addition, References 1 and 2 provide discussion of somewhat similar situations. The NRC staff has reviewed the system design characteristics and the inservice testing programs . IN 91-56 September 19, 1991 Page 3 of 3 of several utilities and has determined that similar conditions, with comparable consequences, may exist at other plant sites. In particular, valves with similar functions are not identified as Category A valves in inservice testing programs. Inservice testing requirements are specified in 10 CFR 50.55a(g), which references the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section XI. Subsection IWV-2200, "Categories of Valves," stipulates that Category A valves are those valves with functions in which the closed valve seat leakage is limited to a specific amount. The leak test requirements for Category A valves are specified in Section XI, IWV-3420, "Valve Leak Rate Test." This information notice requires no specific action or written response. If you have any questions about the information in this notice, please contact one of the technical contacts listed below or the appropriate Office of Nuclear Reactor Regulation project manager. Charles E. Rossi, Director Division of Operational Events Assessment Office of Nuclear Reactor Regulation Technical contacts: Patricia Campbell, NRR (301) 492-1311 Brian E. Holian, NRR (301) 492-1344 Vern Hodge, NRR (301) 492-1861 Attachments: 1. Figure 1. Partial ECCS Piping Diagram for the Palisades Plant 2. List of References 3. List of Recently Issued NRC Information Notices . Attachment 2 IN 91-56 September 19, 1991 Page 1 of 1 LIST OF REFERENCES 1. NRC Information Notice 85-94, "Potential for Loss of Minimum Flow Paths Leading to ECCS Pump Damage During a LOCA," December 13, 1985 2. NRC Information Notice 86-38, "Deficient Operator Actions Following Dual Function Valve Failures," May 20, 1986 .
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