ML20032A662
| ML20032A662 | |
| Person / Time | |
|---|---|
| Site: | Zion File:ZionSolutions icon.png |
| Issue date: | 10/23/1981 |
| From: | Wigginton D Office of Nuclear Reactor Regulation |
| To: | Feuerstein R AFFILIATION NOT ASSIGNED |
| Shared Package | |
| ML20032A663 | List: |
| References | |
| NUDOCS 8111020084 | |
| Download: ML20032A662 (5) | |
Text
l Distribution Docket File (50-295)
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/g[,gi-e OCT 2 31981 D. Eisenhut R. Purple l
q J. Heltemes L
T. Murley
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Ms. Renee Feuerstein S. Varga t
8 6 ISgg rl 8844 No. Lacrosse D. W'gginton 84 gn%. Z
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Skokie, Illinois 60077 C. Parrish Q
Dear Ms. Feuerstein:
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 Comission't (NRC) project manager for licensing matters at Zion.
I must assume that the safety concern you have mantioned is the postulated thermal shock to the reactor vessel following an overcooling transient; we refer to this simply as "thennal shock." For your 1r.fomation, I have enclosed a short synopsis on this issue. The Zion Station vessels have not received the radiation exposure that would make them a safcty 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 information, the NRC maintains a resident inspector at the Zion Site; Joel Kohler can be reached on telephone number 312-746-2313.
Sincerely, Criginal Sinned By(
David Wigginton, Project Manager Operating Reactors Branch No.1 l
Division of Licensing
Enclosure:
i As stated l
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Enclosure NUCLEAR REACTOR PRESSURE VESSEL INTEGRIT.YiWHEN SUBJECTED T0; THERMAL l SHOCK]ND SUBSEQUENT 'REPRESSURIZATION.DURING,
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'AN OVERC00 LING TRANSIENT'
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' '.(. PRESSURIZED THERMAL SH0CK)~ '
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Pressure vessel th5 mal shock'has been considered'for ma6y y5srs in~the
[o'ntext of assuring iittegrity of the vessel when subjected to cold
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j emergency core cooling water during a large. loss of coolant accident' l
(LOCA).
Basbd on'a series of thermal shock expe'iments (unpresiurized) r
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conducted at Oak Ridge National' Laboratory (ORNL) beginning in 1976'and'~'
based on' fracture mechanics analyses 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 j
maintained during subsequent reflooding which-would occur at relatively low pressure due to pre'sbnce of the large break.
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l 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 pressuri.ted 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 believ'ed to represent the most severe (and prolonged) overcooling transient experienced to date.
In that event. a lightbulb being replaced L
in the non-nuclear instrumentation / integrated control system (NNI/ICS) panel was dropped and caused a short to occur while the plant was at l
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.
Subsequent refilling f
by the main feedwater (MFW) system caused a primary system overcooling l
and an actuation of high pressure injection (HPI) and energency feedwater O
(EFW).
Actuation of HPI and EFW caused severe cvercooling. rates (approx-b
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.imately 300 F/h ) until the pumps'were partly secured by plant operators."
0 Actuation of SPI also caused repressurization of the primary system.
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Operators did not recognize until approximately one hour later that 0
primary system [ temperature. had been, reduced.to about 285 F (because of preoccupation with restoration af NNI/ICS equipment).
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{f an. overcooling event, su,ch as that at Rancho Seco in 1978 were to occur even for the vessel with the wor:t material. properties in'the current
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population of reactor vessels, the stiff would not expect a. failure.
The staff co6cludion is.suphorted'by an analysis of the.Radho Seco
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event perfomed 6y the Oak Ridge National Laboratory which' indicated
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that it would be several years before.any B&W-designed facility reached tiie threshold irradiation level for crack, initiation (that.is, small' crack.s growing to larger ones assuming conserfative initial material
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properties for pressurized overcooling events equal in severity to the Rancho Seco event).
Some. reactor vessels in Combustion Engineering (CE) aiid 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 a@
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 beginn'ing in early 1980.
These investigations included' defining the
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cooldown transients and accidehts 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 evaluations of this analytical work indicated that there couid be a problem if pre.ssure vessels having initial material properties (fracture toughnese) less favorable than those fabricated more recently e
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i were subjected to severe pressurized cooldown transients after many years of neutron irradiation.
In order to assess the need for any 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 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 conclusior, that no inmediate action was re,uired on any operating reactor.
The staff has detennined 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 have been held on many occask's with industry representatives for detailed discussions and exchanges of infonnation.
2.
Evaluations are continuing for refinement of the staff's understanding of this safety concern and better definition of wh'at actions thE industry and staff must take to resolve this issue.
A number of efforts are now underway by the NRC staff to develop a better technical basis for a final resolution for this problem.
These programs may show the need for more extensive corrective actica before vessel.s approach their end of design life state.
A new project ha be.en initiated at Oak Ridge National Lab' oratory (0.RNL) to bring together a canprehensive 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 thennal shock test facility are scheduled for FY1982.
The development of a computer code for probabilistic analysis of reacter pressure vessel failure utilizing fracture mechanics and Monte Carlo simulation techniques is continuing.
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i Several potential correcti e 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 1n 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
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temperature cf the energency core cooling system (ECCS) injection d
' water; and 4.
Reducing the probability of the event by control system designs that would prevent repressurization, and/or by operator actions to prevent repres'surization.
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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