ML20032B100
| ML20032B100 | |
| Person / Time | |
|---|---|
| Site: | Zion File:ZionSolutions icon.png |
| Issue date: | 10/23/1981 |
| From: | Wigginton D Office of Nuclear Reactor Regulation |
| To: | Lipson H AFFILIATION NOT ASSIGNED |
| Shared Package | |
| ML20032B101 | List: |
| References | |
| NUDOCS 8111040446 | |
| Download: ML20032B100 (5) | |
Text
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Docket File (50-295)
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T. Murley h (({fFf OCT 2 31981 D. Eisenhut y
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Mr. A Mrs. H. M. Lipson D. Wigl nton 7 u.s.w.ou m m, C
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Chicago, Illinois 60611 j N
Dear Mr. & Mrs. Lipson Thank you for 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 thermal 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 synopsis 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 thermal shock issue or any other matter that you feel presents an undue hazard, please let us know. Also for your infomation, the flRC maintains a re3ident inspector at the Zion Site; Joel Kohler can be reached on telephone number 312-746-2313.
Sincerely, briginal Signed By:
David Wigginton, Project Manager Operating Reactors Branch ?!o.1 Division of Licensing
Enclosure:
As stated n.
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L Enc 1o'sure NUCLEAR REACTOR PRESSURE VESSEL INTEGRIT.Y.WHEN' SUBJECTED TOTHERMLSHCfANDSUBSEQUENTREPRESSURIZATIONDURING AN'0VERC00 LING TRANSIENT
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.s Pressure vessel themal shock'has been considered for many years'in'the dontext of ar uring ihtegrity 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 tjational'i.aboEatory (ORNL) beginning in i976 and based on fracture mechanics analysee verified by the experiments, it was concluded that a postulated flaw would not propagate through the vessel wall during a large LOCA. Therefore, the vessel integrity would be nihintained during subsequent reflooding which would occur'at relatively~
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low pressure due'to presence' of the large break.
As the result of operating experience, it was subsequently recognized that there could be transients in pressurized water reactors (PelRs) in which the vessel could be subjected to severe overcooling (themal shock) followed by repressurization.
In these pressurized themal shock 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 believ'ed to represent the most. severe (and prolonged) overcooling transient experienced to date.
In that event, a lightbulb being replaced 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 the once through steam generators (OTSGs) dried out.
59bsequent refilling by the main feedwater (MFW) system caused a primary system overcooling and an actuation of high pressure injection (HPI) and emergency feedwater
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(EFW).
Actuation of HP: and EFW caused severe Overcooling rates (approx-imately 300 F/hr) until the pumps were partly secured 'by plant operators.
Actuation of'HPI also caused repressurization of' the primary system.
Operators did ~ not recognize until approximately one hour later that plinary-system. temperature had been.reducedito about 285 F (because of preoccupation with restorati.on of NNI/ICS equipment). ~~
If an overcooling event,such as that at Rancho,Seco in 1978 were to occur even for the ves,sel with 'the worst material properties in the current population of reactor vessels, th_e staff would not expect.a failure.
The staff conclusion is supported'by an analysis of'the Rancho Seco event perfomed by the Oak Ridge' National Laboratory which~ indicated that it would be several years before any B&W-designed facility reached tiie threshold irradiation level for crack initiation (that is, small i'
cracks growing to larger ones as~suming conservative initial material properties for' pressurized overcooling events equal in severity to the
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Rancho Seco event).
Some. reactor vessels in Combustion Engineering (CE) and Westinghouse (W) facilities have somewhat higher irradiation histories; however,. other mitigating factors provide a significant 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 These investi'ations included' defining the I
beginn'ing in early 1980.
g cooldown transients and accidents of interest and their respective
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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 l
ORNL.
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l The staff evaluations of this analytical work indicated that there could 1
l be a problem if pressure vessels having initial material procerties (fracture toughness) less favorable than those fabricated more recently l
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were subjected to severe pressurized cooldown transients after many years of neu' tron irradiation'.
In order to assess the need for any
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immediats 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 that there was no need for immediate corrective action.
On May 15,1981, the Westinghouse, Combustion Engineering and Babcock & Wilcox Owners' Groups filed written responses supporting and reiterating their conclusion that no immediate action was required on i ~
any operating reactor.
The staff. has detemined that no immediate licensing actions are required for plants under construction, plants under review for operating licenses, or operating facilities; hwever, the staff has taken the following.
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actions:
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Meetings have been held on many occasions with industry representatives for detailed discussions and exchanges of infomation.
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Evaluations are continuing for refinement of the staff's understanding of this safety concern and better definition of what actions thi industry and staff must take to resolve this issue.
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A number of efforts are now undemay by the NRC staff to develop a
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be.tter technical basis for a final resolution for this problem.- These l
programs may show the need for more extensive correctiv,e action before vessel.s approach their end of design life state.
A new project has been initiated at Oak Ridge National Lab' oratory (0.RNL) to bring together a cmprehensive evaluation of the many aspects of this problem in order to define the best course of regulatory action toward its understanding and resolution.
The Heavy-Section Steel Technology Program at ORNL is continuing, and first tests using a new pressurized themal shock test facility are scheduled 'for FY1982. The development of a computer code for probabilistic analysis of reactor pressure vessel failure utilizing fracture mechanics and Monte Carlo simulation techniques is continuing.
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4 Several potential corrective actions are possible, and will be considered.
These include:
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Reducing tha 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.
Reducing the thennal shock during some transients by raising the temperature of the energency core cooling system (ECCS) injection
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Reducing the probability of the event by control system designs that would prevent repressurization, ano/or by operator actions to prevent repres'surization.
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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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