ML20011B058
| ML20011B058 | |
| Person / Time | |
|---|---|
| Site: | Zion File:ZionSolutions icon.png |
| Issue date: | 10/23/1981 |
| From: | Wigginton D Office of Nuclear Reactor Regulation |
| To: | Berger R AFFILIATION NOT ASSIGNED |
| Shared Package | |
| ML20011B059 | List: |
| References | |
| NUDOCS 8111040112 | |
| Download: ML20011B058 (5) | |
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Ms. Robin Berger S. Varga 1
-125 Forestway Drive D. Wigginton Deerfield. Illinois 60015 C. Parrish
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Dear Ms. Berger:
Thank yourfor your letter to Thomas Murley regarding your concerns for the safety at the Zion Station. Your letter was forwarded to me since I am the Nuclear Regulatory Commission's (NRC) project manager for licensing matters at Zion.
I must assume that the safety concern you have mentioned is the postulated themal shock to the reactor vessel following an overcooling transient; we refer to this simply as " thermal shock." For your information, I have enclosed a short synops.'s on.this issue. The Zion Station vessels have not received the radiation exposure that would make them a safety concern at this time, however, our program is scheduled to resolve the matter before the vessels are susceptable to damage from any overcooling transient. We hope this information will be of benefit to you.
If you have any further questions on the themal shock issue or any other matter that you feel presents an undue hazard, please let us know. Also for your infomation, the NRC maintains a resident inspector at the Zion Site; Joel Kohler can be reached on telephone number 312-746-2313.
Sincerely, g ggg1.SiRDea M David Wigginton, Project Manager Operating Reactors Branch No. 1 Division of Licensing
Enclosure:
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Enclosure
~ NUCLEAR REACTOR PRESSURE VESSEL INTEGRIT.Y:WHEN SUBJECTED 3'
TO THERMAL SH0CK AND SUBSEQUENT REPRESSURIZATION,DURING,
AN OVERC00 LING TRANSIENT-
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'(PRESSURIZEDTHERMALSHOCK) 3__:
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Pressure vessel themal ~ shock has been considered for many years-in'the context of assuring iiltegrity of the vessel when subjected to cold emergency core cooling water during a large loss of coolant accident ~
(LOCA).
Based on a series of themal shock experiments (unpressurized)'
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conducted' at Oak Ridge National Labo~ratory (ORNL) beginning in 1976~and ' '
based on fracture mechanics analyses verified by the experiments, it was 66ncluded that a poste1ated flaw would not propagate through the vessel wall during a large LOCA.
Therefore, the vessel in'tegrity would be rhaintained during subsequent reflooding'which-would occur at relatively ~
i low pressure due to presence'of the large break.
1 As the result of operating experience, it was subsequently recognized that there could be transients in pressurized water reactors (PWRs) in which the vessel could be subjected to severe overcooling (themal shock) followed by repressurization.
In these pressurized themal shock
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transients, vessels would be subjected to pressure stresses superimposed upon the themal stresses resulting from the temperature difference across, the vessel wall.
The Rancho Seco event.of March 20,1978 is believed to represent the most severe (and prolonged) overcooling l
transient experienced to date.
In that event, a lightbulb being replaced f
in the non-nuclear instrumentation / integrated control system (NNI/ICS) panel was dropped and caused a short to occur while the plant was at approximately 70% power.
About 2/3 of the pressure, temperature and level indication was lost. The reactor tripped, feedwater was lost and i
the once through steam generators (OTSGs) dried out.
Subsequent refill.ing by the main feedwater (MFW) system caused a primary system overcooling and an actuation of high pressure injection (HPI) and emergency feedwater 1
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(EFW).
Actuation of HPI and EFW caused severe overcooling, rates (approx-imately 300 F/hr) until the pumps 'were partly secured bj plant operators.
0 Actuation of HPI also caused represserization of the primary system.
Operators did not recognize until appio,amately one hour later that primary system temperatUr~e had bein reduced to about'285 F ibecause of
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preoccupation witb~ restoration'b; NNI/ICS equ:pmenth
.c If an overcooling event such as that at Rancho Seco in 1978 were to occur even for the vessel with }he worst material propert{es in thd current population of reactor vessels, the staff would not extiect 'a failure.
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The staff conclusion is supported by an ~ analysis of the Rancho Seco event performed ~by the 05k ' Ridge ' National Laboratory Ubich'in~dicated '
that it would be se, s1 years before any B&W-designed facility reached 2
dhe threshold irradiati$n level for crack initiationlthat.is, smai1 w
cracks growing to larger 'ones assuming conservative' initial material properties for ~ pressurized overcooling events, equal in severity to the Rancho Seco event).
Some reactor vessels in Cmbustion Engineering (CE) and Westinghouse (W) facilities have somewhat higher irradiation histories; however, other mitigating factors provide a Ignificant margin to failure should a pressurized overcooling event similar to that at' Rancho Seco occur.
In order to define what transient conditions more severe than the Rancho Seco event would be necessary to propagate a flaw through the entire vessel thickness, a number of investigations were initiated by the staff beginning in early 1980.
These investigations included' defining the cooldoan transients and accide'nts of interest and their respective probability, development of a computer code to perfom the themal transient and fracture mechanics analyses, and planning for pressurized themal shock tests in the Heavy-Section Steel Technology Program at ORNL.
The staff evaluati,ons of this analytical work indicated that there could be a problem if pressure vessels having initial material properties (fracture toughness) less favorable than those fabricated more recently y
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were subjected to severe pressurized cooldown transients after many years of neutron irradiation.
In order to assess the need for 'ny immediate action, the PWR industry Regulatory Response Groups (RRGs) and PWR reactor manufacturers were briefed on this ' issue by the staff on March 31, 1981.
In a progress briefing on April 29, 1981, the PWR Owners' Group asserted thv.t there was no need for immediate corrective action.
On May 15, 1981, the 'lestinghouse, Combustion Engineering and Babcock i Wilcox Owners' Groups filed written responses supporting and reiterating their conclusion that no inmediate action was required on any operating reactor.
The staff has determin.
that no immediate licensing actions are required for plants under construction, plants under review for operating licenses, or operating facilities; however, the staff has taken the following actions:
1.
Meetings liave been held on many occasions with industry representatives for detailed disemions and exchanges of infomation.
'2.
Evaluations are continuing for refinement of the staff's understanding of this safety concern and b'etter definition of what actions thi industry and staff must take to resolve this issue.
A number of efforts are now underway by the NRC staff to develop a bettet technical basis for a final resolution for this problem.
These proc ams may show the need for more extensive corrective action before versells approach their end of design life state.
A new project has been iaitiated at Oak Ridge National Lab' oratory (0.RNL) to bring together a aprehensive evaluation of the many aspects of this problem in order to define the best course of regulatory action toward its inderstanding and resolution.
The Heavy-Section Steel Technology Program at ORNL is continuing, and first tests using a new pressurized tiemal shock test facility are scheduled for FY1982.
The development of a computer code
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for probabilistic analysis of reactor pressure vessel failure utilizing fracture mechanics and Monte Carlo simulation tech iques is continuing.
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5 Several potential corrective actions are possible, and will be considered.
These include:
1.
Reducing the neutron irradiation of the pressure vessel by. replacing some or all of the outer row of fuel elements in the core with partially loaded or reflector elements; 2.
Annealing the reactor pressure vessel in-situ to restore a major fraction of the fracture toughness which was lost due to neutron i rradiation.
Annealing is feasible from a metallurgical standpoint, but' practical application is difficult and potentially expensive; 3.
Recucing the thermal' shock during some transients by raising the temperature of the emergency core cooling system (ECCS) injection
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Reducing the probability of the event 4y control system designs that would prevent repressurization, and/or by operator actions to pr: vent repressurization.
The NRC staff and its. contractors have been, and will continue to be, extensively involved 'in the development of the technology of this issue.
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