Information Notice No. 91-56: Potential Radioactive Leakage to Tank Vented to Atmosphere

                                UNITED STATES
                           WASHINGTON, D.C.  20555

                             September 19, 1991

                               TO ATMOSPHERE


All holders of operating licenses or construction permits for nuclear power 


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:

     �    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.


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 

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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

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
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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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