ML19263F291
| ML19263F291 | |
| Person / Time | |
|---|---|
| Issue date: | 05/15/1978 |
| From: | Eisenhut D Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19263F290 | List: |
| References | |
| REF-GTECI-A-12, REF-GTECI-EQ, TASK-A-12, TASK-OR NUDOCS 8001110457 | |
| Download: ML19263F291 (5) | |
Text
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CATEGORY A TECHNICAL ACTIVITY NO. A-12 PROPOSED REVISION 1
Title:
Fracture Toughness and Potential for Lamellar Tearing of Steam 1
Gene _rator and Reactor Coolant Pump Supports lead Responsibility: Division of Operating Reactors Lead Assistant Director:
Darrell G. Eisenhut, Assistant Director for Systems & Projects, D0R Task Manager:
Dick Snaider, D0R 1.
Problem
Description:
During the course of the licensing action for North Anna Power Station Unit Nos.1 and 2 a number of questions were raised as to the potential for lamellar tearingl/ and low fracture toughness of the steam generator and reactor coolant pump support materials for those facilities. Two different steel specifications (ASTM A36-70a and ASTM A572-70a) covered most of the material used for these supports. Toughness tests, not originally specified and not in the relevant ASTM specifications, were made on those heats for which excess material was available. The tough-ness of the A36 steel was found to be adequate, but the toughness of the A572 steel was relatively poor at an operating temperature of 80 F.
In the case of North Anna Unit Nos. 1 and 2, the applicant has agreed to raise the temperature of the ASTM A572 beams in the steam generator supports to a minimum temperature of 225 F prior to reactor coolant system pressurization to levels above 1000 psig. Auxiliary electrical heat will be supplied as necessary to supplement the heat derived from the reactor coolant loop to obtain the required operating temperature of the support materials.
Since similar materials and designs have been used on other nuclear plants, the concerns regarding the supports for the North Anna facilities may be applicable for other PWR plants.
It is therefore necessary to reassess the fracture toughness of the steam generator and reactor coolant pump support materials for all operating PWR plants and those in CP and OL review.
1/ amellar tearing is a cracking phenomenon which occurs beneath welds L
and is principally found in rolled steel plate fabrications. The tearing always lies within the parent plate, often outside the s
transformed (visible) heat-affected zone (HAZ) and is generally parallel to the weld fusion boundary. Lamellar tearing occurs at certain critical joints usually within large welded structures involving a high degree of stiffness and restraint.
Restraint may be defined as a restriction of the movement of the various joint components that would normally occur as a result of expansion and contraction of weld metal and adjacent regions during welding.
("Lamellar Tea ing in Welded Steel Fabrication", The Welding Institute).
r 2213 150 s ooo., yg 7
Lamellar tearing may also be a problem in those support structures similar in design to North Anna. This possibility will be investi-gated on a generic basis.
The scope of this program is presently limited to PWR steam generator and reactor coolant pump supports. Another program, ASYMMETRIC LOCA LOADS (A-2) will investigate vessel supports as part of its scope as part of that effort, a review of the need for including BWR vessel supports is being undertaken. As noted in Section 8 of this report, activity A-12 can be expanded to include BWR supports and other PWR support structures if warranted.
2.
Plan for Problem Resolution:
A preliminary survey of operating PWR plants was made in May,1976 to determine the initial scope of this problem.
Results indicate that five units have designs similar to North Anna and that 12 units use A36 materials. No plants which were surveyed used the A572 material.
The staff concluded that, depending on the heat treatment of the A36 material, a potential material toughness problem may exist.
In addi-tion, it was determined that other materials used in the design of steam generator and pump supports have never been tested to determine toughness properties. Therefore, the potential " toughness problem" may exist for operating plants that did not use A36/A572.
As noted above, the poten-tial for lamellar tearing may also exist for certain support structures.
Based on the above, the continuing action plan for resolution of this concern for operating PWRs is as follows:
a.
Send a generic letter to all PWR licensees stating NRC concerns and requesting information on the design, materials, fabrication and inspection of the steam generator and reactor coolant pump supports for each plant. This transmittal was completed for operating PWRs I
in October 1977.
(A follow-on letter to BWR licensees may be necessitated by information developed in program A-2).
b.
Based on information supplied by the licensees and with the aid of the consultant, categorize the support design and materials as far as practical and select typical designs for further study.
DSS /MTEB will concurrently review fracture toughness and possibility of lamellar tearing for PWRs in the CP and OL stages, based on informa-tion gathered from the 00R review.
c.
Complete preliminary review of typical designs and inform each applicable PWR licensee of the concerns on their particular support system.
2213 151 Utilizing input from consultart, develop and issue specific guidance for resolution of the problems discovered. This will be a joint DSS / DOR task and will result in the issuance of a NUREG document and/or other appropriate document.
Subsequent case-by-case resolution (implementation) will involve requiring those applicants or licensees for whose facility (ies) a problem exists to either:
(1) demonstrate that safety margins are not lower than anticipated or; (2) propose a solution to the problem in accordance with the criteria developed in step d above.
3.
NRR Technical Organizations Involved:
a.
Engineering Branch, Division of Operating Reactors.
Has lead responsibility for review of data generated from licensee responses, control of and coordination with consultant organization, and will coordinate with DSS in development and issuance of criteria.
I Manpower Estimates:
1.0 manyears FY 1978,.6 manyears FY 1979.
b.
Materials Engineering Branch, Division of Systems Safety.
Review infonnation received from operating units and problems identified during review. Coordinate with D0R in development and issuance of cri te ria.
I Manpower Estimates:
.3 manyears FY 1978,.3 manyears FY 1979 c.
Task Manager, Division of Operating Reactors.
Has overall responsibility for coordination of DDR and DSS technical tasks and for the development and issuance of criteria documents.
I Manpower Estimates:
.1 manyears FY 1978,.1 manyears FY 1979.
4.
Technical Assistance Requirements:
Technical assistance for the D0R program is required to provide expertise in evaluating the potential for lamellar tearing and low fracture tough-ness of the support materials. The work will include:
a.
Evaluating utility responses to NRC questions.
b.
Performing a literature search for fracture toughness data on I
the materials in question.
c.
Evaluating the brittle failure potential of support materials.
22l3 152
. d.
Evaluating the potential for lamellar tearing and assessing its consequences.
1 e.
Evaluating any proposed solutions as requested by the NRC.
f.
Preparing a topical report.
The present budget is $100,000, $50,000 of which was carried over from FY 1977. We anticipate Sandia's continued participatio6 in the program completion in FY 79. 00R has requested a budget alloca-1 tion of $35,000 for this effort.
5.
Interactions With Outside Organizations:
Individual licensees of PWR facilities and applicants for PWR licenses.
All PWR licensees have been contacted to gather information at the 1
comencement of the program. Some licensees will become more involved in this study due to the need for site visits and/or the discovery of material problems at their particular facility (ies).
Further inter-action will be a function of the results of our review.
DSS will perform information review during CP and OL stages of review in order to resolve issues prior to licensing.
6.
Interaction With Other NRC Offices:
The Office of Standards Development intends to commence, in FY 1979, work on a program involving Fabrication and Examination of Component Supports. Although an effort is presently being made to incorporate specific guidance in the ASME Code, this new program may result in issuance of a Regulatory Guide.
7.
Schedule for Problem Resolution:
The major milestones for this program are as follows:
a.
Send generic letter to operating reactors complete b.
Obtain consultant complete c.
Select typical designs for lamellar tearing May 15, 1078 study d.
Complete preliminary review of operating June 30, 1978 1
units s
e.
Receive input on generic resolution from Sept. 29, 1978 consultant f.
Issue branch technical position paper /NUREG August 31, 1979 1
Document 2213 153
. 8.
Potential Problems Although this program is presently aimed at a problem known to exist for PWRs, the scope of review could uncover similar problems rA BWR facilities or additional PWR component support problems, nece':sitating a major change in the program. Additionally, significant input from the A-2 study regardina load combinations and LOCA load increases I
could result in the need for re-evaluation of the conclusions drawn from A-12.
2213 154