Protecting People and the EnvironmentUNITED STATES NUCLEAR REGULATORY COMMISSION
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
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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:
ø Current confidence in valve performance and leak tightness
ø Conservatisms in valve leak rate evaluations
ø System design considerations and their mitigating effects
ø The low probability of a LOCA with fuel damage and a large fission
product release to the containment building
ø 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
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IN 91-56
September 19, 1991
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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
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Attachment 2
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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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