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Technical Meeting on Design Extension Conditions for Storage
Facilities for Power Reactor Spent Fuel 27 June – 1 July 2016 651-T1-TM-52204 room C0454
Fuel Storage in Nuclear Fuel Cycle Facilities
Ramon GATER
RRSS/NSNI
1) IAEA Safety standards for fuel storage
2) Status of DEC for NFCF
3) Levels for Practical Elimination
4) Analysis and Categorisation
5) DEC for NFCF
6) Safety Reassessment 1 3
7) FINAS and SEDO
Introduction
Standards for Fuel Safety
“establishes the safety
requirements that apply to
all facilities …. such as
nuclear power plants or
spent fuel reprocessing
plants. In this publication,
the term ‘facility’ is used to
refer to either of these
possibilities.”
GSR Part 5 – Generic Standard
1) SSR 2/1 and NS-G-1.4 are specific standards
that apply to pools on the reactor site. LWR
pools are closely linked to a reactor core and
may contain fuel with partial (or no) burnup.
2) NS-R-5 and SSG-42 are specific standards that
apply to reprocessing with centralised interim
storage facilities for spent fuel.
3) SSG-15 covers the fuel and not the facility that
contains it.
Identification of Standards
1) Design rules and criteria;
2) Design basis;
3) Codes and standards;
4) Design for reliability;
5) Human factors and ergonomic considerations;
6) Material selection and ageing management;
7) Provisions for inspection, testing, and maintenance;
NS-R-5 – Design
Design Basis for NFCF
Usually multiple Design Basis scenarios for an NFCF
1) SSR 2/1 is the principle IAEA reference that defines DEC.
2) Currently, only SSR 2/1 and SSG-42 (reprocessing guide) cover DEC.
3) NS-R-5 and NS-G-1.4 are being updated to cover DEC.
4) SSR-4 (NS-R-5) will cover DEC, and all nuclear fuel facilities that are not reactors, including related chemical hazards.
Status of DEC for NFCF
The government shall ensure that protection strategies are developed, justified and optimized at the preparedness stage for taking protective actions and other response actions effectively in a nuclear or radiological emergency.
A reference level expressed in terms of residual dose shall be set, typically as an effective dose in the range 20–100 mSv, acute or annual, that includes dose contributions via all exposure pathways.
GSR part 7 Requirement 5
….On the basis of the outcome of the justification and the optimization of the protection strategy, national generic criteria for taking protective actions and other response actions, expressed in terms of projected dose or of dose that has been received, shall be developed with account taken of the generic criteria in Appendix II. If the national generic criteria for projected dose or received dose are exceeded, protective actions and other response actions, either individually or in combination, shall be implemented.
Nat. Regulator defines Generic Intervention Levels
GSR part 7 Requirement 5
5.31. The design shall be such that the possibility of
conditions arising that could lead to an early radioactive
release or a large radioactive release is ‘practically
eliminated’.
The design shall be such that for design extension
conditions, protective actions that are limited in terms of
lengths of time and areas of application shall be sufficient for
the protection of the public, and sufficient time shall be
available to take such measures.
The possibility of certain conditions occurring is considered to have been
practically eliminated if it is physically impossible for the conditions to occur or if
the conditions can be considered with a high level of confidence to be extremely
unlikely to arise.
TecDoc 1791 provides guidance
SSR-2/1 Practical Elimination
Some Member States may identify Generic
Intervention Levels (GIL) with protective
measures that are of limited scope in terms of
area and time shall be necessary for protection
of the public, with sufficient time available for
implementation.
This is a decision for Member States
GIL and Protective Measures
Facilities and locations containing;
I. Recently discharged irradiated reactor fuel with a total of more than about 0.1 EBq (1017 Bq) of 137Cs, equivalent to the inventory in a 3000 MW(th) reactor core.
II. Recently discharged irradiated reactor fuel requiring active cooling.
III. Facilities with inventories of radioactive material sufficient to result in doses warranting urgent protective action being taken on the site.
- will generally have the indicated category
Categories I, II and III
Based on Generic Intervention Level (GIL), dry
storage casks are generally Cat III;
1) No potential for off-site doses in excess of
urgent GILs.
2) No potential for on-site doses in excess of
urgent GILs from inhalation. If shielding is
lost, direct shine dose could exceed urgent
GILs.
Dry Storage Casks
1) At-reactor pool;
• (Un)coupled to core
• In/out of containment
2) Dry storage casks
3) Centralised pools (e.g. at reprocessing facility)
4) Centralised dry-stores
5) Other activities and facilities, e.g. Th breeder
Require Accident Analyses for
A graded approach must be taken when considering DEC for NFCF…..
The following list developed from list presented to NUSSC in June 2015:
1) Loss of two or more independent criticality controls
2) Overheating of spent fuel and vitrified high-level waste
3) Uncovering spent fuel or Th/U breeder in a pool
4) Extended blackout of a large reprocessing facility
5) Loss of cooling of Highly Active Liquor
6) Overfilling a UF6 storage or process vessel
7) Major disruption of a plutonium facility
…….these are not listed in SSR-4
Suggested DEC for NFCF
A graded approach must be taken when considering DEC for NFCF…..
The following list developed from list presented to NUSSC in June 2015:
1) Loss of two or more independent criticality controls
2) Overheating of spent fuel and vitrified high-level waste
3) Uncovering spent fuel or Th/U breeder in a pool
4) Extended blackout of a large reprocessing facility
5) Loss of cooling of Highly Active Liquor
6) Overfilling a UF6 storage or process vessel
7) Major disruption of a plutonium facility
…. Some were heavily engineered for a long time
Suggested DEC for NFCF
Severe Accidents Considered
“Conditions …. for which
additional safety measures
may be needed ..”
i.e. Design Extension
Conditions
to be identified using a gap
analysis derived by safety
reassessment.
Safety Reassessment - 1
First check the design basis, using figure1 in SRS-90
Safety Reassessment - 2
Then consider
potential DEC,
figure 2 in SRS-90;
Safety Reassessment - 3
SEDO - Intended as a review by peers,
conducted by a team of international experts
with experience in FCF technology and
operations.
• Scope, duration and countries doing the
reviewing are negotiable.
• Judgments on facility safety performance
based on IAEA NFCF Safety Standards.
SEDO Review Service
Simple, efficient means to share
information on fuel cycle facilities
and learning on safety significant
events:
• Identify causes
• Share lessons learned
• Prevent repeating events
FINAS is the only international
incident reporting system for
NFCF. Reports not made public.
FINAS and iNFCIS Databases
a) For doses for which protective actions and other response
actions are expected to be undertaken under any circumstances
in a nuclear or radiological emergency to avoid or to minimize
severe deterministic effects;
b) For doses for which protective actions and other response
actions are expected to be taken, if they can be taken safely, in a
nuclear or radiological emergency to reasonably reduce the risk
of stochastic effects;
c) For doses for which restriction of international trade is warranted
in a nuclear or radiological emergency, with due consideration of
non-radiological consequences;
d) For doses for use as a target dose for the transition to an existing
exposure situation.
GSR part 7 Appendix II
Exposure Dose
If the dose is projected:
• Take precautionary urgent protective
actions immediately (even under
difficult conditions) to keep doses
below the generic criteria;
• Provide public information and
warnings;
• Carry out urgent decontamination.
Adred marrow 1 Gy
ADfetus 0.1 Gy
ADtissue 25 Gy at 0.5cm
ADskin 10 Gy to 100cm2
GSR part 7 Appendix II
Table II.1 (extract) Acute external exposure (<10 h)
• AD is an average relative biological effectiveness (RBE) weighted
absorbed dose from penetrating radiation.
• Other tables provided for different types of exposure.
