Component integrity

ISSUE TITLE: Fuel channel integrity and effect on core internals (CI 1) ISSUE CLARIFICATION:

Description of issue

This issue is specific to NPPs with PHWR.

The coolant channels in CANDU reactors increase in length and diameter with the passage of time, due to creep, induced by temperature and irradiation. This requires periodic adjustments, which is a cumbersome operation.

Furthermore, several mechanisms have been identified as contributors to fuel channel degradation, which could lead to fuel channel failure. These mechanisms are the following:

(a) deuterium ingress into, and embrittlement of, the pressure tube (PT) material, (b) rolled-joint cracking,

(c) fretting,

(d) formation of blisters,

(e) PT deformation, e.g. elongation and sagging, and

(f) material property changes, e.g. changes in material tensile properties, fracture toughness, and delayed hydride cracking velocity.

Finally, there are other concerns related to fuel channel integrity due to ageing and accidents particularly during loss-of-coolant or loss-of-heat sink (e.g. moderator) events. These are the following:

1. The fuel channel design life has not yet been demonstrated.

2. PT local creep rupture during the high pressure part of the large LOCA blowdown due to:

(a) creep and annealing of irradiated PT, (b) thermal hydraulic temperature gradients,

(c) temperature gradients from bearing pad contact with PT, and (d) temperature gradients from contact of fuel with PT, hence:

i) more rapid ballooning at fuel bearing pad plane causing fuel element/PT contact, and

ii) fuel element sagging, embrittling, or appendage braze melting causing contact between fuel element and PT.

3. Inadequate subcooling on calandria tube to prevent dryout and rupture upon initial contact between ballooned PT and calandria tube (CT):

(a) contact conductance of irradiated pressure tubes and calandria tubes, (b) local moderator flows and temperature gradients, and

(c) effect of fuel element to PT contact on subcooling requirement.

4. Miscellaneous:

(a) axial creep and rupture of PT in the hot ring beside the garter spring, (b) heating and pullout of PT rolled joint,

(c) feeder rupture

(d) pullout of CT rolled joint,

(e) constrained axial expansion of fuel, (f) moderator system reliability, (g) local CT dryout beneath sagged PT,

(h) delayed ECCS repressurisation and PT ballooning, and (i) transport of embrittled fuel from core.

Safety significance

1. Although the CANDU reactor design has the capability to withstand the consequences of a pressure tube rupture, designers and operators must strive to reduce the probability of pressure tube failure. Fuel channel failure consequences are severe, particularly when taking into consideration the potential for damage to other channels and/or core internals.

2. Radiation doses associated with fuel channel replacement adjustments are high.

3. Fuel channel replacement entails a very high economic as well as radiation dose penalty.

Source of issue (check as appropriate)

• ____xx____ operational experience

• ____xx____ deviation from current standards and practices

• ____xx____ potential weakness identified by deterministic or probabilistic (PSA) analyses

MEASURES TAKEN BY MEMBER STATES:

Argentina

CNA-I reactor pressure vessel surveillance program was initiated in 1974, by introducing 30 test specimens distributed in the lower reflector, inside the moderator tank. In 1980, 10 additional capsules containing samples of the same type, but made of A508 class 3 material were introduced, aiming at obtaining a relative reference for neutron exposure. On the other hand, irradiated capsules were withdrawn and examined. No definite conclusion could be obtained about the pressure vessel material behaviour, due to the differences between the neutron spectra present on samples located inside the moderator tank and those spectra on the pressure vessel wall.

Due to the uncertainties associated with the previous results, Siemens - Kraftwerk Union AG carried out an irradiation program in the German VAK reactor in 1985. The experiments were carried out with probes made of the same pressure vessel basis material as well as with A508 class 3 material, simulating the conditions on the CNA-I pressure vessel wall.

Although there are some uncertainties due to the short irradiation time (high acceleration factor), irradiation carried out by Siemens - Kraftwerk Union AG indicated that 35ºC was the temperature for ductile-to-brittle transition at the end of CNA-I lifetime (32 full power years). On the other hand, preliminary results of the most unfavourable LOCA analysis showed that the pressure vessel would always be at a temperature higher than 35ºC at the end of its lifetime.

The Regulatory Body has carried out an assessment of the available information related to the reactor pressure vessel. It was concluded that there are uncertainties regarding the reactor pressure vessel integrity under certain accidental operational situations.

On the other hand, NASA has initiated an evaluation of those accidental scenarios having the most unfavourable stresses for the pressure vessel integrity (Pressurised Thermal Shock analysis), as well as a program of the necessary actions (i.e. improvement in the surveillance program), to be carried out in order to minimise such uncertainties contained in the studies.

In addition, it is important to emphasise that the design characteristics of the plant do not facilitate an optimal mixture of the injection of the emergency system water, due to the fact that the loop seal

always contains cold water, which has a negative contribution to the effects of a thermal shock during the eventual occurrence of an accidental situation.

Therefore, with the purpose of ensuring that the reactor pressure vessel will continue preserving the appropriate safety margin, it has been required that the necessary measures should be taken in order to reach the year 2001 having heated the water contained in the high pressure accumulators of the core emergency cooling system.

The CNA 1 reactor cooling channels are degraded in way which could not be predicted in the original design. The above mentioned degradation consists fundamentally of both cooling channels external cover foils degradation and cooling channels sagging. Besides, due to the fact that to reduce the cooling channels corrosion and erosion effects, these components were covered with a non ferrous alloy containing 60% cobalt, commercially known as Stellite-6, so as a result of the material activation produced by the core neutronic flux, cobalt-60 is formed contributing additionally to the radiation field in certain places of the reactor building and consequently, to the increase of occupational doses. As a result of this, CNA-I normalised occupational dose is higher than in those nuclear power plants where Stellite-6 is not used.

Due to above mentioned Responsible Organisation was required to replace the all the original coolant channels by others without Stellite-6. This task is gradually going on at each programmed outage and up to now 152 channels have been replaced, representing more than 60 % of their total number.

The CANDU reactors cooling channels (pressure tubes) are degraded in way which could not be predicted in the original design. It is known that the degradation mechanisms are: hydrogen uptake, blister formation, rolled joint cracking, sag/deformation and fuel-bundle bearing-pad fretting. The Regulatory Authority issued a requirement regarding the degradation mechanisms and ageing concerns for Embalse NPP pressures tubes. Pressure tubes become susceptible to the formation of the hydride blisters when the hydrogen equivalent concentration exceeds a threshold called “blister formation threshold” (BFT).

The licensee was required to demonstrate before the end of 1998 that hydrogen equivalent concentration for all the pressure tubes is below the BFT based on the specific assessment model for Embalse NPP. The concentration and the uptake rate of equivalent hydrogen must be obtained from the scraping results. In addition, an appropriate sequence for garter springs repositioning (SLARing) must be defined according to the above mentioned results. To verify that the degradation is kept within admissible values and to take the corresponding remedial actions, in service inspections and tests is carried out. For that reason, an in service inspection programme was implemented and put to practice in 1986 with the objective of repositioning all garter-springs and inspection all pressure tubes.

It is important to highlight that except for failures in A14 and L12 pressure tubes caused by anomalies in the LIM RAM tool used for the repositioning of the garter springs during 1995 Embalse NPP programmed shutdown (see significant events), no other anomaly in relation with the pressure tubes was recorded.

The criteria to select the different pressure tubes that were inspected had their basis on: 1) avoiding the operation of the power station when there was contact between pressure tubes and calandria tubes and with pressure tubes predicting level of hydrogen equivalent or superior to the one during the initiation of the zirconium hydride blister formation process (blister formation threshold), and 2) considering the deformation of the pressure tubes which might prevent the use of the repositioning tool or the movement of the garter springs.

Besides, as a consequence of the failure in the A14 and L12 pressure tubes, a programme for the assessment of the material extracted from such tubes has been implemented in order to determine the

level of hydrogen equivalent and the deuterium intake velocity. Apart from this, the future possibility to implement sample scraping in some pressure tubes is being analysed.

The selection criteria for the pressure tubes inspected during the last planned outage were: Priority list proposed by AECL Memo: “Embalse spacer repositioning proposal” - E. G. Price - Feb. 94, and Changes due to new predictive hydrogen intake model taking into account the operating experience In June 1998 AECL completed a pressure tube blister susceptibility assessment to evaluate the potential for hydride blister formation in the pressure tubes (PT) of Embalse NPP (CNE). Several deuterium ingress cases were analyzed using deuterium concentration data from other CANDU 6 reactors, concluding that for each case PTs could potentially reach blister threshold formation (BTF) before the CNE October / November 1998 planned outage. Therefore, were required to complete a long term assessment of hydride blister susceptibility by monitoring deuterium uptake in PTs at Embalse.

Based on the above mentioned assessment, to avoid PT / calandria tube (CT) contact, had been improved the garter springs repositioning program schedule on going during planned outages.

Regarding the above mentioned considerations, the Nuclear Regulatory Authority (ARN) required to utility to constrain CNE operation until fulfil the following:

(1) Demonstrate that there are not PTs containing H/D over BTF using a specific CNE model to determine deuterium content and uptake. The model had been obtained from scrapping results performed during last planned outage. Besides, until fulfils its requirements into the next three months, additional measures during operation for PT failures early detection were recommended.

(2) Submit a new garter springs repositioning program for remanding PTs, according to model results

The utility fulfils the above mentioned requirement as follow:

During the October 1998 CNE planned outage had been implemented the in-core scrape sampling program, which involved a total of ten PTs that were sampled after about 109,600 EFPH equivalent full power hours (EFPH) or 119,400 hot hours. The purpose of PT scrape sampling was to monitor the uptake of deuterium. The samples were selected by the licensee from a list of recommended candidate channels issued by AECL to provide an indication of the variation in deuterium concentration along the length of the PTs.

Scrape sampling was performed using the wet scrape tooling in which a separate tool is delivered to each sampling location in sequence by the fuelling machines and cutting is performed fully immersed in the PHTS coolant under both temperature and flow corresponding to shutdown condition.

The sample analyses were performed by using two techniques. The non destructive measurement of the temperature corresponding to the Terminal Solid Solubility for hydride Dissolution (TSSD) using Differential Scanning Calorimetry (DSC) and Hot Vacuum Extraction – Mass Spectrometry (HVEMS). DSC is normally only performed to provide consistency check on the HVEMS results.

The assessment of CNE scrape sampling results, performed in October 1998, indicated lower deuterium uptake rates in comparison with previous assessment performed using deuterium concentration data from other CANDU 6 reactors. There is a difference in deuterium uptake behaviour between the two PHTS flow loops (higher in loop 1 than loop 2). The differences still are being investigated.

The assessment of CNE scrape sampling results let to perform studies with CNE specifics data determining that until reactor reach 160,800 EFPH, that means approximately October 2004, conditions to blisters formation would not be given.

Canada

Activities by the regulatory body:

(a) To address deuterium ingress, licensees are required to:

− justify operation with Hydrogen Equivalent (Heq) concentrations predicted to the end of the license period,

− periodically monitor pressure tube hydrogen levels in situ by scrape sampling, and

− periodically remove and examine surveillance tubes for deuterium and hydrogen concentration.

(b) To address rolled-joint cracking, licensees are required to:

− establish shutdown procedures that avoid fast fracture,

− establish a Leak Detection System that is active at all times during operation,

− monitor hydriding by period removal and surveillance examination of rolled joints, and

− monitor crack detection through in-service inspection (ISI).

(c) To address fretting, licensees are required to:

− determine the population of at risk by in-service inspection program,

− limit the heatup/cooldown cycles and operating times based on the fret population, and

− limit operation if the probability of initiating a crack is high.

(d) To address blister formation, licensees are required to:

− ensure that reactors are not operated with a PT which has a detected blister in, or which is in contact with, its CT, and which meets or exceeds the current threshold criterion for hydrogen concentration for blister formation under continuous operation; this is achieved by maintenance programs that consist of:

− prediction and monitoring of Heq concentrations,

− predictions and monitoring of PT-to-CT contact, and

− removal of contact by repositioning garter springs before the blister formation threshold is reached.

(e) To address PT deformation, the licensees are required to:

− measure deformation through a comprehensive in-service inspection program,

− shift fuel channels to keep them on their bearings as they elongate,

− reposition garter springs to keep sagged tubes out-of-contact, and

− replace PTs and CTs that have sagged to the point where they are difficult to refuel.

(f) To address material property changes, licensees are required to:

− remove PTs for periodic testing from CANDU lead units, and

− carry out tests to determine material tensile properties, fracture toughness, and delayed hydride cracking velocity.

Activities by the industry:

The Canadian nuclear industry has identified and is acting upon the following key elements of a

“Pressure tube aging management program”:

(a) understanding pressure tube aging (material properties, operating conditions, aging mechanisms, condition indicators, consequences of aging-related degradation and failures, operating experience, research and development),

(b) definition of an aging management program (coordinating activities, documentation, program optimization),

(c) managing aging mechanisms (following procedures, chemistry control of water and annulus gas),

(d) inspection, monitoring and assessment (leak rates, fitness for service assessment), and (e) maintenance/replacement (mitigation of tube degradation, replacement).

China

Pressure Tube (PT) integrity issue has been identified by NNSA during TQNPP PSAR review. It is required by NNSA that the designer should demonstrate the efficient measures have been taken to guarantee the integrity of PT. The measures have been reviewed by NNSA.

According operation experience of Canada, CANDU fuel channels have degraded in different ways.

The problems associated with them are: rolled joint cracking, hydrogen uptake (ingress), formation of blisters, and deformation.

In TQNPP PSAR review, NNSA require that the licensee set up measures to avoid fast fracture due to rolled-joint cracking. There is an Annulus Gas System (AGS) ,which is a leak detection system ,that is active at all time. It can detect any leakage from pressure tubes to meet Leak Before Break (LBB) requirement. In AGS, dry CO2 gas is supplied to the annuli between the pressure tubes and calandria tubes. The dewpoint and rate of change of dewpoint for the recirculated gas is continuously monitored. Sampling and analyzing the gas for moisture contents provides a means for leak detection from pressure tube. The extent of hydriding can be monitored by periodic removal of rolled joints.

In the short term, hydrogen uptake could influence the formation of blisters. We do not expect hydrogen uptake to reach levels high enough to cause embrittlement during operation. NNSA require that the licensee set up measures to ensure that pressure tubes remain ductile.

In CANDU reactors, garter springs keep pressure tubes from contacting their calandria tubes. In early design of CANDU, the garter springs can move during operation. The movement of garter springs allowing the pressure tubes to sag, contact the calandria tubes, and form blisters can crack and eventually fail the tube. To maintain some element of the defense-in-depth approach, NNSA require that the licensee set up measures to avoid the contact of pressure tubes and calandria tubes. In TQNPP, the tight garter springs will be used to avoid its moving.

According to operation experience of CANDU, there are 3 types of deformation of pressure tubes:

• diametrical creep and wall thinning;

• channel sagging; and

• axial elongation.

Although a lot of researches have been done in Canada, there are still many uncertainties associated with the long-term consequences of the above degradation mechanisms. NNSA require that the licensee to increase the level of management by inspecting and testing during the operation in order to pressure tube integrity.

The material (Zr-2.5Nb) of PTs used in TQNPP was purchased from Russia. It is first time that the Russian material is used in CANDU. NNSA required the utility perform the radiation performance test of TQNPP PT material. The testing results should be provided to NNSA for review.

It’s required that the hydrogen content of the material should be strictly controlled during manufacture stage.

India

India is operating three reactors with Zircaloy-2 pressure tubes and ten reactors with Zirconium- 2.5%

Niobium Pressure tubes. Reactors under construction are Zr-2.5% Nb pressure tube with four tight fitting garter springs.

First Generation Reactors of RAPS-1 & 2 and MAPS-1&2 have open annulus between calandria tube and pressure tube with air flowing through the annulus. The reactor RAPS-1, RAPS-2 (before retubing and MAPS-1 & 2 have two loose fit spacers which separate the pressure tube from calandria tube. NAPS 1 & 2 and KAPS-1 have four loose fit spacers whereas from KAPS-2 onwards (including retubed RAPS-2) have four tightfit spacers. From NAPS-1 onwards all reactors have close annulus where dry CO2 is purged continuously to help detection of any leakage through the pressure tube.

RAPS # 2 has an unique combination of zirconium - niobium pressure tubes with an open annulus and thus faced some unique challenges.

The continuous purge of CO₂ also helps in avoiding accumulation of Deuterium/Hydrogen in the annulus.

History of operation:

The first PHWR commissioned in India was RAPS-1 which was made critical in 1973. This reactor operated at low power for substantial period of time due to crack in end shield and therefore has logged in only 7.2 EFPYs till now. RAPS unit-2 was taken for en-masse coolant channel replacement in 1994 at 8.5 EFPYs & has operated since 6/6/1998 after en-masse coolant channel replacement for 5.85 EFPY.

MAPS-1 operated successfully for 10.1 EFPY till 20/08/2003 and its en-masse coolant channel replacement program is in progress. MAPS-2 has been operating since 23-07-2003 after coolant channel replacement and has completed 1.68 EFPY so far.

NAPS-1 has been taken for coolant channel rehabilitation job after completion of 9.69 EFPY &

NAPS-2 has completed 9.35 EFPY so far and KAPS-1&2 are operating after 8.72 and 8.52 EFPYs respectively. Kaiga-1 & 2 are operating after completion of 3.75 & 4.36 EFPYs.

RAPS-3 & 4 have completed 4.34 EFPY & 3.85 EFPY respectively as on 30.09.05.

RAPS-3 & 4 have completed 4.34 EFPY & 3.85 EFPY respectively as on 30.09.05.

Dans le document Generic Safety Issues for Nuclear Power Plants with Pressurized Heavy Water Reactors and Measures for their Resolution | IAEA (Page 49-75)