Final Report — Evaluation of Chemical Effects Phenomena in Post-LOCA Coolant (NUREG/CR-6988)

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Publication Information

Manuscript Completed: December 2008
Date Published:
January 2009

Prepared by:
C.H. Delegard1, M.R. Elmore1, K.J. Geelhood1, M.A. Lilga1,
W.G. Luscher1, G.T. MacLean2, J.K. Magnuson1, R.T. Pagh1,
S.G. Pitman1 and R.S. Wittman1

1Pacific Northwest National Laboratory
902 Battelle Boulevard
Richland, WA 99352

2Fluor Federal Services
120 Jadwin Avenue
Richland Washington, 99352-3448

J.P. Burke, NRC Project Manager

Prepared for:
Office of Nuclear Regulatory Research
U. S. Nuclear Regulatory Commission
Washington, D.C. 20555-0001

NRC Job Code N6381

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During a loss-of-coolant accident (LOCA) in a pressurized water reactor (PWR), the integrity of the reactor coolant system is abruptly lost through a breach of the system pressure boundary. Remedial actions initiated automatically to limit the damage from the accident include 1) actuation of the containment spray system (CSS) for most plants to maintain containment building pressure below pre-established set points to prevent building damage and to scrub radioiodine from the containment air (minimizing offsite doses) and 2) actuation of the emergency core cooling system (ECCS) to provide an emergency supply of borated water from the refueling water storage tank to the core for cooling and to maintain the core subcritical. The reactor coolant, containment spray, and borated water injected by the ECCS drain to the floor of the containment building are collected by the recirculation sump with debris generated by the LOCA and latent debris pre-existing in the wetted parts of the containment building. Ultimately, the collected water is pumped from the sump, through a heat exchanger, into the reactor pressure vessel, and typically also, just after the rupture, through the containment spray nozzles in a recirculation loop. The recirculation flow encounters the sump strainer where solids may accumulate to block coolant flow or be ingested into the pump suction line to potentially damage the pump, irradiated fuel in the core, or other structures.

The Nuclear Regulatory Commission established Generic Safety Issue 191, “Assessment of Debris Accumulation on PWR Sump Performance”, to evaluate the effects of post-LOCA debris on the performance of the ECCS and CSS in recirculation mode at PWRs. Experimental testing and other studies have been completed to determine the impacts of cooling water composition, debris sources, and materials corrosion on the nature of the debris, presuming no fuel cladding failure. However, historical, ongoing, and planned testing and analysis studies were evaluated, and 10 further topics related to chemical effects were identified that deserve additional consideration.

The 10 topical areas are radiation effects (particularly on material corrosion), differences in concrete carbonation between tested systems and existing containment structures, effects of alloy variability between tested and actual materials, galvanic corrosion effects, biological fouling, co-precipitation, and other synergistic solids formation effects, inorganic agglomeration, crud release effects (types and quantities), retrograde solubility and solids deposition, and organic material impacts. Sufficient data or prior related studies were available to sufficiently address some of the questions raised in the 10 topic areas. However, within these 10 broad areas, topics meriting additional consideration also were identified and are the focus of this report.

The topic with the greatest perceived influence on ECCS performance is the interactions of organic materials (lubricants and coatings) with inorganic solids. The effects of radiolysis on redox potential and thus metal corrosion have the next most influence. Of similar influence are the effects of biological growth in the post-LOCA system and the impacts of dried borate salts on hot fuel cladding and reactor pressure vessel materials. Of lesser, but not insignificant, influence are galvanic corrosion, inorganic agglomeration, and crud release effects on increasing and altering solids delivered to the post-LOCA coolant. Changes in concrete carbonation and differences in alloy corrosion rates were judged to have minor impacts on ECCS functionality.

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