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Resolution of Generic Safety Issues: Issue 53: Consequences of a Postulated Flow Blockage Incident in a BWR (Rev. 1) ( NUREG-0933, Main Report with Supplements 1–35 )


Historical Background

In response to a 1967 ACRS concern relative to the potential of melting and subsequent disintegration of a portion of a fuel assembly due to inlet orifice flow blockage, GE submitted NEDO-10174380 in May 1970. As a result of a staff review of another topical report on new fuel design, the applicability of NEDO-10174380 to the then new (8 x 8) fuel was questioned. In late 1977, GE submitted NEDO-10174, Revision 1.288 This revision is still awaiting staff review.381

Safety Significance

In a BWR, each fuel bundle is surrounded by a channel box. (The bundle plus channel box is referred to as a fuel "assembly.") Coolant is metered to each fuel assembly by individual orifices in the fuel support castings. All core flow must come through these orifices. Even the water between the channel boxes (outside of the fuel assemblies) now comes from holes drilled in the fuel assembly nosepieces; it does not come directly from the lower plenum.

The safety concern is that, if an orifice became blocked with the reactor operating at power, its corresponding fuel assembly could be deprived of coolant flow, but would still be producing thermal energy and could be damaged or even become molten. Moreover, if a highly overheated assembly were to suddenly refill with water (from dislodging of the blockage at the orifice), there would be the possibility of a steam explosion with consequent pressure pulses on the primary system.

Possible Solutions

Some mitigating features already exist. The first is the effect of the void reactivity coefficient. Reduced flow and consequent increased voiding will reduce power in the blocked assembly and in the immediately surrounding assemblies. This effect tends to compensate for most partial blockages. Second, even if the orifice is completely closed, the blocked assembly is still open at the top and has holes drilled in the nosepiece down below. Thus, some flow would still be available. (Older fuel designs did not have holes in the nosepiece. However, these older designs also did not have finger springs to seal the lower end of the channel box to the nosepiece. Thus, these older designs have as much or more leakage flow than the new designs.)

NEDO-10174, Revision 1288 contains a summary of the calculations of the core's response to an orifice blockage event assuming these two features. The report came to the following conclusions:

(1) The only mechanism capable of causing a major flow blockage is that induced by a foreign object.

(2) Fragmentation, crudding, or fuel swelling cannot cause major flow blockages.

(3) A fuel assembly is capable of withstanding very severe blockages before losing adequate cooling.

(4) For orifice blockages greater than 98%, fuel and cladding melt are expected to occur. However, this will not result in failure propagation to adjacent assemblies, local high pressure production or offsite doses in excess of a small fraction of 10 CFR 100 guidelines. For this worst-case event, no action is required of the reactor ECCS. However, the reactor must be scrammed by the main steam line radiation monitor.

(5) For orifice blockages between 95% and 98%, clad melting is expected, but fuel melting is not calculated to occur. For this case, the consequences are less severe than in (4) above.

(6) For orifice blockages between 79% and 95%, boiling transition and attendant cladding heatup are calculated to occur. No clad nor fuel melting is calculated. However, cladding failure is not precluded. The off-gas radiation monitor will provide an alarm to the reactor operator if fission product releases are significant.

(7) For orifice blockages less than 79%, nucleate boiling is maintained. Therefore, the fuel and cladding are unaffected.

If these conclusions are accepted, it is not clear what more needs to be done to prevent damage from blockage. PNL suggested (for prioritization purposes) adding more holes to the nosepieces of the fuel assemblies.64 These holes would normally be closed off by internal flapper valves. If the orifice were blocked, these springless valves would open to allow more coolant into the assembly.


Frequency Estimate

Experience has shown that the presence of loose objects in the primary system of reactors is not a rare occurrence. Nevertheless, blockage of an orifice is expected to be rare:

- coolant velocities are relatively low in the lower plenum-- so much so that thermal stratification might occur if the RWCU system did not continuously draw some water out of the lower vessel head drain. Thus, it is difficult for a non-floating object to be levitated by coolant flow.

- Most materials which could be carried up to the orifices (e.g., the Browns Ferry rubber shoe cover and the Duane Arnold aluminum can) chemically decompose very rapidly at reactor coolant temperatures and pressures.

- most loose parts, whether of internal or external origin, are made of stainless steel, Zircaloy, or some other material which does not float. In BWRs, such objects tend to settle to the bottom of the lower plenum.

Orifice blockage events may or may not be discovered by inspection. However, since blockages change power level in an assembly, blockage would also affect power maps and spent fuel isotopics. Thus far, to the best of our knowledge, no blockage events have occurred.

Presently, about 370 BWR-years of experience have been accumulated. No events have been reported in this interval. Thus, to 95% confidence, the frequency of detectable blockage events is limited by:

In reality, of course, we expect the frequency of such events to be much less than this figure for the reasons given above.

We will make a further assumption that all degrees of blockage are equally likely. (In reality, it is expected that modest blockages would occur more often then severe ones, so this is a conservative assumption.) NEDO-10174, Revision 1288 concluded that 79% blockage would result in DNB and that 98% blockage would result in fuel melting.

F(DNB) (0.19)(8.1 x 10-3/RY) = 1.5 x 10-3/RY

F(Melt) (0.02)(8.1 x 10-3/RY) = 1.6 x 10-4/RY

Consequence Estimate

Localized DNB failures are not severe events. A special CRAC calculation of a rod drop event (in which 770 fuel pins were assumed to fail) gave a result of 7 x 103 man-rem under our usual assumption of 340 people per square mile. Scaling this figure down to one 62-pin fuel assembly, the result of a DNB event becomes 5.6 x 10-4 man-rem. Using the upper limit frequency given in the previous section and assuming a 40-year plant life, the DNB hazard is, at most, about 3 x 105 man-rem/plant and is probably much less. Consequently, DNB failures will not be considered further.

One assembly melting will produce more of a radioactive release, although (according to the GE study) pressure pulses are not expected to be severe. The effects of such an event can be bounded by scaling a BWR-4 release to one assembly. A BWR-4 release is a complete core-melt with enough containment leakage to prevent containment failure due to overpressure. Since one assembly is not very likely to overpressurize the containment, this is a very conservative estimate. In reality, the release would be limited by the high steam line radiation isolation and also by the offgas treatment system isolation on high outlet radiation. Thus, since a typical large BWR has 764 fuel assemblies and a complete core-melt producing a BWR-4 release results in 6.1 x 105 man-rem, we can set a limit on consequences of:

Currently, there are 24 BWRs in operation with an accumulated experience of about 370 RY. Additionally, 23 more BWRs are under construction. Assuming a 40-year plant life, there are 590 RY remaining for operating BWRs and 920 RY for future BWRs. This represents a total remaining reactor life of 1,510 RY. Therefore, the total risk reduction associated with this issue is (1.6 x 10-4)(800)(1,510) man-rem or 193 man-rem.

Cost Estimate

Industry Cost: PNL postulated64 an engineering solution consisting of one-way flapper valves to admit coolant into the lower portion of the fuel assembly. Each BWR would have to pay an incremental cost on each new fuel assembly for the additional holes and flapper valves in the lower tie plate. On the average, a plant replaces about 1/3 of its core during annual refueling outages. Assuming 600 and 750 fuel assemblies per core for backfit and forward-fit BWRs, respectively (based on BWR design information), one obtains the following refueling rates:

Backfit BWR: 133 fuel assemblies/RY
Forward-fit BWR: 167 fuel assemblies/RY

Only the incremental cost due to design changes in the assemblies is credited here. Assuming this amounts to $250/fuel assembly, the operation/maintenance costs (interpreted as refueling costs) are as follows:

Backfit BWR: (133 fuel assemblies/RY) ($250/fuel assembly)= $33,000/RY
Forward-fit BWR: (167 fuel assemblies/RY) ($250/fuel assembly)= $42,000/RY

For the 24 BWRs in operation, the backfit costs are ($33,000/RY)(590 RY) or $19.5M. Forward-fit costs are ($42,000/RY)(920 RY) or $38.6M for the 23 BWRs under construction. Thus, the total industry cost is $58.

NRC Cost: NRC costs are estimated to be 4 man-weeks for review of the GE topical report, 12 man-weeks for review of engineering solutions, and 1 man-week/plant for supporting implementation. Therefore, the total NRC cost is estimated to be approximately $120,000.

Value/Impact Assessment

Based on an estimated risk reduction of 193 man-rem and a total cost of $58.2M, the value/impact score is given by:


Based on the above value/impact score, this issue should be placed in the DROP category.


0064.NUREG/CR-2800, "Guidelines for Nuclear Power Plant Safety Issue Prioritization Information Development," U.S. Nuclear Regulatory Commission, February 1983, (Supplement 1) May 1983, (Supplement 2) December 1983, (Supplement 3) September 1985, (Supplement 4) July 1986, (Supplement 5) July 1996.
0288.NEDO-10174, "Consequences of a Postulated Flow Blockage Incident in a Boiling Water Reactor," General Electric Company, October 1977, (Rev. 1) May 1980.
0380.SECY-93-049, "Implementation of 10 CFR Part 54, `Requirements for Renewal of Operating Licenses for Nuclear Power Plants,'" U.S. Nuclear Regulatory Commission, March 1, 1993. [9303030337, 9303100375].
0381.Memorandum for W. Minners from O. Parr, "Prioritization of Proposed Generic Issue on CRD Accumulator Check Valve Leakage," August 13, 1984. [8408280264]