JPS6150143B2 - - Google Patents
Info
- Publication number
- JPS6150143B2 JPS6150143B2 JP56172742A JP17274281A JPS6150143B2 JP S6150143 B2 JPS6150143 B2 JP S6150143B2 JP 56172742 A JP56172742 A JP 56172742A JP 17274281 A JP17274281 A JP 17274281A JP S6150143 B2 JPS6150143 B2 JP S6150143B2
- Authority
- JP
- Japan
- Prior art keywords
- treatment
- cold working
- nuclear reactor
- scc
- aging treatment
- Prior art date
- Legal status (The legal status is an assumption and is not a legal conclusion. Google has not performed a legal analysis and makes no representation as to the accuracy of the status listed.)
- Expired
Links
- 230000032683 aging Effects 0.000 claims description 25
- 239000000243 solution Substances 0.000 claims description 13
- 229910045601 alloy Inorganic materials 0.000 claims description 8
- 239000000956 alloy Substances 0.000 claims description 8
- 238000004519 manufacturing process Methods 0.000 claims description 6
- 230000007797 corrosion Effects 0.000 claims description 5
- 238000005260 corrosion Methods 0.000 claims description 5
- 238000005336 cracking Methods 0.000 claims description 5
- 230000035882 stress Effects 0.000 claims description 5
- 238000000034 method Methods 0.000 claims description 4
- 239000006104 solid solution Substances 0.000 claims description 3
- 238000005482 strain hardening Methods 0.000 description 25
- 238000001816 cooling Methods 0.000 description 5
- 239000000463 material Substances 0.000 description 4
- XLYOFNOQVPJJNP-UHFFFAOYSA-N water Substances O XLYOFNOQVPJJNP-UHFFFAOYSA-N 0.000 description 4
- 229910001090 inconels X-750 Inorganic materials 0.000 description 3
- OKTJSMMVPCPJKN-UHFFFAOYSA-N Carbon Chemical compound [C] OKTJSMMVPCPJKN-UHFFFAOYSA-N 0.000 description 2
- 229910002804 graphite Inorganic materials 0.000 description 2
- 239000010439 graphite Substances 0.000 description 2
- 229910000816 inconels 718 Inorganic materials 0.000 description 2
- 238000002844 melting Methods 0.000 description 2
- 230000008018 melting Effects 0.000 description 2
- 229910000990 Ni alloy Inorganic materials 0.000 description 1
- QVGXLLKOCUKJST-UHFFFAOYSA-N atomic oxygen Chemical compound [O] QVGXLLKOCUKJST-UHFFFAOYSA-N 0.000 description 1
- 229910052799 carbon Inorganic materials 0.000 description 1
- 230000007423 decrease Effects 0.000 description 1
- 230000001627 detrimental effect Effects 0.000 description 1
- 230000000694 effects Effects 0.000 description 1
- 238000010438 heat treatment Methods 0.000 description 1
- 229910001026 inconel Inorganic materials 0.000 description 1
- 230000007246 mechanism Effects 0.000 description 1
- 229910052760 oxygen Inorganic materials 0.000 description 1
- 239000001301 oxygen Substances 0.000 description 1
- 239000000126 substance Substances 0.000 description 1
- 210000002268 wool Anatomy 0.000 description 1
Classifications
-
- Y—GENERAL TAGGING OF NEW TECHNOLOGICAL DEVELOPMENTS; GENERAL TAGGING OF CROSS-SECTIONAL TECHNOLOGIES SPANNING OVER SEVERAL SECTIONS OF THE IPC; TECHNICAL SUBJECTS COVERED BY FORMER USPC CROSS-REFERENCE ART COLLECTIONS [XRACs] AND DIGESTS
- Y02—TECHNOLOGIES OR APPLICATIONS FOR MITIGATION OR ADAPTATION AGAINST CLIMATE CHANGE
- Y02E—REDUCTION OF GREENHOUSE GAS [GHG] EMISSIONS, RELATED TO ENERGY GENERATION, TRANSMISSION OR DISTRIBUTION
- Y02E30/00—Energy generation of nuclear origin
- Y02E30/30—Nuclear fission reactors
Description
本発明は新規な原子炉用ばねの製作法に係り、
特に耐応力腐食割れ（SCC）性の向上に好適な
原子炉用ばねの製作法に関する。
第１図及び第２図に示される従来の原子炉用ば
ね３及び６は主に溶製あるいは溶製後に固溶化処
理を施した後、成形を含む冷間加工を行ない、次
いで時効処理を施して製作されていた。これらは
冷間加工＋時効処理により原子炉用ばね材として
要求される高温強度ならびにばね特性を確保する
ための処理製作法であるが、耐SCC耐に関して
必らずしも最適化されていない。
本発明の目的は原子炉用ばねとして要求される
高温強度及びばね特性を確保すると共に耐SCC
性がより優れた原子炉用ばねの製作法を提供する
ことにある。
原子炉用ばね（例えばBWR制御棒駆動機構の
ストツプピストンシール用イクスパンシヨンスプ
リングやドライブピストンシール用イクスパンシ
ヨンスプリング）は、高応力作用下で隙間が形成
された箇所で使用される場合が多く、そのため隙
間SCC対策が重要課題となつている。現在、原
子炉用ばねは強度と腐食性に優れたインコネル
X750合金を用い、固溶化処理後に冷間塑性加工
を施しその後に時効処理（直接時効あるいは二段
時効処理）する工程で製作されている。固溶化処
理後の冷間とそれに引き続く時効処理により原子
炉用ばね材としてのばね特性及び高温強度向上に
寄与するものである。しかし、発明者らは固溶化
処理と時効処理との中間に30％〜40％未満の冷間
加工を施すと逆に耐応力腐食割れ性を低めること
を見いだした。すなわち、固溶化処理と時効処理
との中間に実施する冷間加工度を０％〜60％に変
え、耐隙間SCC性に及ぼす冷間加工度の影響を
高温高圧純水中隙間付定歪試験により検討した。
また同時に時効処理条件（直接時効及び二段時
効）についても同様の検討を加えた。その結果、
次の様な新たな事実を発見した。固溶化処理と
直接時効処理との中間に30％以下（10〜30％）の
冷間加工を施した場合には隙間SCC感受性が認
められる。固溶化処理と二段熱処理との中間に
20％以下（10〜20％）の冷間加工を施した場合、
隙間SCC感受性が認められるが、30％以上にな
ると隙間SCC感受性は著しく減少する。固溶
化処理と時効処理（直接時効処理と二段時効処理
を含む）との中間に60％程度の冷間加工を施すと
耐隙間SCC性が、極めて良好となる。
以下、本発明の一実施例を説明する。素材はイ
ンコネルX750合金である。その主な化学成分は
72.92％Ni，15.48％Cr，6.91％Fe，0.57％Ａ，
2.60％Ti，0.95％Nb＋Ta，0.04％Ｃである。表
は高温高圧純水中隙間付定歪試験結果を示す。試
験条件は次の通りである。試験温度：288℃、圧
力：86Kg／cm2、溶存酸素：8ppm、隙間形成材：
グラフアイトウール、ひずみ：約1.0％、試験時
間：500h。
The present invention relates to a novel method for manufacturing a spring for a nuclear reactor,
In particular, it relates to a manufacturing method for nuclear reactor springs suitable for improving stress corrosion cracking (SCC) resistance. Conventional nuclear reactor springs 3 and 6 shown in FIGS. 1 and 2 are mainly produced by melting or solution treatment after melting, followed by cold working including forming, and then aging treatment. It was manufactured by These manufacturing methods use cold working and aging to ensure the high-temperature strength and spring properties required for spring materials for nuclear reactors, but they are not necessarily optimized in terms of SCC resistance. The purpose of the present invention is to ensure high temperature strength and spring characteristics required for nuclear reactor springs, and to provide SCC resistance.
An object of the present invention is to provide a method for manufacturing a spring for a nuclear reactor with superior properties. Nuclear reactor springs (e.g. expansion springs for stop piston seals in BWR control rod drive mechanisms and expansion springs for drive piston seals) are sometimes used in locations where gaps are formed under high stress conditions. Therefore, countermeasures against gap SCC have become an important issue. At present, springs for nuclear reactors are made of Inconel, which has excellent strength and corrosion resistance.
It is manufactured using X750 alloy through a process of solid solution treatment, cold plastic working, and then aging treatment (direct aging or two-stage aging treatment). Cold treatment after solution treatment and subsequent aging treatment contribute to improving spring properties and high-temperature strength as a spring material for nuclear reactors. However, the inventors have found that performing cold working of less than 30% to 40% between the solution treatment and the aging treatment conversely lowers the stress corrosion cracking resistance. In other words, the degree of cold working carried out between solution treatment and aging treatment was varied from 0% to 60%, and the influence of the degree of cold working on clearance SCC resistance was examined by a constant strain test with a gap in high-temperature, high-pressure pure water. This study was conducted by
At the same time, a similar study was conducted regarding aging treatment conditions (direct aging and two-stage aging). the result,
We discovered the following new facts. Crevice SCC susceptibility is observed when cold working of 30% or less (10 to 30%) is performed between solution treatment and direct aging treatment. Between solution treatment and two-stage heat treatment
When subjected to cold working of 20% or less (10-20%),
Although interstitial SCC susceptibility is observed, the interstitial SCC susceptibility decreases markedly when it exceeds 30%. If cold working of about 60% is applied between solution treatment and aging treatment (including direct aging treatment and two-stage aging treatment), the gap SCC resistance will be extremely good. An embodiment of the present invention will be described below. The material is Inconel X750 alloy. Its main chemical components are
72.92%Ni, 15.48%Cr, 6.91%Fe, 0.57%A,
2.60%Ti, 0.95%Nb+Ta, 0.04%C. The table shows the results of a constant strain test with a gap in high-temperature, high-pressure pure water. The test conditions are as follows. Test temperature: 288℃, pressure: 86Kg/cm 2 , dissolved oxygen: 8ppm, gap forming material:
Graphite wool, strain: approximately 1.0%, test time: 500h.
【表】【table】
【表】
固溶化処理（1066℃×1h→水冷）と直接時効
処理（704℃×20h→空冷）との中間に試験片の
断面減少率が10％〜60％となるような冷間加工を
施した場合、Ａ処理（冷間加工なし）に比べてＢ
処理（冷間加工：10％）、Ｃ処理（冷間加工：20
％）及びＤ処理（冷間加工：30％）の方が隙間
SCC感受性が大きく示され、30％以下の冷間加
工は隙間SCC性に関して有害であることが判
る。しかし、Ｅ処理（冷間加工：60％）ではＡ処
理に比べて耐隙間SCC性が改善される。これら
のことから、固溶化処理と直接時効との中間に施
こす冷間加工は試験片の断面減少率を60％程度に
することが良好である。
固溶化処理（1066℃×1h→水冷）と二段時効
処理（843℃×24h→空冷＋704℃×20h→空冷）
との中間に前述と同様の冷間加工を施した場合、
Ｆ処理（冷間加工なし）に比べてＧ処理（冷間加
工：10％）及びＨ処理（冷間加工：20％）の方が
若干の耐隙間SCC性の改善が認められるが、割
れ感受性が比較的大きいことから実用上での採用
は不適当である。一方、Ｉ処理（冷間加工：30
％）及びＪ処理（冷間加工：60％）ではＦ処理に
比べて割れ深さが著しく減少し耐隙間SCC性も
著しく改善される。これらのことから固溶化処理
と二段時効処理の中間に冷間加工を施す場合は試
験片の断面減少率が30％以上となるような冷間加
工を施す必要がある。
また、インコネル718合金（52.47％Ni，18.37
％Cr，0.40％Ａ，0.85％Ti，5.06％Nb，＋
Ta，）のＫ処理、Ｌ処理、Ｍ処理、Ｎ処理、Ｏ処
理及びＰ処理について上記インコネルX750合金
と同様の試験を実施した結果、表に示したように
インコネルX750合金とインコネル718合金の隙間
SCC感受性に及ぼす固溶化処理と時効処理（直
接時効処理及び二段時効処理）との中間に施す冷
間塑性加工の影響はほぼ同等であることが明らか
となつた。
以上のように本実施例によれば、固溶化処理と
時効処理との中間に試験片の断面減少率が40％〜
70％になるような冷間加工を施せば、耐隙間
SCC性がより一層に優れた原子炉用ばねが製作
できる。
本発明によれば、原子炉用ばねの耐隙間SCC
性が向上できるので、析出強化型Ni合金製のば
ね寿命も長くなり、ひいては原子炉の信頼性の向
上に効果がある。[Table] Between solution treatment (1066℃ x 1h → water cooling) and direct aging treatment (704℃ x 20h → air cooling), cold working was performed so that the area reduction rate of the test piece was 10% to 60%. When processed, B compared to A treatment (no cold working).
Treatment (cold working: 10%), C treatment (cold working: 20
%) and D treatment (cold working: 30%) have better clearance.
It can be seen that the SCC susceptibility is large, and cold working of less than 30% is detrimental to the gap SCC property. However, the E treatment (cold working: 60%) improves the gap SCC resistance compared to the A treatment. For these reasons, it is preferable that the cross-section reduction rate of the test piece be approximately 60% in the cold working performed between the solution treatment and the direct aging. Solid solution treatment (1066℃ x 1h → water cooling) and two-stage aging treatment (843℃ x 24h → air cooling + 704℃ x 20h → air cooling)
If the same cold working as mentioned above is applied between
Compared to F treatment (no cold working), G treatment (cold working: 10%) and H treatment (cold working: 20%) show a slight improvement in gap SCC resistance, but cracking susceptibility is relatively large, making it inappropriate for practical use. On the other hand, I treatment (cold working: 30
%) and J treatment (cold working: 60%), the crack depth is significantly reduced compared to the F treatment, and the gap SCC resistance is also significantly improved. For these reasons, when performing cold working between solution treatment and two-stage aging treatment, it is necessary to perform cold working such that the cross-sectional area reduction rate of the test piece is 30% or more. Also, Inconel 718 alloy (52.47% Ni, 18.37
%Cr, 0.40%A, 0.85%Ti, 5.06%Nb, +
As a result of conducting the same tests as the above Inconel X750 alloy for K treatment, L treatment, M treatment, N treatment, O treatment, and P treatment of Ta,), the gap between Inconel X750 alloy and Inconel 718 alloy was found as shown in the table.
It has become clear that the effects of cold plastic working applied between solution treatment and aging treatment (direct aging treatment and two-stage aging treatment) on SCC susceptibility are almost the same. As described above, according to this example, the cross-sectional reduction rate of the test piece was 40% to 40% between the solution treatment and the aging treatment.
If cold working is applied to reduce the gap to 70%, the gap resistance will be improved.
Nuclear reactor springs with even better SCC properties can be manufactured. According to the present invention, gap resistance SCC of nuclear reactor spring
This improves the properties of precipitation-strengthened Ni alloy springs, which in turn increases the reliability of nuclear reactors.
第１図及び第２図は原子炉用ばねの断面図及び
第３図は原子炉用ばねの製作工程を示すフロー図
である。
３…フインガースプリング、６…イクスパンヨ
ンバネ、１…タイプレート、２…チヤンネルボツ
クス、４…グラフアイト、５…チユーブ。
FIGS. 1 and 2 are cross-sectional views of a spring for a nuclear reactor, and FIG. 3 is a flow chart showing the manufacturing process of a spring for a nuclear reactor. 3...Finger spring, 6...Expansion spring, 1...Tie plate, 2...Channel box, 4...Graphite, 5...Tube.
Claims (1)
２％Ａ、0.7〜３％Ti，0.7〜８％Nbを含み、残
部Niよりなる合金を、固溶化処理し、次いで断
面減少率で40〜70％の冷間塑性加工を施した後、
時効処理することを特徴とする耐応力腐食割れ性
に優れた原子炉用ばねの製作法。 ２ 重量で、14〜25％Cr，８％以下のMo，20％
以下のFe，0.4〜２％Ａ，0.7〜３％Ti，0.7〜
８％Nbを含み、残部Niよりなる合金を、固溶化
処理し、次いで断面減少率40〜70％の冷間塑性加
工を施した後、時効処理することを特徴とする耐
応力腐食割れ性に優れた原子炉用ばねの製作法。[Claims] 1. 14-25% Cr, 20% or less Fe, 0.4-25% by weight
After subjecting an alloy containing 2% A, 0.7 to 3% Ti, 0.7 to 8% Nb, and the balance being Ni, to solid solution treatment, and then cold plastic working with an area reduction rate of 40 to 70%,
A method for manufacturing nuclear reactor springs with excellent stress corrosion cracking resistance characterized by aging treatment. 2 By weight, 14-25% Cr, 8% or less Mo, 20%
The following Fe, 0.4~2%A, 0.7~3%Ti, 0.7~
An alloy containing 8% Nb and the remainder Ni is solution treated, then subjected to cold plastic working with an area reduction rate of 40 to 70%, and then subjected to aging treatment to achieve stress corrosion cracking resistance. How to make an excellent nuclear reactor spring.
Priority Applications (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| JP56172742A JPS5877559A (en) | 1981-10-30 | 1981-10-30 | Manufacture of spring for nuclear reactor with superior stress corrosion cracking resistance |
Applications Claiming Priority (1)
| Application Number | Priority Date | Filing Date | Title |
|---|---|---|---|
| JP56172742A JPS5877559A (en) | 1981-10-30 | 1981-10-30 | Manufacture of spring for nuclear reactor with superior stress corrosion cracking resistance |
Publications (2)
| Publication Number | Publication Date |
|---|---|
| JPS5877559A JPS5877559A (en) | 1983-05-10 |
| JPS6150143B2 true JPS6150143B2 (en) | 1986-11-01 |
Family
ID=15947470
Family Applications (1)
| Application Number | Title | Priority Date | Filing Date |
|---|---|---|---|
| JP56172742A Granted JPS5877559A (en) | 1981-10-30 | 1981-10-30 | Manufacture of spring for nuclear reactor with superior stress corrosion cracking resistance |
Country Status (1)
| Country | Link |
|---|---|
| JP (1) | JPS5877559A (en) |
Families Citing this family (5)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| JPS60204849A (en) * | 1984-03-28 | 1985-10-16 | Toshiba Corp | Sealing ring for control rod driving mechanism of nuclear power plant |
| JPH0647701B2 (en) * | 1984-12-14 | 1994-06-22 | 株式会社東芝 | Electrical connection terminal clip for magnetron filament repairing |
| JPH0742560B2 (en) * | 1984-12-14 | 1995-05-10 | 株式会社東芝 | High temperature spring manufacturing method |
| JPH0684535B2 (en) * | 1984-12-27 | 1994-10-26 | 株式会社東芝 | Method for producing nickel-based alloy |
| US5827377A (en) * | 1996-10-31 | 1998-10-27 | Inco Alloys International, Inc. | Flexible alloy and components made therefrom |
Citations (1)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| JPS5531130A (en) * | 1978-08-25 | 1980-03-05 | Hitachi Metals Ltd | Heat treating method for ni alloy used in warm water |
-
1981
- 1981-10-30 JP JP56172742A patent/JPS5877559A/en active Granted
Patent Citations (1)
| Publication number | Priority date | Publication date | Assignee | Title |
|---|---|---|---|---|
| JPS5531130A (en) * | 1978-08-25 | 1980-03-05 | Hitachi Metals Ltd | Heat treating method for ni alloy used in warm water |
Also Published As
| Publication number | Publication date |
|---|---|
| JPS5877559A (en) | 1983-05-10 |
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