Radioactivity level in the coolant water

[oksuz_]

Hi,

Can anyone give me a source from which I can obtain the radioactivity level in the coolant water circulating in the primary circuit?

Thanks in advance.

 

[gmax137]

Standard Tech Specs for Westinghouse plants (NUREG-1431) says below 1 uCi/gram (Equivalent I-131). I'm pretty sure this is the spec number for all PWRs. [edit] NUREG-1432 for the CE plants has the same value. Actual activity is much much less in operating plants.

 

[oksuz_]

Thank you for the information. I have made a quick search and seen that there is a couple of volume of NUREG-1431. Do you know at which volume has this info? I would like to know How NRC determined this figure.

Another thing is that this number somehow seems relatively big to me when considering the processes which make the coolant water radioactive. As far as I know, the processes are as follows - Triton production via consecutive neutron capture of Hydrogen
- (n,p) reaction of O-16 making it Nitrogen-16
-Neutron-activated metal core structures. Corrosion makes these radioactive nuclei dissolve in water
- And trace amount of fission fragments which escape from fuel pin.

Please correct me if I am wrong about the processes.

As you mention that the actual radioactivity level is much less. Is this owing to that the processes given above are not prominent when considering normal reactor operation?

I am going to try making a rough calculation to get a number related to the radioactivity level. At least for the first two processes since their calculations require less and easily-obtainable information.

 

[Hiddencamper]

In normal operation, the contamination levels in the fuel are less. If you have leaking fuel cladding, you may approach those limits.

The limits are high enough to support some limited fuel leakers. In BWR plants, we’ve even seen some fish mouth type ruptures where local power was suppressed and core operation was able to continue through the cycle.

With no leakers though, your cleanup or chemical volume control systems will help keep the water pretty pure. Chemistry is normally very carefully maintained to minimize corrosion products and crud.

 

[Astronuc]

Hi,
Can anyone give me a source from which I can obtain the radioactivity level in the coolant water circulating in the primary circuit?

Is one looking for activity limits, or actual activity values?

Radiochemistry in Nuclear Power Reactors is a good basic reference.
https://www.nap.edu/catalog/9263/radiochemistry-in-nuclear-power-reactors

Very little tritium comes from successive neutron capture in hydrogen. Most tritium comes from B in the coolant or burnable poison assemblies, and Li-6 in the LiOH buffer the coolant, even though LiOH is enriched in Li-7 and depleted in Li-6. In BWRs, boron is used in control blades, and some of the T produced in the control blades leaks into the coolant.

Corrosion products (metal ions) dissolved in the coolant may traverse the core, or may accumulate on the fuel (crud deposits), and become activated. Reactor coolant systems have resin filters to remove metal ions from the coolant.

Fission products can be released from failed (breached) fuel rods, but are also present as 'tramp' uranium.

Utilities are required to have programs in place to limit exposure to workers and keep coolant activity as low as reasonably achievable. There is also an economic incentive to maintain low coolant activity, since it is more costly to dispose of the waste products collected on the filters. Utilities have water chemistry programs to mitigate corrosion and prevent stress corrosion cracking of structural alloys.

The dose limits for iodine and noble gases were established during the AEC days and inherited by the NRC. I have seen very high coolant activities in some of the earliest operating plants in the 1970s that would be unacceptable today, and I know of two plants that were within hours of shutting down due to high iodine activity in the reactor coolant. In one case, it was a single high power fuel rod that had failed and degraded.

There were several BWR plants in the 1980s and 1990s that had to shutdown to remove failed fuel, before it became common practice to use power suppression testing (flux tilting) to identify the location of failed fuel by how the coolant activity responded when control blades where inserted in the core.

 

[gmax137]

Thank you for the information. I have made a quick search and seen that there is a couple of volume of NUREG-1431. Do you know at which volume has this info? I would like to know How NRC determined this figure.

Another thing is that this number somehow seems relatively big to me when considering the processes which make the coolant water radioactive.

The Tech Spec is in 3.6.16 if I recall correctly; that is in Volume 1. But I am sorry if I have led you astray; the Tech Spec is a legal limit. The value (1uCi/g) is selected as a design value, used to design the shielding and physical layouts of the plant systems (allowing for the plant operators to run the plant within the dose limits of 10CFR20). The value also gets used in various safety analyses which must consider a maximum activity (steam generator tube rupture, other events with RCS leakage). It is included in the Tech Specs to preserve the validity of the analyses. The numerical value probably is what was used during the design of the Nautilus reactor in the 1950s.

But all of this is irrelevant to a discussion of what the actual activity values are, under normal operating conditions. Astronuc's post above has much useful information.

 

[Astronuc]

The value (1uCi/g) is selected as a design value, used to design the shielding and physical layouts of the plant systems (allowing for the plant operators to run the plant within the dose limits of 10CFR20). The value also gets used in various safety analyses which must consider a maximum activity (steam generator tube rupture, other events with RCS leakage).

This is the basis for maximum allowable activity. In reality, coolant activity is much lower. As I recall, we typically see numbers on the order of 10^-5 to 10^-4 uCi/g. I did a lot of failed fuel simulations back in the 90s and early 2000s, and part of the analysis was looking at the background activity before failure. Failed fuel releases not only fission products and uranium to the coolant, but also transuranics, e.g., Np-239, Pu nuclides, and smaller amounts of others, which may persist in the reactor coolants system for several cycles (years).

From RADIATION PROTECTION ASPECTS OF PRIMARY WATER CHEMISTRY AND SOURCE-TERM MANAGEMENT, April 2014
https://www.oecd-nea.org/rp/docs/2014/crpph-r2014-2.pdf

The main contributors to the radiation field generation (and then to the collective dose during outage) are activated corrosion products. The most important radionuclides from a radiation perspective are ⁵⁸Co, ⁶⁰Co, ^110mAg, ^124,5Sb, ⁵⁹Fe, ⁵⁴Mn, ⁵¹Cr, ⁹⁵Zr, and ⁹⁵Nb. Two main sources are usually defined for these radionuclides: out-of-core corrosion products (steam generator corrosion products, etc.) and fuel assembly and/or materials corrosion products (reactor internals, etc.). The first possibility to limit radiation field generation is to limit corrosion of materials and the second is to limit concentration of elements such as Ni and Co in these materials (e.g. playing on plant design, flow and chemistry of the primary coolant). From a radiation protection perspective, “the most important material issue is the corrosion resistance of the material”.

Cobalt comes from activated Ni as well as tramp Co in Ni. Fe, Mn, Cr come from stainless steel. Silver may come from control rods that have cracked or worn through. Sb may be from secondary neutron sources (Sb-Be). Zr, Nb would likely come from fuel cladding using Zr-Nb alloys. Sn is a minor constituent of Zircaloy (Zr - 1.3-1.5 Sn - 0.2 Fe - 0.1 Cr) and ZIRLO (Zr - 1 Nb - 1 Sn - 0.1 Fe). I believe most of the world use Bq/ml or MBq/kg (or GBq/m³) is discussing coolant activity.

In 2005, adverse industry trends in Radiation Protection were a key factor in the development of the NEI/EPRI/INPO RP 2020 Initiative that had the stated goal of ‘Taking Radiation off the Table.’ EPRI was charged with taking the technical lead for Radiation Source Term Reduction. In response to this initiative, the EPRI Chemistry and LLW Technical Advisory Committee strongly recommended that EPRI restart PWR radiation field data collection efforts to help quantify the effects of plant changes such as replacement steam generators, core uprating, adverse radiological incidents, and various changes in shutdown and normal chemistry procedures. These changes have caused unpredictable fluctuations in dose rates throughout the out-of-core surfaces, and a more fundamental understanding is required.

In 2007, the programme was reinstated and currently 129 units have submitted data to the programme. Several projects beginning in 2007 have used the collected data to evaluate a consider the effect of parameters such as plant age, chemistry control methodology, effective full power year (EFPY), coolant chemistry, cobalt source terms, and start–ups and shutdowns. These factors have been evaluated and published in other EPRI reports.