ML19247E434
| ML19247E434 | |
| Person / Time | |
|---|---|
| Issue date: | 04/04/1981 |
| From: | Dircks W NRC OFFICE OF THE EXECUTIVE DIRECTOR FOR OPERATIONS (EDO) |
| To: | |
| References | |
| TASK-2.K.2.13, TASK-PII, TASK-SE, TASK-TM SECY-81-286, TAC-45202, NUDOCS 8105260144 | |
| Download: ML19247E434 (3) | |
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POLICY ISSUE (Informat. ion)
For:
The Comissioners From:
William J. Dircks Executive Director for Operations
Subject:
PRESSURIZED THER."AL SHOCK Purocse:
This paper presents background infor.ation en the issue of pressurized thermal shock of reactor pressure vessels and describes actions underway and planned by the staff for cealing with the issue.
Discussion:
The issue of pressure vessel thermal shock has been considered by NRC for many years. The early concerns were centered around the integrity of the vessel when subjected to cold emergency core cooling (ECC) water during a large break, loss-of-coolant accident (LOCA). A number of analyses have been made by industry experts, NRC staff and NRC contractors.
Based cn these analyses, as well as ther al shock experiments (unpressurized) at CRNL, the staff has concluded that a crack cannot propagate through the ves';el wall during a large break LOCA.
For nor al operation and anticipated operatier.a1 occurrences, NRC regulations (10 CFR 50, 'ppendix G) place requirements on vessel fr.acture toughness aimed Pt providing an adequate marcin of protection acainst fracture, taking into account the potential for such factors as ther' a1 shock. There may be some FWR transient secuences, however, in which the vessel could be subjected to thermal shock at the same time that primary system pressure remained high.
In these pressurized ther al shock transients, the vessel would be subjected to tensile stresses superimposed up::n the thermal stresses resulting from the thermal gradient across the vessel wall.
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The Commissicners The probability of pressure vessel failure due to pressurized thermal shock depends on the following factors:
(a) the frequency tnd severity Of evercooling transients; (b) the prcbability nat the crimary systeT. remains pressurized or is repressurized through operator actions; (c) the existence of a flaw of sufficient size to propagate at the locaticn of high thermal stresses; and (c) the frac'ture toughness, or resistance to crack propagation, cf the vessel, wnich depends On the copper conten: Of weld material and on the irraciation history Of the vessel.
L veral overcocling transients have occurred in operating FWRs, tne most serious of wnich was a transient at the Rancho Seco plant on March 20, 1978. The.NER staff repuested B&W to perform a fracture mechanics analysis of the vessel for tne transient conditions experienced, and the staff performed an independent analysis. The staff concluded that, althouch the Appendix G limits were exceeded, the Rancho Seco vessel was not damaged to tra extent that it reduced its expected service life. The staff stressed, however, that the safety implications were minimal only because the transient occurred very early in plant life when the fracture toughness of the vessel remained high.
The TMI Ac:icn Plan included a task (II.K.2.13) that requires a detailed analysis for all PWRs of the potential for thermal shock of reactor vessels resulting from ceid safety injection ficw duri.no small break LOCAs. This work is eroceedino on s chedul e.
In addition, the Office of Research has a pressurized thermal shock test facility under construction at ORNL (discussed in SECY-79-4Eg), and the first tests are scheduled for 1982.
During the past year, RES has studied evercooling transients more severe than the Rancho Se:0 transient and has investigated a range of vessel material properties. One of the results of a recent fracture mechanics analysis carried out by CENL indicated that, if the Rancho Secc transient had cccurred after 10 effective full-power years (more than twice its current level), the pr:bability of failure of the Rancho Secc vessel would have been very high.
Aftcr reviening these analyses, NRR called a meeting with the PWR incastry Re;ulatory Respcnse Groups (RRis) and the PWR reactor manufacturers en March 31, 1951. The RRG representatives agreed
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The Commissioners to send NER a report by May 15,19S1, describing their on-coing work and acdressing the issue of pressuri:ed thermal shock. They were, subsecuently, rLquested to provide a generic bas'is for continued safe operation of the plants.
On April 10, 1931, Demetries Easdekas wrote a letter to Ccngressman Ucall in which he reiterated his concern about pressurized thermal sheck transients. He further suggested that these PWRs with high copper content weld material that have operated for 4 effective full-power years be shut down until the issue is resclved in the technical arena.
The staff received a pr0gress tricfing from the ?WR Owners Group en April 29, 1951. At this b-iefing, the Owners Group representatives assertec that there was no need for immeciate corrective actions.
These assertions were based en the icw prctability of severe overcooling transients, as well as the hign fracture toughness of the vessels at this time. They agreed to provice more technical backup in their May 15 report.
The NRR staff has made an independent revier of this issue and has concluded that no immediate licensing actions are required for operating reactors (Enclosure 1). The staff will review the ERS report when it is submitted, and in addition, the NRR and RES staffs will prepare a state-of-the-art repcrt en pressurized tha" mal shock within the next few months. We will keep the Com:.ssion informed cf the prcgress of these reviews.
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William J. Dircks Executive Director for Operations
Enclosure:
Memo, Eiser. hut to Denton etd cpril EE,1??l, " Thermal Shock to PWR Reacter" e
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