AN OPTIMAL CONTENT AND EXTENT OF THE LICENSING DOCUMENTATION FOR DUKOVANY NPP MODERNISED FUEL CYCLES
Tinka I., Tinková E., Nuclear Research Institute Řež, division Energoprojekt,
Czech Republic
ABSTRACT
The documentation, required for permission to carry out changes influencing nuclear safety, has to cover according to Czech Atomic law:
1) description and substantiation of changes under preparation,
2) updating of documentation approved in the frame of the Nuclear Power Plant (NPP) start-up and operation,
3) expected time schedule of changes,
4) demonstration that consequences of changes being performed do not influence nuclear safety adversely.
Documentation under point 2 above has to be approved by the Regulatory Body directly and its content is more or less given (safety analysis report, limits and conditions, etc). The other items (1, 3 and 4) have no unified form or fixed structure. Thus the most demanding step is connected with updating of mentioned documentation under point 2. Preferably it is desirable to make changes in approved documentation minimally.
A procedure consisting of the following three levels has been elaborated for Dukovany NPP as an effective tool for licensing of modernised fuel cycles [1] :
- safety evaluation of individual fuel cycles as level 1,
- enlarged safety evaluation as level 2 with two “sublevels”:
a) without safety analyses
b) including restricted number of analysed cases
- elaboration of Amendment of the Safety Analysis Report, in scope of Reactor
and Safety analyses, as level 3. Usually partial modifications in Limits and Conditions follow as well.
STANDARD SAFETY EVALUATION
This level of safety evaluation represents a standard step repeated for each of individual fuel cycles. The sense of this level is to show that all initial parameters of the core (including whole cycle burnup) are bounded by those, used for safety analyses and thus to confirm the validity of Safety Analysis Report (SAR) for evaluated fuel cycle. If this is confirmed without any exception, no further analyses are required. The variety of fuel assemblies arrangement in the core is typical reason for evaluated parameters changes. If necessary, additional analysis of individual cases can be needed, but still without no consequence for safety documentation.
In practice, this evaluation is performed by Dukovany physicists and archived in the form of a set of tables containing all key physical parameters playing basic role as input data for safety analyses. This form is fully acceptable for the Czech State Office for Nuclear Safety (SONS).
Additionally to this procedure, based on all results for all units, periodical evaluation is realised. This evaluation serves as independent check-up pursued by Energoprojekt and also as an occasion to change bounding values for further safety analyses. A graphic form, rather than tables, is much more convenient for such purposes, because this form enables a global view on parameters evolution, as shown in Example 1 and Example 2. Up to now it was not needed to split SAR into versions for different units. Thus, values demonstrated in these examples, contain data (undistinguished) for all four units of Dukovany NPP (in principle, of course, it is possible to obtain similar figures for individual units).
Presented type of figures is produced for each of checked parameters and usually contains (see examples 1 and 2) minima and maxima obtained by calculation (according to methodology elaborated for these purposes), then corresponding bounding values used in safety analyses and again minima and maxima, including uncertainties. In Example 2 it is illustrated tendency in the parameter (here Doppler coefficient for end of cycle) to be higher. Generally, it is not desirable to have too large margins between set of calculated extremes and bounding values for analyses, then corresponding reduction should follow. It means, for the nearest planned analyses change in right direction is expected, as demonstrated in Example 2.
From figures of Example 1 and 2 type it can be directly seen, whether corresponding parameter is deviated from range of bounding values or not. If not, licensing procedure remains on level 1.
If checked parameter falls out of bounding range, then summary table, containing list of analysed events and relevant bounding values applied, has to be consulted for identification, which events can be influenced by this parameter. Following decisions are possible:
- deviation is very small and impact on discussed events can be easily evaluated as insignificant, based on available sensitivity studies licensing process remains on level 1
- deviation is more significant, but discussed events are lowly sensitive to such changes licensing process remains on level 1
- deviation is so significant, that discussed events has to be reanalysed under level 2 evaluation
In summary, this level 1 of safety evaluation can be characterised as the checking process without any impact on existing safety documentation. It is also one part of enlarged evaluation, described in the following text.
Example 1: Doppler temperature coefficient for nominal power at the beginning of cycle (BOC)
Example 2: Doppler temperature coefficient for nominal power at the end of cycle (EOC)
Abbreviations in graphs:
SAR bound.: bounding values used in Safety Analysis Report (SAR)
uncert.: uncertainties included
change: foreseen changes resulting from last end expected evolution of given parameter
cycle 23 – 28: expected cycles with 1444 MW nominal reactor power (under preparation)
ENLARGED SAFETY EVALUATION
Basic question for enlarged safety evaluation refers to circumstances requiring such type of evaluation. One possibility was mentioned above and it concerns values of checked parameters, exceeding range of key parameters used in valid SAR. Furthermore, there are cases, when enlarged evaluation is more reliable way to reach SONS approval, even if all parameters are found inside of checked ranges. Examples 3 and 4 show typical parameters for fuel assemblies, implemented up to now on Dukovany NPP units. Implementation each of these assemblies was accompanied by one of assumed levels of evaluation (not the same), although characteristics of assemblies are nearly identical (perhaps except assemblies with/without burnable Gd):
1) 3,82 % - implementation of fuel assemblies with radial enrichment profilation: level 3
2) 4,38 % - 1st generation of Gd assemblies (pitch 12,2 mm): level 3
3) 4,25 % - 2nd generation of Gd assemblies (pitch 12,3 mm): level 3
4) 4,25 % /opt. - see case 3, but radially optimised enrichment: level 2 (without analyses)
5) 4,25 % /opt. - see case 4, but implementation on unit with instrumentation and control system before reconstruction: level 2 (with analyses of selected events)
6) 4,38 % - modified 2st generation of Gd assemblies (pitch 12,3 mm): level 3
Under enlarged safety evaluation two sublevels are possible: first sublevel without additional safety analyses (case 4 above), second one with spectrum of selected events, additionally analysed (case 5 above). The second sublevel differs against level 3 in type of documentation, which still remains as supporting, without direct impact on SAR. Moreover, the spectrum of additionally analysed events make only a small fraction of spectrum contained in SAR.
Overall enlarged safety evaluation typically contains the following items:
1. Characteristics of individual fuel assemblies and fuel cycles
- properties of fuel assemblies in infinite arrangement:
- infinity multiplication factor and its dependencies,
- non-uniformity in power distributions (peaking factors),
- thermal characteristics of individual assemblies,
- properties of the core during fuel cycles burnup:
- boron acid concentration,
- maximum of relative fuel assembly power,
- enthalpy rise hot channel factor,
- heat flux hot channel factor,
- reactivity coefficients.
2. Requirements and restrictions for fuel operation
3. Safety evaluations – tables of key checked parameters (same as level 1 evaluation)
- nuclear characteristics of the core:
- reactivity coefficients and kinetics parameters,
- control assemblies worth,
- thermal-hydraulic characteristics of the core:
- axial power distributions,
- coolant temperatures and DNBR,
- specific parameters for specific events
4. Additional analyses and their results (minority against full spectrum in SAR)
5. Summary fuel cycles evaluation and consequences for safety analysis report
Major impact on SAR is not expected, but if needed, corresponding recommendations or proposals are contained under this item.
In summary, this level 2 of safety evaluation can be characterised as the checking process with supporting nuclear and thermal parameters of the fuel and fuel cycles, mostly in comparison with previous ones. Direct impact on existing safety documentation is not expected more then a few les significant modifications. Otherwise level 3 is under consideration.
Elaboration of Amendment of the Safety Analysis Report
This level 3 of evaluation does not need special description: It is very close to full elaboration of the SAR in parts of:
- Reactor: fuel system design, nuclear design, thermal and hydraulic design
- Safety analyses: whole spectrum including radiological consequences analysis
The difference against full elaboration of corresponding parts of SAR consist in possibilities to utilize unchanged results either in reactor or in safety analyses parts. Practical using of this approach is planned for licensing modified Gd fuel of the 2nd generation for new nominal reactor power level 1444 MW.
Example 3: Multiplication factor k, nominal power, equilibrium Xe, Sm, boron acid concentration 3 g/kg
Example 4: Relative pin power distribution, fresh fuel, no Xe, Sm
3,82 %: 4 % (84 rods), 3,6 %(24), 3,3 %(18)
4,25 %: 4,4 % (84), 4,0 %(30), 3,6 %(6), 4,0 % (6 with Gd)
4,25 % opt.: 4,6 % (54), 4,4 %(30), 3,7 %(30), 3,3 %(6), 4,0 % (6 with Gd)
4,38 %: 4,6 % (84), 4,0 %(30), 3,6 %(6), 4,0 % (6 with Gd)
CONCLUSION
Very well established licensing procedure is very important for NPP, because implementation of changes has to be prepared in advance enough, to obtain approval just when needed. Number of experts taking part in licensing process depends on extent of changes, but number of each other communicating institutions is nearly the same: NPP and its supporting organization(s) and SONS together with its supporting organization(s). That is why attentively prepared licensing process can significantly save time and costs and thus essentially can reduce time from application to approval of the SONS with licensed changes.
This has been main reason for introducing three levels of supporting/safety documentation with specified content as narrowly as possible, which is suitable for current modifications, especially in fuel cycles modernization. Based on this the extent of documentation reviewed by the SONS experts is known and it enables to plan particular steps more precisely.
REFERENCES
[1] Tinka I. Veselý P. Gerža J.: Licensing of new fuel cycles at Dukovany NPP. “Stav a perspektívy palivových cyklov jadrových elektrární VVER 440”. Kongresové centrum SAV, Smolenice 6.-7.09.2006 (Slovakia).