Question 1: Three Stages of Decommissioning
Describe the three internationally recognised stages of nuclear decommissioning. For each stage, state the key activities and give an approximate timescale.
Answer:
Stage 1 — Post-Operational Clean Out (POCO) / Initial Decommissioning: Removal of fuel from the reactor, draining of circuits, removal of process materials, decontamination of accessible areas. This removes the major source term (spent fuel) and significantly reduces the site hazard. Timescale: typically a few years after final shutdown.
Stage 2 — Care and Maintenance (Safestore) / Partial Dismantling: The reactor building is sealed, auxiliary buildings may be demolished, but the reactor pressure vessel and biological shield remain intact. The site enters a period of passive surveillance with minimal staffing. The purpose is to allow short-lived activation products (particularly Co-60, T½ = 5.27 years) to decay, reducing dose rates for eventual dismantling. Timescale: approximately 50—85 years.
Stage 3 — Final Dismantling / Site Clearance: Complete dismantling of all remaining radioactive structures (RPV, bioshield), remediation of contaminated land, final radiation survey, and delicensing. The site is released for unrestricted use (greenfield end-state) or restricted use (brownfield end-state). Timescale: may occur 85—100+ years after shutdown.
Question 2: Safestore vs Immediate Dismantling
(a) Explain the safestore (deferred dismantling) approach to decommissioning and state two advantages. (b) State two disadvantages of safestore compared with immediate dismantling. (c) Under what circumstances might immediate dismantling be preferred?
Answer:
(a) Safestore involves sealing the reactor building after Stage 1 (POCO) and deferring final dismantling for a period of approximately 85—100 years. During this time, short-lived radionuclides (especially Co-60) decay significantly, reducing dose rates and simplifying later dismantling. Advantages: (1) Worker doses during dismantling are much lower because radioactivity has decayed; (2) The volume of radioactive waste may be reduced because some material will have decayed below clearance levels.
(b) Disadvantages: (1) The site remains under institutional control for a very long period, requiring ongoing surveillance, maintenance, and security — this requires funding and organisational continuity over decades; (2) Knowledge and records may be lost over such long periods, making future dismantling more difficult; the workforce with direct experience of the plant will no longer be available.
(c) Immediate dismantling may be preferred when: the site is needed for redevelopment; the operator has strong financial resources and experienced workforce available now; or regulatory/public pressure demands prompt site clearance. It is also preferred for smaller facilities where the radiological challenge is manageable.
Question 3: Decommissioning Worker Dose Calculation
During decommissioning of a reactor, an activated steel component contains Co-60 with a total activity of 5 × 10¹¹ Bq. The specific gamma ray constant for Co-60 is Γ = 3.41 × 10⁻¹³ Sv·m²·Bq⁻¹·h⁻¹. (a) Calculate the dose rate at 2 m from the component. (b) A worker must spend 30 minutes at this distance. Calculate the dose received. (c) Compare this with the annual occupational dose limit. (d) If the same work is deferred by 25 years, calculate the residual Co-60 activity and the new dose rate at 2 m. (Co-60 T½ = 5.27 years.)
Answer:
(a) Ḋ = AΓ/r² = (5 × 10¹¹ × 3.41 × 10⁻¹³) / 2² = 0.1705 / 4 = 0.0426 Sv/h = 42.6 mSv/h.
(b) Dose = 42.6 × 0.5 = 21.3 mSv.
(c) The annual occupational dose limit is 20 mSv. This single 30-minute task would deliver 21.3 mSv, exceeding the annual limit. The work cannot proceed as described — remote handling, additional shielding, or shorter exposure time is required.
(d) After 25 years: number of half-lives = 25/5.27 = 4.74. Decay factor = 2^4.74 = 26.7. Residual activity = 5 × 10¹¹ / 26.7 = 1.87 × 10¹⁰ Bq. New dose rate = (1.87 × 10¹⁰ × 3.41 × 10⁻¹³) / 4 = 1.60 × 10⁻³ Sv/h = 1.60 mSv/h. A 30-minute exposure now gives 0.80 mSv — well within limits. This illustrates the value of the safestore approach.
Question 4: Waste Classification Calculation
During decommissioning, the following waste streams are generated. Classify each using the UK system (LLW: ≤12 GBq/te beta-gamma, ≤4 GBq/te alpha; ILW: exceeds LLW limits but does not require active cooling; HLW: requires active cooling due to heat generation).
(a) 500 kg of activated concrete containing 3.0 × 10¹⁰ Bq of beta-gamma activity. (b) 200 kg of contaminated PPE containing 1.5 × 10⁶ Bq of beta-gamma activity. (c) 50 kg of reactor internals containing 8.0 × 10¹³ Bq of beta-gamma activity and generating measurable heat.
Answer:
(a) Concentration = 3.0 × 10¹⁰ / 0.5 te = 6.0 × 10¹⁰ Bq/te = 60 GBq/te. Exceeds 12 GBq/te → ILW. No significant heat generation → confirmed ILW.
(b) Concentration = 1.5 × 10⁶ / 0.2 te = 7.5 × 10⁶ Bq/te = 0.0075 GBq/te. Well below 12 GBq/te → LLW. (Could potentially be Out of Scope Waste if below the EPR 2016 threshold values.)
(c) Concentration = 8.0 × 10¹³ / 0.05 te = 1.6 × 10¹⁵ Bq/te = 1.6 × 10⁶ GBq/te. Vastly exceeds LLW limits → ILW or HLW. Since it generates measurable heat requiring active cooling, this is HLW.
Question 5: Co-60 Decay and Waste Reclassification
An activated bioshield concrete block (mass 5 tonnes) contains Co-60 with an initial activity of 2.0 × 10¹² Bq. The half-life of Co-60 is 5.27 years. (a) Calculate the activity after 50 years of storage. (b) Express the activity concentration in GBq/te at that time. (c) Would the block then qualify as LLW? (d) How many years of storage would be required for the block to fall below the LLW beta-gamma threshold of 12 GBq/te?
Answer:
(a) Number of half-lives = 50/5.27 = 9.49. Activity = 2.0 × 10¹² / 2^9.49 = 2.0 × 10¹² / 718.4 = 2.78 × 10⁹ Bq.
(b) Concentration = 2.78 × 10⁹ / 5 = 5.57 × 10⁸ Bq/te = 0.557 GBq/te.
(c) Yes — 0.557 GBq/te is well below the 12 GBq/te LLW threshold, so after 50 years the block qualifies as LLW. It could be disposed of at the LLWR near Drigg.
(d) Initial concentration = 2.0 × 10¹² / 5 = 4.0 × 10¹¹ Bq/te = 400 GBq/te. Need to reach 12 GBq/te. Required attenuation = 400/12 = 33.3. Number of half-lives: 2^n = 33.3 → n = ln(33.3)/ln(2) = 3.51/0.693 = 5.07 half-lives. Time = 5.07 × 5.27 = 26.7 years. After approximately 27 years, the block would qualify as LLW.
Question 6: NDA and Regulatory Framework
(a) What is the NDA and when was it established? (b) State the NDA’s primary role in decommissioning. (c) Name the independent nuclear safety regulator in the UK and state its role. (d) State four licence conditions relevant to decommissioning.
Answer:
(a) The Nuclear Decommissioning Authority (NDA) was established in 2005 under the Energy Act 2004.
(b) The NDA is responsible for the strategic management of the UK’s nuclear legacy — it oversees the decommissioning and clean-up of 17 nuclear sites. It is a strategic body, not an operator; it contracts Site Licence Companies (SLCs) to carry out the work. Its annual budget is approximately £3.2 billion.
(c) The Office for Nuclear Regulation (ONR), established as an independent statutory body in 2014 under the Energy Act 2013. ONR is responsible for nuclear safety, security, and conventional health and safety regulation at licensed nuclear sites.
(d) Four relevant licence conditions: LC14 — Safety documentation (requires an adequate safety case); LC25 — Operational records (records of operations must be kept); LC32 — Accumulation of radioactive waste (waste must not accumulate unnecessarily); LC35 — Decommissioning (the licensee must prepare and implement an adequate decommissioning programme).
Question 7: Decontamination Methods
(a) Distinguish between chemical and mechanical decontamination methods. Give two examples of each. (b) For each example, state what type of surface or material it is most suitable for. (c) A stainless steel vessel in a reprocessing plant has internal surface contamination of 10⁵ Bq/cm². Suggest a decontamination approach and explain your choice.
Answer:
(a) Chemical methods dissolve or react with the contamination layer: (1) Acid washing — using dilute nitric acid or citric acid to dissolve surface oxide layers and remove contamination; (2) Foam decontamination — application of chemical foam that clings to surfaces, allowing prolonged chemical contact without generating large liquid waste volumes. Mechanical methods physically remove the contaminated surface layer: (1) Abrasive blasting (grit or CO₂ pellet blasting) — removes surface layers by impact; (2) Scabbling — uses pneumatic or microwave tools to chip away the surface layer of concrete.
(b) Acid washing: suitable for metal surfaces (pipes, vessels, tanks) where contamination is in the oxide layer. Foam: suitable for large, complex surfaces including walls and ceilings. Abrasive blasting: suitable for metal surfaces and structural steel. Scabbling: suitable for concrete surfaces (floors, walls, bioshield).
(c) For a stainless steel vessel with high surface contamination (10⁵ Bq/cm²): chemical decontamination using successive acid washes would be the first approach. Stainless steel resists corrosion, so controlled acid treatment can dissolve the contaminated oxide layer without damaging the base metal. If chemical methods do not achieve adequate decontamination, mechanical methods (abrasive blasting) or sectioning for disposal as radioactive waste may be required.
Question 8: Site End-State Assessment
(a) Define greenfield and brownfield end-states for a decommissioned nuclear site. (b) State the dose constraint that applies to members of the public from a delicensed site. (c) A final radiation survey of a decommissioned site shows residual Cs-137 contamination giving a dose rate of 0.15 μSv/h at 1 m above the ground across a 2-hectare area. Estimate the annual dose to a hypothetical resident spending 8 hours per day outdoors on the site for 365 days per year. Would this meet the criterion for unrestricted release?
Answer:
(a) Greenfield: the site is returned to a condition where it can be used for any purpose without restriction — all radioactive material has been removed and residual contamination is below clearance levels. Brownfield: the site is released for restricted use — some structures or residual contamination may remain, but the site is safe for specified uses (e.g., industrial) with appropriate controls.
(b) The dose constraint for public exposure from a single source (such as a delicensed site) is 0.3 mSv/year (source constraint). The overall site constraint is 0.5 mSv/year. The statutory public dose limit is 1 mSv/year from all artificial sources.
(c) Annual outdoor time = 8 × 365 = 2,920 hours. Annual dose = 0.15 μSv/h × 2,920 h = 438 μSv = 0.438 mSv. This exceeds the 0.3 mSv/year source dose constraint, so the site would NOT meet the criterion for unrestricted (greenfield) release at this contamination level. Further decontamination or remediation would be required, or the site could be released as brownfield with land use restrictions (e.g., industrial only, limiting occupancy time).