Stress corrosion cracking of CRDM penetrations

The most promising coolant additive is zinc, which as been shown to reduce the radiation activity of the primary coolant as well as increase the resistance of Alloy 600 material to PWSCC. The zinc interacts with chromium in the oxide film on the Alloy 600 components and forms a more protective (stable) oxide coating, which delays initiation of PWSCC [164].

With the addition of 20 ppb of zinc, the PWSCC initiation time for Alloy 600 reverse U-bend specimens is increased by a factor of 2.8, and, with 120 ppb of zinc, the initiation time is increased by a factor of 10 [165]. With the addition of 20 ppb of zinc and a crack-tip stress intensity in the range of 40 to 50 MPa m(36 to 45 ksi inch) the PWSCC crack growth rates are reduced by about a factor of 3.3. EPRI and the Westinghouse Owners’ Group implemented zinc addition in June 1994 at Farley Unit 2 for field demonstration. The duration of this demonstration is about 39 months [166]. The zinc is being added in the form of zinc

acetate, which has a high solubility in the PWR coolant at operating temperature. Adding zinc is expected to mitigate PWSCC in both new and old plants. However, it may take longer for zinc to be incorporated into the oxide film present in an older plant because the film is likely to be thicker and more stable.

By the end of August 2003, there were 14 plants through out the world operating with zinc addition to the primary coolant water. Table 35 identifies the 14 plants operating with zinc addition.

TABLE 35. PWR PLANTS OPERATING WITH ZINC ADDITIVE TO COOLANT WATER

Farley Unit 2 (10)*

Farley Unit 1 (16)

Diablo Canyon Unit 1 (9) Diablo Canyon Unit 2 (9) Palisades (14)

Brigham (28) Bibles A (19) Bibles B (17) Angora Unit 2 (1) Sequoia Unit 1 (12) Sequoia Unit 2 (11) Beaver Valley Unit 1 (15) Callaway (13)

Ft. Calhoun (21)

*(first cycle of zinc addition)

7.2.2. Reduced upper head temperatures

The reactor upper head temperatures can be lowered somewhat by making minor modifications to the internals of certain RPVs to increase the bypass flow. This has been tried in France, but the results were not entirely satisfactory in terms of crack growth rate to ensure 40 years of safe operation. Fig. 54 of paragraph 6.4.2 compares results of similar plants with cold or hot dome with different operational temperatures.

In Japan, the coolant temperature of the vessel head was lowered to the cold-leg temperature by increasing the bypass flow to the vessel head in 11 PWRs, which had enough margin in the primary coolant flow rate so as to increase the bypass flow to the vessel head.

7.2.3. Surface treatments

There are several different inside surface treatments being considered for mitigating Alloy 600 CRDM nozzle cracking, including special grinding, nickel plating and peening. Grinding techniques are being developed in France and Japan to remove the surface layer where cracks might have initiated, but remain undetected, and then produce compressive stresses on the regenerated surface [167]. Nickel plating can protect the treated surfaces from the PWR coolant, stop existing cracks from propagating and repair small cracks. Nickel plating has been qualified for steam generator tubes and has been applied to about 1100 tubes in Belgium and Sweden in the last 8 years. Al of these tubes, except for the first few, are still in service, whereas unplanted sister tubes are degrading [168]. The nickel plating does not provide structural strength for the CRDM nozzle. Peening with shot or other methods replaces high tensile residual stresses on the surface with compressive stresses. It has been used to prevent PWSCC initiation in steam generator tubes. However, shot peening is not effective if cracks already exist.

7.2.4. Stress improvement methods

Porowski et al. [169] have proposed a mechanical stress improvement method that redistributes the residual stresses in the nozzle and produces a layer of compressive stresses on the inside surface of the nozzle. The method consists of applying a compressive axial load at the nozzle ends, which are accessible. Analysis of the application of this method shows that the imposed axial compressive stresses interact with the residual tensile stresses on the inside surface, and the resulting plastic flow removes the residual tensile stresses from the sites on the nozzle inside surface near the partial penetration weld. The analysis results also show that the residual stresses on the inside surface are reduced, and the surface becomes near stress-free after removal of the applied axial load. This method has not been implemented on CRDM but can be considered. It was implemented only some dissimilar metal welds.

7.2.5. Alloy 600 head penetration repairs

Two options exist for the repair of Alloy 600 RPV head penetrations which contain SCCs.

The first method involves grinding out the SCC and filling the resulting cavity with a suitable weld metal. The welding process should be such that residual stresses are minimized.

Following the welding process, grinding is again performed to contour the surface of the weld repair to that of the head penetration. The weld filling material is usually Alloy 182.

The second method to repair head penetrations with stress corrosion cracks is to insert a thin liner (tube) of thermally treated Alloy 690 (TT) or austenitic stainless steel into the degraded head penetration. The head penetration in question is then pressurized and the liner will expand onto the head penetration tube and seal the crack.

EDF and FRAMATOME have developed some repair processes like grinding of the inner surface, with and without repair by welding. Repair by cutting and replacement of part of the CRDM nozzle has been studied. None of them has been used for the moment in France due to dosimetry, the number of nozzles concerned and difficulties in repairing the downhill nozzles.

7.2.6 Head penetration replacement

Head penetration replacement can take the form of either replacing the RPV closure head with a new closure head or replacing each head penetration; the new head penetration should be made from material other than Alloy 600. In several plants, where replacement of existing RPV closure heads has occurred, thermally treated Alloy 690 has been chosen as the material of construction for penetrations in replacing Alloy 600. Test results and limited field experience associated with other Alloy 690 components exposed to PWR primary coolant indicate that Alloy 690 material is not susceptible to PWSCC damage. In addition, new weld materials, Alloy 52 and 152, have been used in place of Alloy 82 and 182. The new materials have better resistance to PWSCC.

Based upon the fact that a large number of reactor pressure vessel head penetrations were exhibiting PWSCC, 29 plants in the USA have elected to replace reactor vessel heads rather than repair the damage penetrations (White el al). Table 36 presents the status as of September 2003 of reactor pressure vessel head replacement in the USA. Recognizing that Alloy 600 was susceptible to PWSCC, the replacement reactor pressure vessel heads utilized Alloy 690 instead of Alloy 600. Alloy 690 is considered to be less susceptible to PWSCC than Alloy 600. In the USA, over fifty percent of the operating PWR utilize Alloy 690 for head penetrations.

EDF has decided to replace all the reactor vessel heads of the 50 plants concerned for economical reasons: cost of inspection, cost and difficulties of repair, uncertainties on the shutdown time of plant affected by degradation. To-day 41 reactor vessel heads has been replaced with Alloy 690 penetrations. 9 others will be replaced in the next six to eight years in order to finish all the replacements before 2010.

Currently there are 23 operating PWR plants in Japan. Of these twenty-three PWR plants, 22 of the plants had RPV heads with CRDM with Alloy 600 thermal treated (TT-600) and one plant has that with Alloy 690 thermal treated 690 (TT-690). Japanese PWR utilities addressed the issue of RPV head penetration cracking in two ways. The first way is to lower the coolant temperature in the RPV head for eleven of the 22 plants. The margin in primary coolant flow rate was too small to increase the bypass flow to the RPV head in 11 plants to lower the temperature in the head. The RPV heads for these 11 plants were replaced with new heads that utilized Alloy 690 thermal treated (690-TT).

TABLE 36 (WHITE ET AL.) ANNOUNCED HEAD REPLACEMENT PLANS AS OF SEPTEMBER 2003

Status Year No. Plant

1 Davis-Besse 2 North Anna Unit 2 3 North Anna Unit 2

4 Oconee Unit 3

Already Replaced 2002

5 Surry Unit 1

6 Crystal River Unit 3 7 Ginna

8 Oconee Unit 1

9 Surry Unit 2 2003

10 TMI Unit 1

11 Oconee Unit 2

12 Farley Unit 1

13 Kewaunee 2004

14 Turkey Point Unit 3 15 Millstone Unit 2 16 Point Beach Unit 2 17 Turkey Point Unit 4 18 ANO Unit 1

19 Farley Unit 2

20 Point Beach Unit 1 2005

21 H.B. Robinson Unit 2 22 St. Lucie Unit 2 23 Beaver Valley Unit 1 24 Calvert Cliffs Unit 1 25 St. Lucie Unit 1 26 Cook Unit 1 2006

27 Fort Calhoun

28 Calvert Cliffs Unit 2 Replacing Next

Refueling Outage

2007

29 Cook Unit 2

8. REACTOR PRESSURE VESSEL AGEING MANAGEMENT PROGRAMME