ML19257B023
| ML19257B023 | |
| Person / Time | |
|---|---|
| Site: | Dresden  |
| Issue date: | 08/17/1979 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19257B022 | List: |
| References | |
| REF-GTECI-A-15, REF-GTECI-DD, TASK-A-15, TASK-OR NUDOCS 8001150114 | |
| Download: ML19257B023 (14) | |
Text
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DRESDEN STATION UNIT 1 CHEMICAL DECONTAMINATION SAFETY EVALUATION REPORT ENGINEERING BRANCH DIVISION OF OPERATING REACTORS INTRODUCTION / BACKGROUND Decontamination is required in order to reduce radioactivity levels so that improved inspection and maintenance can be performed on the Dresden Unit I nuclear facility which has been in service since 1960.
Extensive tests were conducted to evaluate the potential corrosion effects from decontamination of Commonwealth Edison Company's (CECO's) Dresden-1 Nuclear Power Station with a proprietary cleaning solution, Dow solvent NS-1.
A significant amount of this study was concerned with measuring the potential IGSCC effects of Dow NS-1 solvent on sensitized and stressed 304 stainless steel.
The corrosion research program covered several thousand individual corrosion tests of all the basic Dresden-1 BWR construction materials that will be exposed to Dow solvent NS-1 under conditions of time and temperature exceeding those proposed for the actual decontamination. The program evaluated general weight loss, corrosion by examination of pitting, crevice, and galvanic corrosion effects and stress corrosion cracking for each major metal or alloy family in the CECO system.
A corrosion trial matrix was developed that related every material to be tested with: (1) the fom of the material (wrought, welded, cast, forged etc.); (2) type of heat treatnent and condition of the surface (annealed, sensitized, hardened, precipitation hardened, scaled, descaled before exposure, etc.); (3) type of corrosion trial (general corrosion, effect of galvanic couples, crevice pitting, stress corrosion cracking, etc.); (4) environmental modifications to be considered in each trial aeration; nitrogen padding; deliberate con-tamination with heat-produced scale, metal and ions such as Cl-, F, Pb",
or 5"; use of reducing or oxidizing environment, or static or dynamic con-ditions etc.).
In this corrosion study, " worst possible case conditions" included:
Sensitization of alloys at temperature 100-150*F above those used in actual construction of the reactor facility.
Sensitization of alloys for times of up to 50 hours5.787037e-4 days <br />0.0139 hours <br />8.267196e-5 weeks <br />1.9025e-5 months <br />.
1750 168 soonso //
. Stressing the alloys after they were sens;tized, with the brittle carbide precipitates su.^ rounding the grain boundaries in place, as opposed to actual practice where components were stress relieved after they were fabricated and stressed.
Stressing the alloys to levels grossly in excess of their yield points.
Exposure of specimens to NS-1 solvent at 29F higher than expected maximum operating temperatures.
Exposure of specimens for 25-300% greater than ex-pected cleaning times.
Exposure of heavily scaled specimens whereas Dresden-1 components actually were cleaned and free of heat-treatment scale before the unit was put into operation.
Exposure of specimens to NS-1 solvent which contained contaminants at much higher con-centrations than would be present during the actual cleaning operation.
Use of configurations such as those found in double U-bend stressed specimens and severely creviced specimens, which would not be found in an opnating unit and that are particularly prone tc produce adverse corrosion effects.
Testing on this worst possible case basis was to demonstrate that the majority of the alloys at Dresden-1 will resist significant corrosion by Dow solvent NS-1.
In those few cases where specimens were retested under conditions which approached a worst possible case basis but were closer to the most probable operating case basis. These worst possible case conditions were used during the initial phase of corrosion screening.
Alloys that successfully passed such extreme corrosion tests were eliminated from the text matrix so that more time could be spent evaluating alloys which appeared susceptible to corrosion.
Negative results are difficult to prove in corrosion testing.,
N n wered questions remain even after tests of long duration:
e.g., ti,W the specimen have failed if the test had been extended for a longer p;r to( e time, or div the test environment inhibit the corrosion mechanisni.iue:
eqdy? Answer-ing these questions directly by experimentation is time cantuming, therefore 1750 169 the approach taken in this study involved testing the alloys under condi-tions known to cause failure. Once the results from these test conditions were well understood, the severity of the conditins was relaxed by a measured amount and the test was repeated. Subsequent results then were used to judge the probability that a particular corrosion mechanism would operate during the actual decontamination of Dresden-l.
For example, some 304 stainless steel specimens which were oversensitized and overstressed in relation to the alloys found in Dresden-1 underwent intergranulce stress corrosion cracking in either Dow solvent NS-1 g deionized water under sufficiently severe conditions of heat treatment, stress levels, contamination levels, and configurations. liowever, speci-mens tested under identical conditions, but with the amount of cold-worked areas reduced, showed no tendency toward such IGSCC, even when stressed to well beyond their yield points or when highly sensitized by exposure at temperatures and time periods in excess of those anticipated for the actual cleaning.
The discussion below takes into account the Dow formal 8 volume report, General Electirc supporting work, and the responses to NRC questions.
DISCUSSION A.
Decontamination Solution Dow chose a proprietary solvent called NS-l as being appropriate for dissolving BWR type corrosion products.
It is understood to consist of a chelating agent, a mild organic acid, some non-corrosive inorganic ions, and a corrosion inhibitor. The process acidity will be a pH of 3.3-4.0.
Undersirable impurities are to be held at a reasonably low concentration.
Iron concentration (resulting from dissolution) will be held below 1200 ppm, and residual solvent capacity at 5%.
The 1aboratory investigation included the effect of temperature, corrosion inhibitor concentration, nitrogen blanket, and corrosion products. The Dow NS-1 solution can be used safely at 250*F for 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />, under a nitrogen blanket.
Consideration was also given to residual effects of solvent retained in crevices. This was done by a sequence of decontamination condition followed by actual primary plant operating condition (i.e. high tempera-ture water exposure). No significantly adverse effects were observed in tests performed by General Electric Co. This result is consistent with the ?xpectation that the organic constituents in the solvent will 1750 170
s decompose into low boiling end products which will dissipate into the main coolant.
The test work was done with an inhibitor for prevention of significant corrosion.
In response to an NRC question, Dow stated that the inhibitor concentration would not be checked during the plant process because it was detemined in previous tests that absence of an inhibitor would not cause excessive corrosion. Their position is that removal of the in-hibitor "would require total reiteration of the test program."
The inhibitor does decrease corrosion of 1020 carbon steel but it does not have much effect on SA516 low alloy steel. However, the galvanic attack on SA516 in contact with a relatively large surface of Type 304SS is extensive (about 19 mils loss per decontamination).
This situation would prevail in a vessel at a weld clad defect o nly. Since the stress relieved cladding is Type 304L which proved to be immune to cracking, there is no apparent problem.
- Thus, the Dow position is acceptable.
B.
Plant Materials Plant materials were.seleeced for test on the basis of a review of component drawings. A second review with the latest drawings was made in response to questions raised at the NRC meeting on January 16-18, 1978.
One of the main items considered was the prevailing condition of vessel walls clad with Type 304SS.
The heat treatment condition on the vessel would cause this material to be sensitized.
However, according to listing most (if not all) the cladding is Type 304L (low carbon), which would sensitize only lightly if at all.
The low carbon version of the alloy is generally immune to problems of sensitization.
In any case, extensive testing was done with both the normal and low carbon alloys.
The choice of other materials and metallurgical conditions was thorough with only minor exceptions. Type 410 is extensively used as bolting, 1750 171 pins, nuts etc. within the reactor vessel and in some valves. The specimens tested appear to be in an as received condition, and it is not clear that the material was actually tested in the heat treated condition (hardened) of the item in service. This point, as well as possible use in a pressure boundary component, was checked by CECO.
Vendor records show that Type 410 stainless steel had been heat-treated to 3/4 maximum hardness. Supplementary tests were con-ducted by General Electric to determine the effects of:
1.
Bimetallic coupling to very 1arge area of AISI Type 304 Stainless Steel and also, possibly, to carbon steel or low-clloy steel.
2.
Tensile stress level approaching yield.
3.
Severe cold-working.
The primary objective of the Type 410H corrosion test was to detemine whether or not the heat-treated Type 410 stainless steel is susceptible to stress cracking, particularly of hydrogen type, under the Dow Solvent NS-1 cleaning environment.
In addition, the effects of heat-treatment, tensile stress level, galvanic effect and cold-working were also studied.
The G.E. tests results showed that:
1.
AISI Type 410H heat-treated to the vendor specifications did not crack under the anticipated NS-1 cleaning condition.
2.
Heat-treatment is the major factor in high corrosion rate of AISI Type 410H in NS-1.
3.
Galvanic effect on the corrosion rate of Type 410H is also evident, but is less important then the heat-treatment. Unstressed Type 410H of about 3/4 hardness may corrode at the rate of 300 mils per year (or 3 to 4 mils per 100 hour0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> cleaning cycle) when galvanically 1759 9 2 coupled to AISI Type 304.
If a galvanic couple to Type 405 also exists (as in the case of Dresden-1 core support system) the corrosior, rate may be reduced to about one half.
4.
The effect of stress level on the corrosion rate of Type 410H seems to be very significant. The corrosion rate of alloy 410H could be as high as 3,000 mils per year or 40 mils per 100 hour0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> cycle, provided that the stress level is as high as yield.
However, there is not a reliable record concerning the stress level on the AISI Type 410H parts in the Dresden core support system. The best esti' nates calculated from the design torque limits and the nomal engineering practices at the time of construction range from 15% to 20%
yield stress of the material. At this low stress level, the corrosion rate of 410H is expected to be less than about 15 mils per 100 hour0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> cleaning cycle. This is not considered to present a problem during the cleaning cycle or during subsequent plant operations.
Nitrided surfaces (hard face) were not tested.
Components (pumps and valves) with this isam will be inspected after decontamination. Al so,
possible malfunction during decontamination is being factored into the process (independent method of draining the solution).
C.
Type of Specimens The corrosion testing program was thorough. Various types of specimens were tested for general corrosion, stress corrosion cracking, crevice corrosion, galvanic action, weld condition etc.
Included were worst case conditions, combination of conditions (stress and crevice), and realistic circumstances like the one mentioned above about decon-tamination followed by primary plant conditions. There was also a dynamic loop test.
Stress corrosion of sensitized Type 304SS was investigated in great detail. Many types of specinans were testd, namely single U-bends, double-bends, presence and absence of heat treatment scale, corro-sion products present, crevice defects (WOL specimens), and electro-chemical techniques. The cracking influence of the decontamination solution was compared with that of deionized water (containing initial air) which is the medium of plant operation. Also tested for SCC were tensile specimens of irradiated material in a dynamic loop and galvanic coupling of carbon and low alloy steels to 304 stainless steel.
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. D.
Additional Tests Tests were performed on the effectiveness and corrosion effects with the use of the Dow dilute copper rinse following chemical cleaning.
The copper rinse is performed at 120*F of the following compositions:
p 2 (0.25% W), NH 0H (0.5%.of 30% aqua ammonia).
NS-1 (5.0% W), H 0 4
Components evaluated were: (1) copper washers used in the main re-circulation pumps; (2) silver-clad 304SS 0-rings in the reactor pressure vessel head gasket.
The results showed that light deposits of copper dissolved very rapidly.
No replating of the dissolved copper from the copper rinse occurred. The reduction in linear dimension at the edge of the copper washers will be 0.1-0.2 mils during the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> copper rinse. Since the thickness of the copper washers in the main recirculation pumps is approximately 1 1/2 inches, a loss of 0.2 mils should not present operational problems. The plating of the RPV head gasket is considered to be of pure silver, and ASTM 13413 (99.90%) grade silver corroded at about 0.01 micron per day which should not cause a leak during the rinse stage or subsequent RPV operation.
E.
COMPONENT INSPECTION PROGRAM The chemical cleaning licensing submittal committed Commonwealth Edison to a component inspection program. A Component Inspection program will be implemented specifically for the Dresden 1 chemical cleaning. This inspection program will include a baseline inspection (of selected com-ponents) before the cleaning with post cleaning inspections immediately after cleaning before (reactor) operations and at (the) two subsequent (refueling) outages. This component inspection program will be to insure that the chemical cleaning has caused no degradation (of the reactor system) and will be separate frcm the nomal Dresden 1 Inservice Inspection Program.
The component Inspection Program will be based on the requirements of the ASME Boiler and Pressure vessel code,Section XI, but will only irclude those categories that are relevant to the chemical cleaning process. The acceptance criteria and component selection will remain constant, and 1750 174
. will be based on the code in effect at the time of the first inspection.
The conditional approval from the NRC dated December 9,1975, for the Chemical Cleaning required that:
"A pre-service inspection program for the primary coolant boundary will be formulated and submitted to our review and approval prior to returning the reactor to service."
This inspection program is to be distinct from the required ASME Section XI in-service inspection program, however, components that are common both programs will be included.
The program is divided into four time periods for effects on materials:
1.
Pre-Cl eaning A.
Component inspection Meridonal Head Weld, at 7* Loc.,1 Ft.
Circumferential Head Weld, From 186* to 206,
1 Ft.
Vessel-To-F1ange Weld, From Bolt #18 to Bolt Hole #24, 3.5 Ft.
Unloading line system welds Pipe welds:
Recirculation suction and return lines, rises, downcomers (cross-tie), recircu-lation bypass loop, clean-up supply and return, vessel drain, poison system.
B.
Metal Surveillance Program The need to know the effects of the chemical cleaning on the materials of construction during the cleaning and subsequent reactor operation will be addressed by this program.
The conditional approval of the cleaning required that:
"A post-cleaning surveillance program which includes additional (metal) sur-1750 175 veillance samples and a specimen withdrawal and examination schedule will be submitted for review and approval prior to returning the reactor to service."
There will be five sets of samples prepared.
Each set will be identical except for the 410 SS samples.
Each set consists of:
1.
Pressurized Tubes at 100% of 550*F Yield Strength (6 specim' ens per material)
A.
304 stainless steel - furnace sensitized - as welded.
B.
304L stainless steel - furance sensitized - as welded.
C.
Inconel 600 - furance sensitized - as welded.
Incoloh 800 - furance sensitized - as welded.
D.
E.
Zircoloy 2 - as welded.
2.
Bent Beam (6 specimens)
A.
410 stainless steel - tempered to 3/4 hard.
3.
Flat - weight loss coupons (3 specimens per material)
A.
304 stainless (sensitized at 1200 F for 50 hours5.787037e-4 days <br />0.0139 hours <br />8.267196e-5 weeks <br />1.9025e-5 months <br />).
B.
304L stainless (sensitized)
C.
316 stainless (sensitize 6' O.
347 stainless.
E.
Inconel 600 (sensitized)
F.
Incoloy 800.
G.
Zircalloy 2.
H.
405 stainless.
I.
410 stainless.
J.
Monel 400.
K.
Mastelloy B 1750 176
- The five sets will be used as follows:
Set 1.
Chemical cleaning set - to be removed after cleaning.
410 SS bent beam sample will be stressed to 100% 550*F yield strength.
Set 2.
Long tem set in reactor pressure vessel. (RPV).
In during cleaning and remains in for reactor operation.
410 SS bent beam sample will be unstressed during cleaning and then stressed to 100% 550*F Y. S. for reactor operation.
Set 3.
Long tem control set in RPV.
Placed in for reactor operations.
410 SS bent beam sample to be stressed to 100% 500 F Y. S.
for reactor operations.
Set 4.
Long tem test loop set.
Inserted in RPV for cleaning, then transferred to test loop for reactor operation.
If test loop does not operate during reactor operation then set will remain in RPV.
410 SS bent beam to be unstressed during cleaning, then stressed to 1003 550 F Y. S. for operations.
Set 5.
Long tem test loop control.
Inserted in test loop for reactor operations (or in RPV if test loop not operational.
410 SS bent beam to be stressed to 100% 500 F Y. S. for operations 2.
Post Cleaning A.
Component inspection same as pre-cleaning, in addition to: (1) nozzle-to-vessel welds as well as dissimilar metal welds:
reactor inlet / outlet, steam drum riser /downcomer, and safe ends; (2) one reactor coolant recirculation pump will be dis-assembled; and (3) two valves will be disassembled.
B.
Metal Surveillance Sets 1, 4 will be removed from turning vane.
Set 3 will be inserted in turning vane.
Sets 4, 5 will be inserted in test vessel "A" of corrosion test loop (if test loop is not operable, then sets 4, 5 will be in turning vane).
C.
CECO. will attenpt to inspect the following welds from inside the vessel by remote UT:
beltline region and upper vessel shell welds, and upper nozzle-to-vessel and safe end welds (including inner radius).
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. Beltline region and vessel shell welds are inaccessible from the vessel 0.D.
Nozzle welds are accessible from the vessel 0.D.
Bottom head and lower shell weld accessibility will be determined after cleaning.
Inspectability of all welds unknown at this time.
D.
Determine accessibility and inspectability of safe end welds on steam drum and steam generators after chemical cleaning.
CECO. will remove the insulation collars which presently prohibit access to the weld for performing inspections.
After removal of the insulation collars, the inspectability of the welds will be determined.
CECO.'s present plans are to meet the section XI code requirements for inspection, i.e., dye penetrant and ultrasonic of the welds.
F.
Evaluation As can be visualized from the large and diverse number of materials, specimens, and test conditions examined, the data are extensive.
However, for the purpose of overall judgement, they can be broken down into two or three general categories, namely general corrosion, cracking, and in a subsidiary group local corrosion (crevice, pitting etc.).
The laboratory data indicate that metal loss (general corrosion) resulting from the process will be of the order of magnitude of several hundredths of a mil for series 300SS, alloys of copper, or nickel and some speciality alloys; tenths of a mil for the 400 series is especially wide (extending into both of the other categories). The galvanic corrosion of the low alloy steels is some four times higher when it is coupled with a 300 series steel with a very high ratio of the latter to the former.
All but this last condition can be dismissed as being negligible metal losses. The extreme galvanic condition would prevail only under the circumstances where there would be a penetration (crack) of the Type 304L cladding to expose a very small surface of the low alloy 1750 178
vessel wall. Although the cladding is expected to withstand the NS-1 solution. Thus, one may conclude that the expected metal loss (general corrosion) is acceptable.
The extensive testing of stressed specimens of Type 304SS confirmed the fact that this material in a strongly sensitized condition will undergo intergranular cracking in some environments when stressed and when a specific local condition obtains on the surface. The local condition may be scale which fractures under stress, or a variable solvent condition which pemits non-unifom passivation, or crevices fomed by double U-bends, or some similar accomodating geometric condition. At some pipe welds, there is a combination of high stress points on a locally sensitized region in a variable oxygen enviroment.
The presentation of the data in the Dow report is very confusing. It is apparent, however, that the negative results (no cracking) are associated with what might be called clean and uniform conditions on the surfaces (low strain rate tests, or scaled surfaces with relatively low strain (G.E. beam tests) and most of the Dow experiments with single U-bends. On the other hand, the test data for the Dow solvent shows cracking in double U-bends with heat treatment scales as well as ferric (and nickel) ion with and without scales. Single U-bends, however, cracked only when heat treatment scale was not removed (higher strain than the G.E. beams).
Similar data are shown for deionized water (initially saturated with air). The data comparison serves as the basis for the conclusion that the decontamination solution is no more aggressive with respect to cracking of sensitized Type 304 than is of the specially designed specimens under the worst case circumstances discussed above which are more severe than actual conditions during the decontamination process The statement in the abstract of the Dow report that cracking did not occur in the specimens which were sensitized after bending is consistent with the above discussions. This particular sequence of treatment probably minimizes stress, and surface irregularities sufficiently to prevent attack at a specific point. Al so, the heat treatment on a cold worked specimen results in a different metallur-gical condition (grain structure).
In any case, the result applies favorably to the vessel wall cladding situation.
Furthemore, the vessel cladding is Type 304L (low carbon, and therefore minimal sensitization). Thus it is a reasonably sound conclusion that the vessel cladding will not be adversely affected.
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The pipe welds must be considered in a somewhat different. light.
The possible sensitization condition (and local stresses) is unknown.
However, again one can accept the logic of the conclusion that the propensity towards cracking in the decontamination solution is no worse than in primary plant water.
Although the Dow NS-1 solution will dissolve the copper constituents in crud deposits, concern had been expressed that some of the copper will redeposit on exposed carbon and low alloy steel.
Dow demonstrated that a rinse with the NS-1 solution raised to a slightly basic pH with hydrogen perioxide would easily dissolve the copper. Concern was then expressed over possible stress corrosion cracking. This question was resolved by G.E. in a test program with bent beams of sensitized Type 304 SS in a crevice arrangement wtih carbon and low alloy steels. No evidence of cracking was found.
CONCLUSION We conclude that the plant materials will not be significantly damaged by the decontamination solution, either during the process itself or upon immediately returning to normal power operation. There are sufficient stress corrosion data on most of the materials, and on the rates of crack penetration on specimens simulating defects in the vessel caldding.
All of these data were obtained under conditons that assumed return of the unit to operation shortly after the decontamination operation. Cad ( r these conditions the bulk of the residual NS-1 solvent (trapped in cro ices) will decompose into harmless constituents with no additional corrosion occurring during the post-cleaning operation.
However. Dresden Unit 1 will not return to service until June,1980, ten months after the August,1979, decontamination operation. Thus, there is sone concern of the consequences of leaving residual NS-1 solvent at wet layup conditions for this period of ten months. The type 410 stainless steel is used for pins, screws, and bolts in the core support structure where there would be crevices that may not be adequately rinsed of copper deposits after the copper rinse procedure.
The 410 SS showed more than normal corrosion rates when stressed to 85% yield in the NS-1 solution. Since, little is known of thermal stresses on bolting stresses or residual, the 410 SS components may be at higher stress than the calculated design torque limits of 15-20% yield in the crevice areas (bolts etc.) and susceptible to IGSCC.
The fracture mechanics tests on simulated clad specimens with cracks showed that the NS-1 solution can by galvanic corrosion cause undercutting of the vessel clad in the vicinity of the crack. There is the possibility that the NS-1 solvent or other potentially harmful impurities may be entrapped in the cavity.
With air in water during the 10 month wet layup period, local galvanic cells 1750 180
. could cause corrosion to continue. Also, there is the possibility of crack extension underneath the cladding defects (if they exist) on the reactor vessel and should be carefully evaluated by fracture mechanics personnel within NRC.
The concern for stressed 410 SS components of the core support structure can be resolved by requiring that the license include in his Metal Sur-veillance Program type 410 SS surveillance specimens that are stressed heavily and contain crevices. The licensee's proposed component inspec-tion program is acceptable, provide the additional surveillnce specimens are includedin the program.
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