ML20236H148
| ML20236H148 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim, Seabrook, 05000000 |
| Issue date: | 05/30/1987 |
| From: | Piercebjorklun AFFILIATION NOT ASSIGNED |
| To: | Zech L NRC COMMISSION (OCM) |
| Shared Package | |
| ML20236H087 | List: |
| References | |
| NUDOCS 8708050075 | |
| Download: ML20236H148 (2) | |
Text
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30 May, 1987 s
Lando Zech, Chairman U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Sir:
I At your instruction Mr. Varga of your staff has responded to my mailgram
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of April 28, 1987 in which I demanded proof that a loss of coolant accident could not occur in the reactor core at Seabrook or Pilgrim.
Mr. Varga replies l
that single valve failures (not multiple valve failures) and expansion of Zirconium rods under temperature increase are considered in the LOCA (loss of coolant accident) computer analyses for Seabrook & Pilgrim nuclear stations.
Surely it is clear that computer models and analyses are not proof.
They are l
conjecture based on assumptions.
Those assumptions underlying the LOCA analyses have never been verified.
They cannot be verified without inducing a meltdown and contriving a witness to the meltdown.
Heat measuring and mechanical devices fail at those extreme temperatures and certainly people do.
Simulated heat is not equivalent to full scale and full temperature atomic heat generated at melt-down of large reactor cores.
Zircaloy at 23000F cooled by water which boils and turns to steam at 3120F actually undergoes a fundamental transformation.
It crumbles upon cooling.
Under escalating temperatures in a reactor core which 0
could go from 600 F to 20000F in a matter of seconds, no cooling medium could effectively arrive in time and in quantity sufficient to prevent disaster.
In fact steam in the core conceivably would not serve a cooling function it would add to the thermal gain of the system resulting in rupture and explosion.
I would like to quote from testimony obtained from your records and memoranda by A.E.C. officials in 1971, coments which were ignored in the formulation of the rules cited by Mr. Varga.
These official comments demonstrate that computer postulates cannot match actual thermal circumstance in reactors and do not constitute proof of anything:
- 1) Mr. Rittenhouse, Oak Ridge staff of AEC, stated vader oath in February 1971:
"Certainly many of the things we toss around in computer codes and use to predict maximum temperatures or to predict the course of a loss of coolant z
accident... have not been verified....I have worked in the fuel rod failure program with questions of fuel cladding and swelling, subsequent blockage and possible effects of this blockage on cooling effectiveness.
As far as these points in which I am an expert there is not the information available to confirm exactly wha +. effect these material related phenomena may have on the ECCS (Emergency Core Cooling System) in the event of a loss of coolant accident."
- 2) Normantauben, AEC Regulatory Staff agineer, stated that he "did not think certain heat transfer coefficients used in industry computer methods n
were conservative enough.."
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- 3) Milton Shaw, Director of AEC Division of Reactor Development, stated categori-8 oo cally "although test information is available on the response of simulated (not actual) fuel rod bundles to a range of emergency coolant flow conditions M*
j no assurance is yet available that emergency coolant.can be delivered at the,
rates intended and in the time period prior to Zircalory clad and subsequent oc fuel melting".
Computer methods relied upon by the AEC to predict ECCS S@
performance "are unable to describe important physical phenomena (that occur g
during loss of coolant) and therefore unable to confidently define safety g
margins."
4)
W. Cottrell, Director of Nuclear Safety Program at Oak Ridge:
"We are not certain that ECCS system adopted by the AEC will provide reasonable assurance that such systems will be effective in loss of coolant accidents... There seems little chance of preventing at least some degree of fuel rod swelling N
and rupture... total blockage is poggibl%"
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- 5) Broc'kett at Idaho Laboratories:
" reactor cooling problems are so complex and data so limited that complete and correct analysis of cooling accidents i
is beyond the scope of currently used techniques and in some cases beyond present scientific knowledge." He warned against " illusory margins of safety". Ifit is possible during a LOCA for fuel temperature to increase several thousand degrees in a few seconds as shown in some computations, then no coolant could be introduced quickly enough and that which does enter would vaporize increasing the steam binding slow down of coolant entry.
- 6) Stanley Szawlewicz of AEC Division of Reactor Development in a memo to members of Steering Comittee on Wash 470 stated that rebuttal of Wasb 470 (a prediction of 45,000 dead in a 1000 megawatt reactor meltdown l
accident) would require " experimental proof" that safety systems would prevent catastrophic accidents.
Such proof would be " difficult and i
expensive" and waiting for it would delay plans for reactor construction.
This was before Seabrook had even broken ground.
l l
- 7) A Rosen-Colman memo to Hanauer in 1971 "The consummate message is that I
the system performance cannot be defined with sufficient assurance to provide a clear basis for licensing."
Rosen-Colman's memo..." the multiplicity of computer coding techniques, of various reactor system designs, of assumptions regarding fundamental physical events...
have all served to raise serious question regarding the basis for reasonable assurance concerning the operation of these complicated systems".
There is ample evidence in your files that the AEC & NRC repeatedly ignored l
or repressed legitimate safety concerns expressed by ACRS (Advisory Committee l
l on Reactor Safety).
It is stated clearly in your files that a Dr. R. Doan warned in 1971 that "the possibility exists that somecne will take legal action against further reactor construction (and licensing) on the grounds that in the light of this report (Wash. 470 - Brookhaven report predicting 45,000 dead) AEC was being irresponsible in granting licenses."
I think the time has come to address this last point and I would ask Attorney General J. Shannon of Massachusetts to convene a grand jury to hear evidence and ask the Massachusetts Senators and Representatives to call for Congressional hearings on the matter of safety record of nuclear reactors, daily effluent, monitors, history of accidents, etc. before Seabrook gets a license.
Sincerely, 4.. k %
MA -
c v !'i%(
Patricia Pierce-Bjorklund copies to:
Gov.
Dukakis, Attorney General Shannon, Senators Kennedy, Kerry, Markey Representatives Mavroules, Atkins
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gmang k
UNITED STATES
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NUCLEAR REGULATORY COMMISSION l;
j W ASHING TON, D. C. 20555 4
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May 20, 1987 Ms. Patricia Pierce-Bjorklund 15 Spring Street Essex, Massachusetts 01929
Dear Ms. Pierce-Bjorklund:
Your mailgram of April 28, 1967 to Chairman Zech was referred to me for response.
Before addressing the issues raised in your mailgram, I need to remove any
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confusion about Pilgrim and a low power license.
Pilgrim currently holds a full power operating license that was issued September 15, 1972. Therefore, I am assuming your telegraphic referral to Pilgrim is in regards to restart of the plant.
Below I have repeated each of your issues and my response:
Issue We demand proof prior to the issuance of low power license to Seabrook or Pilgrim Nuclear Stations that the following circumstances cannot occur.
1.
Valve failure permitting loss of coolant from the reactor core through the failed valve so that emergency coolant entering the cooler would not also exit through th.e same failed valve thereby leaving the core without coolant.
Response
l As a part of the licensing bases for nuclear power plants, licensees l
must perform safety analyses to demonstrate that the emergency core l
cooling systems conform to the requirements of 10 CFR 50.46 and Appendix K to 10 CFR 50 for postulated loss-of-coolant accident:;
(LOCAs). These LOCA analyses are performed for a spectrum of postulated break sizes at the worst location in the primary system and directly account for possible loss of emergency core cooling through the break.
In addition, single failure analyses are performed for the emergency core cooling systems to demonstrate that adequate core cooling will be provided assuming a gngle active valve failure in these systems.
Since the. staff has concluded in Section 15.6.5 of the Seabrook Safety Evaluation Report dated March 1983, and in Section 9 of the Pilgrim Safety Evaluation dated August 1971, that the emergency core cooling systems for these two plants me M he requirements of 10 CFR 50.46, there is no single active valve failure which would prevent the core from being adequately co61ed iiIthe event of a LOCA.y l
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o Ms. Patricia Pierce-Bjorklund May 20,1987 I
Issue
]
2.
Increase in reactor core temperature causing the Zirconium casings on uranium rods to swell thereby blocking passages
-l between rods and preventing emergency core coolant from circulating thereby resulting in uncontrollable (sic) chain reaction and meltdown.
)
Response
The staff recognizes that swelling of the Zirconium casings on l
the fuel rod.c ay occur during a LOCA and result in blockage of the l
coolant chans.s. - Section B of Appendix K' to 10 CFR 50 requires that l
the anal effecis'ytical models used to perform LOCA analyses account for the of fuel rod swelling and rupture.
In addition, 10 CFR 50.46(b)(4) requires that the core geometry remain amenable to cooling following a LOCA. Therefore, since the LOCA analyses for the Seabrook and Pilgrim i
Nuclear Stations demonstrate chiffo'rmince to the requirements of 10 CFR 50.46, swelling of thE" fuel rods will not prevent the emergency core coolant from providing adequate core cooling in the event of a LOCA.
Sincerely, j
0 k
a, i
o Division of Reactor P
'ects I/II
{
Office of Nuclear Reactor Regulation
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L.