ML20002D317
| ML20002D317 | |
| Person / Time | |
|---|---|
| Site: | McGuire, Mcguire  |
| Issue date: | 01/16/1981 |
| From: | Curtiss J, Israel S, Johnston W Office of Nuclear Reactor Regulation, NRC OFFICE OF THE EXECUTIVE LEGAL DIRECTOR (OELD) |
| To: | Atomic Safety and Licensing Board Panel |
| References | |
| NUDOCS 8101200385 | |
| Download: ML20002D317 (38) | |
Text
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UNITED STATES OF At1 ERICA NUCLEAR REGULATORY COMf11SSI0ft y
9 BEFORE THE ATOMIC SAFETY AND LICENSING BOARD
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%c knz In the Matter of
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Docket Nos. 50-369 Id DUKE POWER C0:1PAflY 3
50-370 o
(llillian B. itcGuire fluclear Station, Units 1 and 2)
NRC STAFF ANSWERS TO CESG I!! TERR 0GATORIES AND REQUESTS FOR DOCUf1EilTS The following are the NRC Staff answers to interrogatories propounded by Intervenor Carolina Environmental Study Group (CESG) on November 29, 1980, Decenher 1, 1980 Decenber 28, 1980, and J':nuary 8,1981.
In each instance involving a request for documents pursuant to 10 C.F.R. 9 2.741, the docu-ments provided in response have been nade available in the public document roon locateri in Charlotte, North Carolina for inspection and copying.
NRC STAFF ANSUERS 1.
CESG Interrogatory:
In regard to the zirconiun-water reactinn by which hydrogen is forned, please provide the following infomation:
a.
The weight of zirconiun in the cladding for a core.
8101130038$
g
c'.
s' b.
The weight of hydrogen which would be released by total reaction of the zirconfun; the corresponding volume in cubic feet (or other volume measure) at specified conditions of temperature and pressure.
c.
The tenperature at which this reaction proceeds at a neasurable rate.
d.
The equation for the kinetics of this reaction, specifying the rate in relation to temperature for the specific thickness and configuration of zircalloy sheath versus temperature.
e.
Work papers and printouts relating extent of hydrogen reaction to time for all core conditions assuned including the low power testing case given in the Laubman [Lauben) affidavit of November 7, 1980, and nost particularly the case of about 650 days operation at full power and whatever other full power cases considered.
If different periods of continuous operation at full power were assumed, please indicate.
r IRC Staff Answer:
a.
There are 50,013 pounds of Zircaloy-4 in the cladding of the core in each of the McGuire reactors.
Approximately 987 of Zircaloy-4 is zirconiun, for an approximate total of 49,895 pounds of zirconiun.
b.
The equation for the stoichionetric reaction of zirconium and steam, including molecular weights, is as follows:
Zr + 2H 0 --> Zr02 + 2H2 2
91.22 36.032 123.22 4.032 The weight of hydrogen generated would be:
49,805 lbs. Zr x 4.032 lbs. H
= 2205.4 lbs.
H,,
7 91.22 lbs. Zr l
The nunber of mols of hyarogen would be:
2205.4 lbs. H
= 1094 nols H lbs.Hf/nolH 2.015 2
Assuming a perfect gas at 32'F and 16tnosphere there are 359 ft.3/mol.
The volume of hydrogen under these conditions would be:
1094 nols H,, x 359 ft.3 nol 392,746 ft.3H 2
c.
Although the reaction rate can be " measured" over a period of nonths at operating conditions with low temperatures, this is very insignificant as a hydrogen source. Ilhen analyzing accidents, the reaction heat source becones significant at about 1700'F.
This is also the approximate temperature at which hydrogen generation begins to be significant.
d.
The reaction equation that was used for the ficGuire hydrogen generation analysis conducted by the flRC Staff was the equation of Cathcart el al.1/ or total oxygen consunption.
Since the reaction f
is prinarily linited by gas diffusion through the oxide and since the anount of oxide is directly related to the oxygen consuned, the basic equation is as follows:
~/
Cathcart et al., Zirconiun I1etal-Water Oxidation Kinetics IV.
1 Reaction TEtTStudies OR'lL/flVREG-17, August 1977.
c,
- e dk/dt = 1/k
- /2 (1) where:
2 K = gas of 0 consumed per cm 2
t = time in seconds 2
/2 = reaction constant (gn/cn ) 2/sec.
6 Experinental detemination has led to the following equation for 2
6 /2:
26 /2 =.1311 exp (39940/RT)
(2) where:
R = gas constant (cal /gr mol/*k)
T = reaction temperature (*k) 39940 = experimentally detemined activation energy (cal /g-nol)
This describes the temperature dependence of the reaction rate.
Equation (1) is usually integrated over a small time step where the temperature can be assuned to be relatively constant. Then:
K = Ko + 6 $t (3) at = tine step (seconds) 2 The equation for 6 is usually expressed in English units and is based on the stoichionetric thickness of zirconiun consuned assun-ing a density of 6.5 gn/cc.
k
o e.
Three calculations were discussed in the Lauben affidavit of November 1, 1980. A copy of the computer printcut for these 3 calculations has been placed in the Public Docunent Roon in Charlotte, North Carolina.
(This computer printout also includes a fourth calculation at 100% power.)
In addition, the topical report for the TOODEE 2 computer progran has been placed in the Public Docunent Roon.
It should be noted that nunerous nodifi-cations and improvements in this program have been made since issuance of the topical report.
For the calculations perforned, the most significant nodification is the inclusion of a boiloff nodel. The nodel assunes saturated water in the lower plenun and core.
All water below the nixture level is assumed to boil accord-ing to the local heat source.
A bubble rise velocity of 3 f t./sec.
is assuned.
The nodel is progranned in subroutine MAINL2 between statenents f1A421881 and !!AN22020.
2.
CESG Interrogatorv:
In regard to hydrogen entering the containment atmosohere:
a.
Docunent (reference) the infomation base with regard to the conposition of hydrogen / air /stean mixtures and rate of conbustion used in Staff studies.
(What effects do temperature and pressure have on rate of propagation as a function of conposition? What are the rates of propagation?)
b.
What heat of reaction for hydrogen / oxygen combustion has been used? For the several conditions of pressure, tenperature and composition, what peak pressures were calculated for adiabatic reaction in relation to a containment with the volume assumed for
- 1) ficGuire, 2) for T 11 - 2 ? Provide work papers and equations used.
. }
. NRC Staff Answer:
Staff studies assuned that the hydrogen generated as a result of a postu-lated accident was mixed throughout the containment such that combustion of hydrogen would result in a unifomly applied pressure to the containment struc ture.
These studies are described in a document entitled SECY-80-107,
" Proposed Interin Hydrogen Control Requirements for Small Containments," a copy of which has been placed in the Public Document Roon in Charlotte, North Carolina.
Calculations perfomed _by the Staff for inclusion in SECY-80-107 used a heat of reaction of 60,000 BTV/lb.
These calculations are described in a nemo-randum from II.C. fliistead to W.R. Butler, a copy of which has been placed in the Public Document Roon in Charlotte, North Carolina.
Figures 1 and 5 of this menorandun provide estinates of adiabatic pressures for ice condenser and dry contain mts, respectively.
3.
CESG Interrogatory:
'In regard to the ignition of the containment atnosphere at T11-2, what mechanisms has the Staff considered? What conclusions has it reached? What mechanisns for ignition has it considered for ilcGuire? Provide work papers.
NP.C Staff Answer:
The Staff has considered, for T!il-2, the likelihood of both deflagrations and local detonations.
The Staff has not yet detemined the precise necha-nisn of ignition at T!11-2.
The licGuire review, which is currently underway, will consider both unifom deflagration and local detonation.
Further
e.'
discussion of TMI issues is set forth in a nemorandun from K.I. Parczewski to Victor Benaroya, dated December 23, 1980, and in a memorandum from L. Rubenstein to R. Tedesco, dated December 1,1980, copies of which have been placed in the Public Document Roon in Charlotte, North Carolina.
4.
CESG Interrogatory:
Provide the AEC or NRC statements, prior to licensing, in regard to risk of accidents and public health and safety for 1) Femi, 2) Browns Ferry, and Three !!ile Island.
NRC Staff Answer:
The Staff objects to CESG Interrogatory 4 on the ground that the infomation sought bears no relation to the McGuire licensing proceeding and does not appear to be reasonably calculated to lead to production of adnissible evidence in this proceeding.
5.
CESG Interrogatory:
Provide working papers and any others in regard to:
a.
Estinates of the ficGuire contaianent response to pressurization at, and in excess of the design level of 15 psi.
Include any infomation available on the amount of uncertainty in the safety factors for yield and for failure.
b.
Provide for T!11-2, the corresponding infomation to a. foregoing, c.
To what extent did the zirconiun/ water reaction progress at T'11-2 just prior to the hydrogen ignition.
d.
For the identical amount of hydrogen present in TF11-2 just prior to ignition, what would the hydrogen concentration be in a T1cGuire volume containment? What pressure would be reached on igniting
.e,'
this amount of hydrogen in a ficGuire volume containment? Would the containment, given McGuire design specifications, fail?
If so, give the probable nature of the fail.
NRC Staff Answer:
a.
The design pressure of the McGuire containment is 15 psig. Cased on a study conducted by the Staff's consultant, the AMES Laboratory, the mean or the expected pressure strength of the McGuire contain-nent vessel is found to be 84 psig with a standard deviation (6 )
of 12 psig. Using a variation of 36 to determine bounding values, a lower bound of 48 psig and an upper bound of 120 psig nay, therefore, he established.
Since this range of pressures is deternined on the basis of the strength of the steel without any consideration of the relation between pressure and leakage, the Staff reconnended the lower bound value of 48 psig as the realis-tic capacity of the McGuire containment under unifora static loading.
The lower bound value of 48 psig corresponds to the containment capacity at or near the yield of the steel shell. The expected value of 84 psig can be taken as the pressure at which failure of the containnent in the forn of either excessive leakage or fracture of the steel shell occurs.
Variation in parameters or uncertainty in computing the pressures at yield and at failure, are described in detail on pages 74-78 of the aforementioned AMES study, a copy of which has been placed in the Public Docunent Roon in Charlotte, North Carolina.
J
- l Since the ficGuire containnent-is designed in accordance with the AStiE code, the area replacement rule in the code is satisfied for all penetrations.
Hence the strength of the penetrated portion of the shell is greater than that of the unperforated shell.
This is the basis upon which the N1ES study rests for both the Sequoyah and licGuire containments.
For the portion of the containment which foms the boundary of the conpartment, a preliminary study has been performed for the Sequoyah containment by ATIES Laboratory, Indicating that the strength of this portion of the containment is controlled by the strength of the horizontal stiffener contained therein.
The strength of the horizontal stiffener at Sequoyah (51 psi) is greater than that of the containment shell considered as a whole (38 psi).
Sinilarly, 1
the Staff has concluded that based upon t relationship observed at Sequoyah, the strength of the horizontal stiffener is greater than that of the McGuire containment shell considered as a whole (48 psi).
b.
The Staff does not have the detailed infon,ation on TMI-2 con-tainnent as on l1cGuire.
However, based on the infomation pre-sented to the Staff by Sargent and Lundy on the Zion containment which is similar to the T!11-2 containnent and is of prestressed concrete type, the T!11-2 containnent capacities at yield and at failure are estinated as follows:
k R,
At yield 2.80 x 60 psig = 168 psig At failure 3.00 x 60 psig = 180 psig Where 60 psig is the design pressure of TMI-2 and factors 2.8 and 3.0 are based on the Zion containment with some nodification.
As to the uncertainties in arriving at the containment capacities, reference is made to tables on pages 21, 22, 23 and 24 and Figures on pages 25 and 26 of viewgraphs presented by Sargent and Lundy, copies of which have been placed in the Public Document Roon in Charlotte, North Carolina.
c.
The extent of the zirconium / water reaction that occurred at T!11-2 is not known with precision and estimates have varied as to the degree of degradation. The staff estinates that a 45% netal-water reaction reasonably characterizes the TMI-2 accident.
Various estinates have been nade of the containnent composition atmosphere prior to ignition.
The differing estimates stem from apparent discrepancies in the measured atmosphere pressure, temperature, and oxygen concentration.
The range of estimates includes pre-ignition hydrogen concentrations from 7-11 volune percent.
Further discussion of this subject is contained in NUREG/CR-1219, January 1979, as well as in a nemorandun from W.C.1111 stead to N.C. Ibsely, dated June 18, 1979, and in a nemorandun from W.R.
Butler to R.L. Tedesco, dated April 25, 1979, copies of which are i
available in the Public Document Roon in Charlotte, North Carolina.
o d.
The effect of burning a given amount of hydrogen in the McGuire plant depends on several variables:
a.
The concentration at which combustion is initiated; b.
The rate at which combustion consumes the hydrogen; and c.
The rate at which heat is removed fraa the containnent atmosphere, both through active and passive mechanisms.
In addition, one must incorporate the effects of hydrogen control features, such as deliberate ignition, before an accurate calcula-tion can be perforned.
The Staff has not, as yet, perforned the calculations described above.
6.
CESG Interrogatory:
In the event of the failure of a 'kGuire containment, what are the nost probable scenarios for post failure events? What is the likelihood of continuing boiloff of coolant and core meltdown? Provide whatever work papers Staff has generated in regard to post hydropen ignition sequences, including specifics in regard to atnosphereic releases of radionuclides.
NRC Staff Answer:
The probability of the various events leading to coolant boiloff, the genera-tion of hydrogen and the subsequent failure of containment was previously i
l I
w
0 9 estinated in the Reactor Safety Study, WASH-1400, for the Surry plant.
These estimates have now been revised and updated and applied to the Sequoyah plant (an ice condenser plant). They are to be reported in NUREG/CR-1659, Volune 1.
This report is in the process of being published and is estinated to become available in the next 90 days. A draft version of chapter 8 entitled " Accident Process Analysis" has been placed in the Public Docunent Roon in Charlotte, North Carolina.
The Sequoyah plants are upperhead injection PWRs with ice-condenser contain-nents as are the ficGuire plants. Although a detailed comparison of all their similarities and differences has not been made, they are essentially identical with respect to those features bearing on hydrogen-related accidents.
7.
CESG Interrogatory:
Is a supplemental FEIS in regard to class 9 accidents planned for !!cGuire?
If so, what is its status? What is the present targeted release date?
Provide copies of all drafts.
N NRC Staff Answer:
In accordance with the' Connission's Statenent'ef Interin Policy concerning
" Nuclear Power Plant Accident Considerations Under the Nation ~ai Environnental Policy Act of 1969" (45 Fed. Reg. 40101 (June 13,1930)), the Staff does not intend to prepare a supplemental environmental impact statenent addressing Class 9 accidents.
It should be noted, however, that this issue is currently before the Licensing Board in this proceeding for decision.
See "NRC Staff l
Response to CESG's ' Additional Contentions'" (December 15,1980).
l
r-8.
CESG Interrogatory:
Provide copies of all drafts and connents-on drafts for NUREG-0396.
Provide working papers-in relation to the range of class 9 accidents including all naterials on atmospheric releases.
MRC Staff Answer:
Copies of connents received on ilVREG-0396 have been placed in the Public Docunent Roon in Charlotte, North Carolina.
Supporting references are listed on page I-53 of flVREG-0396.
A copy of NUREG/CR-1131 "Exanination of Offsite Radiological Protective fieasures for Nuclear Reactor Accidents Involving Core 'lelt," has been placed in the Public Docunent Roon in Charlotte, North Carolina.
9.
CESG Interrogatory:
In flVREG-0396, P. I-9, "Less than 30% of all core nelt accidents would result in high exposure outside the recrended planning distances."
Iden-tify the less than 30% of sites with pt-ular reference to ficGuire and Catauba.
NRC Staff Answer:
The discussion on the identified page of NUREG-0395 pertains to "30% of all core 9elt accidents." NUREG-0395 contains no reference nor does it allude to "305 of sites," the subject identified by CESG in this interrogatory.
w
10.
CESG Interrogatory:
List, by plant and dates of occurence, departures from normal operation both prior and subsequent to the T!!I-2 accident in which there was a concern that the prinary systen would "go solid" or in which it did "go solid." Have Oyster Bay and Crystal River been involved in such incidents? If chere have been such incidents identify plant type and power rating. Provide a full account of all such incidents.
NRC Staff Answer:
Concerns about prinary systems being water solid are related to the notential for overpressurization of the vessel at low temperatures. A water solid condition orecludes hydrogen generation because the chenical reaction which nicht oroduce hydrogen will not occur when the core is filled with water.
Accordingly, the Staff objects to CESu interrogatory 10 on the ground that the infomation sought bears no relation to the contentions advanced nor is it reasonably calculated to lead to the production of admissible evident.e in this proceeding.
11.
CESG Interrogatory:
Glow plugs are being used to satisfy interin hydrogen control requirenents at Sequoyah.
Provide working papers and sunnary papers, pro and con, in relation to critiqueing and evaluating this means or hydrogen control.
Provide the cycle of operation for the glow plugs--autonatic or operator control; power supply; testing schedule. Uhat is the largest scale on which the glow plug systen has been tested? Provide all available findings.
'RC Staff Answer:
The infomation requested is set forth in Sequoyah SER Supplenents 3 and 4, cocies of which have been placed in the Public Docunent Roon in Cnarlotte,
florth Carolina. The interin operation procedure calls for actuation of the igniters by the operator after indication of a loss-of-coolant accident.
Details of the igniter systen are discussed in a docunent submitted by TVA entitled " Report on the Safety Evaluation of the Interin Distributed Igni-tion Systen." The " largest scale on which the glow olug systen has been tested," assuming this is a reference to the size of the test chanber, has been in a 1000-gallon snherical test vessel.
Results of the testing progran are described in the above-..antioned TVA report.
12.
CESG Interrogatory:
Provide the working papers on which the :lRC's position on risk, i.e. conse-quences times probability, is based.
Provide any ref.2rences from the earlier literature which were used in the reaching of this definition of risk.
NRC Staff Answer:
The flRC does not have a criterion for evaluating risk, i.e., consequences times probability, associated with nuclear power plant operation.
It is the Conaission position that past estinates ('.!AS'i-1400) of the absolute risk of reactor accidents are unreliable.
Probabilistic techniques are used to supplenent engineering judgnent where appropriate.
The_ najor study on risks associated with nuclear power plants is the Reactor Safety Study. (UAS:l-1400) which was published in final forn in 155.
The Connission issued an Interin General Statenent of Policy (39 Fed. Reg. 30964
( Aug. 27,1974)), regarding the draft version of WASH-1400, which stated M M{lMM@[
o =-
~
.m o
that "...the contents of the draft study are not an appropriate basis for licensing decisions." Subsequently, the study was reviewed by an ad hoc review group, chaired by Dr. H. Lewis, whose findings were published in
" Risk Assessnent Review Group Report to the U.S. Nuclear Regulatory Connis-sion," NUREG/CR-0400, September 1978.
On January 19, 1979, the Connission concurred with the Lewis Review Group that the Reactor Safety Study's numerical estinate of the overall risk of reactor accidents is not reliable.
Following the THI-2 accident, several groups conducted independent reviews of the accident and the design, licensing, and operation process. The Rogovin report, "Three fiile Island, A Report to the Connissioners and to the Public," reconnended increased use of probabilistic methods to identify inportant accident sequences and major contributors to these sequences.
The "NRC Action Plan Developed as a Result of the Tl11-2 Accident," flVREG-0660, contains a task for inplementing reliability and risk assessnent in the licensing process as Nell as a task for the development of a safety l
pol icy.
A proposed plan for developing an NRC position on risk is presented in NUREf-0735, " Plan for Developing a safety Goal," Oct.1980, a copy of which i
has been placed in the public document roon in Charlotte, North Carolina.
An expanded version of this proposal is currently under consideration by the Connission.
Once the Connission adopts a plan, it is expected to take several years to develop a tractable policy.
13.
CESG Interrogatory:
lihat is the NRC's current procedure for reaching the cunulative probability
-of a given type of accident? For example there are estinated probabilities for a variety of sequences which would result in a LOCA.
How are these
-conbined to arrive at the cunulative probability of any kind of LOCA.
What provision is made for possible unidentified sequences, or unencountered sequences, such as a sticking, unidentified power-operated-relief-valve?
(Pre T!11).
NRC Staff Anster:
The NRC does not have a specific procedure for calculatino the cunulative probability of a given type of accident; however, the nethods described in ilASH-1490 are generally used.
Briefly, event trees are drawn which identify the initiating condition and the success or failure of certain critical functions (events) such as reac-tivity control and core cooling.
Conbinations of these events result in nultiple possible sequences which are then judged to be either benign or serious (core nelt).
The probabilities for the sequences are then calculated by detennining the probability of each of the events in the sequence and appropriately conbining the event probabilities.
The cunulative probability for a type of accident, such as a LOCA, is the sun of the probabilities of all the sequences associated with a given category.
The potential for unidentified sequences is part of the uncertainty in the analysis and nust be considered when interpreting the results of risk assessnent.
I i
-14.
CESG I'nterrogatory:
Provide details on the equipment which Staff will approve for the real time determination of the in-containment hydrogen concentration? Has Duke proposed such acceptable equipment?
N7C Staff Answer:
Ite, II.F.1 ( Additional Accident Monitoring Instrunentation), Subpart 5 (Containment Hydrogen Monitor), of NUREG-0737, " Clarification of THI Action plan Requirelents," presents the Staff position on hydrogen monitoring capability in containments in the event of an accident.
A copy of this docuqent has been placed in the Public Document Roon in Charlotte, North Carolina.
Operating license applicants with an operating license date before January 1,1932 nust have design changes completed by January 1, 1982.
A post-implementation review will be performed prior to January 1, 1992. The Staff will report on the acceptability of the hydrogen monitoring provisions at the McGuire Nuclear Station following receipt of the appro-priate documentation fro 1 the Duke Power Company.
- 15. CESG Interrogatory:
Please nrovide the following docunents:
NSAC-1 and supplements NUREG/CR-1219 NUREG-75/014 Safety Guide 7 and subsequent guides dealing with hydrogen NP.C Staff Answer:
Conies of the foregoing docunents have been placed in the Public Docunent j
Room in Charlotte, North Carolina.
(The document referred to as NUREr,-75/014 l
i
is the sane as NASH 1400.
The document referred to as Safety Guide 7 is the saac' as Regulatory Guide 1.7).
15.
CESG Interrogatory:
Any infornation, thought, speculation which Staff has in regard to the amount and kind of particulates released during the high tenperatures phase of the zirconiun/ water reaction; during core neltdown.
Include nechanisms of particle generation and release.
Include particle size distribution.
T.C Staff Answer:
An early reoort "Reconnended Property and Reaction Kinetics Data for Use in Evaluatin, a Light-Water-Cooled Reactor Loss-of-Coolant Incident Involving Zircaloy a or 304-SS-Clad 00," (Reference 1) was prepared under U.S. A.E.C.
2 contract to critically review and assemble the physical and chenical proper-ties of cladding in contact with stean, water, and nuclear fuel, and the kinetics of their interaction.
It contains the relevant phase diagrans and rate eauations, whila 93 references nrovide an introduction to a nore thorough study of the subject.
1 ore recent treatises on zirconium enbrittlenent and the potential claddin' intera:tions during core neltdown are given in References 2, 3, 4, and 5.
Although the IRC does not require applicant analysis of core neltinwn or accident scenarios more severe than the design basis LOCA, the studies described in Reference 2 were prepared under U.S.N.R.C. contract to inves-tigate the capability of oxidized and ruptured zirconiun cladding to with-stand both thernal shock loads during core reflooding and cost-accident
u handling loads.
Experinental evidence in the report is given on the generation, size, amount, and release of the embrittled cladding particles.
7esearch (Ref. 3) underway in Gemany has addressed ultimate core degradation.
In these experinents, heated U0 fuel pellets have been allowed to react 2
with zirconium cladding. The results obtained provide new understanding on the formation and behavior of liquified fuel (i.e., UO dissolved in either 2
nolten zirconiun metal or the eutectic liquid forned between zircnniun netal and its oxide).
Results similar to these have been obtained fron U.S.
U.R.C.-funded inoile experinents (Ref. 4) in the Power Burst Facility under power-cooling nisnatch conditions.
During a loss-of-coolant accident, which progresses beyond the oresent design basis, the zirconium cladding will become embrittled as the netal surface rapidly oxidizes in the stean atmosphere.
The degree of oxidation, and resultant embrittlement, is prinarily dependent upon the netal's tenpera-ture anri the tine-at-temperature.
As time progresses, the developing oxide will serve as an increasingly nore effective diffusion Darrier to further s
oxidation. This is because oxygen atons will have to migrate across the oxide layer to reach the netal/ oxide interface.
Sinultaneously, the oxygen content in the substrate netal will be increasing. At a critical oxide layer thickness, the oxide will begin to cre.ck and possibly spall (flake) due to the lattice volune expansion that accompanies the oxidation process.
This cracking and/or spalling phenonenon will yield diffusion paths and
fresh netal-surfaces that will then allow local corrosion rates that are greater than those of the surrounding surfaces.
In spite of the above described degradation, the zirconiun-cladding will have sufficient ductility, at these elevated tenperatures, such that the rods will not catastrophically crunble, but instead will renain in their original configuration.
However, some relatively large-sized fragnents may be lost fron the fuel rod later at the time of the core reflood when the cladding is thernally shocked.
The studies described in Reference 2 found that these fragnents are generally Dieces of the full thickness of the claiding that are fractions of an inch in size.
Under very severe oxidation conditinns these cladding fragnents, together with oieces of fuel nay lose geonetry and forn a debris pile or rubble bed.
The particle sizes have been estinated to be of the order of 0.1 inch and larger.
During a hypothesized core neltdown accident, depending on the postulated scenario, the cladding and fuel tenperatures continue to rise and above 1900 C the cladding renaining in netallic forn begins to nelt and react with the fuel pellets.
Although downflow (by gravity) nay carry this liquid neltdown to the bottom of the reactor vessel, no particles are forned until the liquid refreezes in contact with water and nay fragnent or for, large solid chunks of nixed fuel, zirconiun and steel.
The anount of additional zirconiun-water reaction to take place after in.ner-sion will be a function of the.aqount of unreacted naterial, the particle
. size or dispersion that nay result as the particle contacts the water. The initial particle size will be difficult to detennine inasmuch as it is a function of its formation nechanisn (i.e., zirconium-water or zirconium-urantun interaction), prior loading history on the cladding (i.e., reflood stresses, etc.), and other variables.
It is believed (Ref. 5) that high initial netal temperatures result in a greater dispersion or a lower nean particle size, which in turn, allows a greater degree of total reaction.
References:
1.
4.C. Brassfield, J.F. Llhite, L. Sjodahl, and J.T. Bittel, "Reconnended Property and Reaction Kinetics Data for Use in Evaluating a Light-
'later-Cooled Reactor Loss-of-Coolant Incident Involving Zircaloy-4, or 304-SS-Clad 00," General Electric report GE'iP-482, April 1963.
2 2.
II.it. Chung and T.F. Kassner, Argonne !!ational Laboratory, "Enbrittle-nent Criteria for Zircaloy Fuel Cladding Applicable to Accident Situa-tions in Light '.!ater Reactors; Sunnary Report," U.S.fl.R.C. report flu' LEG /CR-1344, January 1980.
3.
S. Hagan, et al., Projekt Nukleare Sicherheit Halbjahresbericht, "Experinentalle Undersuchung dew Abschenisphize van U3 Zircaloy 2
Brennelenenten bei versagender flot Kulung," Projekt 4241, p. 416-428, Kft, 1977/2.
/.
4.
S. L. Seiffert and R.R. Habbins, Idaho National Engineering Laboratory, "0xidation and Enbrittlement of Zircaloy-4 Cladding from High Tenpera-ture Film Boiling Operation," U.S.N.R.C. report NUREG/CR-0517, April 1979.
5.
J.V. Cathcart, et al., Oak Ridge Natianal Laboratory, "Zirconiun lietal-Uater Oxidation Kinetics IV Re action Rate Studies," 0RNL report ORNL/NUREG-17, August 1977.
6.
L. Baker, Jr., and L.C. Just, Argonne National Laboratory, " Studies of
'letal-Mater Reactors at High Temperatures III Experimental and Theore-tical Studies of the Zirconiun-Uater Reaction," Argonne report ANL-6548, ltay 196?.
17.
CESG Interroaatnry:
Provide all 9 uke /NRC correspondence in relation to Duke's position to defer any decision to install a systen for hydrogen control under degraded core accident conditions pending the outcone of the degraded core rulemaking proceeding (Darrell G. Eisenhut to Willian 0. Parker, Nov. 14,1980).
Provide any records of phone conversations or ninutes of neetings.
NRC Staff Answer:
The only corresoondence in the Staff's possession relating to " Duke's position to defer any decision to install a systen for hydrogen control under degraded core accident conditions pending the outcome of the degraded core rulemaking proceeding" is a letter fron Willian 0. Parker, Jr. to Harold R. Denton, dated August 27, 1930, a copy of. which has been pl:.ed in the Dublic Docunent Roco in Charlotte, North Carolina.
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18.
CESG Interrogatory:
If there is a successor document to NUREG-0395, or drafts of such a document, please provide.
NRC Staff Answer:
There is no successor document to NUREG-0396, either in draft or final form.
- 19. CESG Interrogatory:
In tenis of present Staff siting criteria, please indicate the consider-ations vis-a-vis ficGuire, bearing in nind the consequences of class 9 accidents including atmospheric release due to containment breach.
NRC Staff Answer:
The present Staff siting criteria are set forth in 10 CFR Part 100.
With respect to the McGuire Nuclear Station, the Staff reviewed and reported on the confornance of the site with respect to 10 CFR Part 100 in the Staff Safety Evaluation Report (NUREG-0422,11aren 1978), in Section 2.1, and concluded there that the site net the criteria of Part 100 and was, there-fore, acceptable.
In regard to the consequences of accidents and siting criteria,10 CFR Part 100 requires that the reactor site plus plant design features be such i
that the consequences to an individual located at the exclusion boundary or at the outer boundary of the low population zone (LPZ) from the largest l
design basis accident postulated, he within the dose guidelines given in
Part 100.- This design basis accident, while postulating a radiological release inside containment that is connensurate with a degree of core nelt, does not consider the breach of-containnent.
However, Part 100 also requires that the nearest population center be no closer than one and one-third times the outer radius of the LPZ. With regard to this criterion, the Statenent of Consideration accompanying the oro1olgation of Part 100 (27 Fed. Reg. 3509 ( April 12,1962)) states that:
Further, since accidents of greater potential hazard than those connonly postulated as representing an upper limit are conceivable although highly innrobable, it was considered desirable to provide for protection against excessive expo-sure doses to people in large centers, where effective oro-tective neasures night not be feasible.
Neither of these nhjectives were readily achievable by a single criterion.
Hence, the population center distance was added as a site requirenent when it was found for several projects evaluated that the specification of such a distance requirement would aporoximately fulfill the desired objectives and reflect a nore accurate guide to current siting practices.
Thus, the population center distance criterion in Part 100 does reflect a consideration of accidents beyond the design basis.
For the McGuire Nuclear Station, as stated in Section 2.1 of the Staff's Safety Evaluation Report, the outer boundary of the LPZ is 5.5. niles, while the nearest ponulation center, Charlotte, North Carolina, is about 11 niles away fron the plant.
The distance to the nearest population center is nore than one and one-third times the LPZ radius, and, therefore, confon1s to the guidelines of 10 CFR Part 100.
i l-
llith respect to explicit consideration of Class 9 accidents, see the NRC Staff response to CESG Interrogatory 7.
20.
CESG Interrogatory:
In Figure I-18, NUREG-0395, what linear wind velocity was assuned? What were the widths of the plune at 5,10,15, and 25 miles? How many people were exposed in the respective intervals?
f1RC Staff Answer:
No single wind velocity was assumed for Figure I-18.
Rather, a conposite of meteorological data was utilized, a detailed explanation of which is contained in Aopendix VI to " Reactor Safety Study: An Assessnent of Accident Risks in U.S. Connercial Nuclear Power Plants," WASH-1400, (NUREG-75/014), October, 1975.
The plume width is a function of:
(1) stability class of atmospheric conditions; and (2) downwind distance.
Stability class information is presented on pages I-14 through I-16, NUREG-0396.
Figure I-?, NilREG-0396, can be used to detemine plune width, as a function of stability class, for downwind distances up to 50 miles.
The average number of people exposed in a radial interval is a function of the.3rea of that interval and the population density within that area.
For a plume sector of 22h', the area of an interval is given by the equation:
2 A=oi (R
- R,,2)/16 1
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where R and R are the maximum and minimum, respectively, radii of the 3
2 interval. As stated in Figure I-18, a unifom population density of 100 persons per square mile was assumed. Therefore, the average number of people exposed in an interval is equal to "A" (in square miles) multiplied by 100.
- 21. CESG Interrogatory:
Provide infomation on the quantitative effect of wind velocity on plume width.
NRC Staff Answer:
The atmospheric stability class detemines plume width.
The stability class is a function of wind velocity.
The stability class can be determined fron Figure I-1, NUREG-0396.
Plume width can be detemined from Figure I-2, NUREG-0396.
22.
CESG Interroaatory:
What are the effects on dosage of plume direction reversal? What is the effect of stagnation on plumes of different ages in respect to time of release?
If they exist, provide papers giving examples.
NRC Staff Answer:
Plume direction reversal is possible in areas affected by terrain-induced flow conditions.
A characteristic example of such a condition is the land /
sea breeze effect in a coastal environment.
The distribution of concentra-tion of material laterally across a plume front increases fron a negligible L
j
O level at the edge of the plume to a maximum at the plume centerline and then decreases back to a negligible level.
It is possible, though highly unlikely, that a person could be fixed at a point during plume passage and reversal such that he would experience the relative peak centerline concentration twice. The atmospheric transport and diffusion models used for dose projec-tions required by NUREG-0654 account for such terrain-induced flows.
Under stagnant conditions, a person within a slow-moving plume is likely to receive a greater exposure than a person within a fast-moving plume.
A stagnant plume centerline is likely to meander so that a person at a fixed point would not likely be constantly exposed to the peak centerline concen-tration. As plume transport time increases for a discrete parcel of air, the plume ages and plume depletion and radioactive decay become more sig-nificant, with the result that a person within the stagnant plume experiences a lower exposure rate.
23.
CESG Interrocatory:
What are the "ates of alpha, beta and gamma decay of the mixed particulate and volatiles release assumed in the atmospheric release of NUREG-0396?
What is the composition, identifying the nuclides and transuranics in par-ticulates and the nuclides ir the gaseous component, assumed for the initial release? How much decay, in relation to operating core conditions, has been assumed at the time of release?
. NRC Staff Answer:
The infomation requested by this question is provided in Tables VI-2-1 and VI-3-1 of " Reactor Safety Study: An Assessment of Accident Risks in U.S.
Commerical Nuclear Power Plants," WASH-1400, (NUREG-75/0614), October, 1975, a copy of which is available in the Public Document Room in Charlotte, North Ca roli na.
24 CESG Interrogatory (second round of discovery, question 1):
In NUREG-0396, Figures 1-17 and I-18 refer to protective actions including sheltering with different shielding factors for different times.
a.
What transit exposure is considered on leaving shelter?
b.
What provision does the NRC contemplate for persons after leaving shelter?
In the case of a severe atmospheric release, how many in the path will be able to promptly return to their residence? What provision is envisaged for persons not able to return? If the path of the release is over the most populous area, how many people will be displaced from their residences? How long will they be displaced?
c.
If there are pemanent displacements of people, what planning does the NRC have in mind?
d.
What are the NRC's decontamination plans for land, water, and chattels in the event of a substantial atmospheric release?
NRC Staff Answer:
The reference document for Figures I-17 and I-18, NUREG-0396, is NUREG/CR-1131, " Examination of Offsite Radiological Emergency Protective Measures for Nuclear Reactor Accidents Involving Core Melt." As stated on page 56 of NUREG/CR-1131, people are not assumed to receive a specific exposure while relocating.
As stated in the footnote, however, an effective exposure tire
of 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> while sheltered (with Shelter Factor (SF) for ground contamination
= 0.2) might, in fact, be due to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> of exposure while sheltered (SF=0.2) and 1/2 hour while relocating (SF=0.8).
The Federal Emergency Management Agency (FEMA), as directed by the President in his statement of December 7,1979, is the lead agency for all emergency activities associated with the offsite planning and response for peacetime radiological accidents at nuclear reactor facilities.
In November 1980, NUREG-0654, FEMA-REP-1, Revision 1 " Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Pcwer Plants," was published jointly by NRC and FEMA.
Section J of this document is entitled " Protective Response" and includes criteria for State and local agencies to register and monitor evacuees at relocation centers in host areas which are at least 5 miles and preferably 10 miles beyond the boundaries of the plume exposure emergency planning zone.
In the event of a severe atmospheric release, the number of people who would be instructed to evacuate and the length of time before they could return home would depend not only on the nature and quantity of the release, but also the meteorological conoitions at the time of the release.
Decisions for recovery and reentry are a State responsibility and would be made on the basis of monitoring by licensee, State and federal authorities.
Section J also requires State plans to identify criteria for putting dairy animals on stored feed, and to identify procedures for detecting contanination, for estimating the dose consequences of uncontrolled ingestion, and for imposing
y protection procedures such as impoundnent, decontanination, processing, decay, product diversion and preservation.
25.
CESG Interrogatory (second round of discovery, question 2):
P' lease provide a copy of flVREG-0611; FEt1A's Report to the President, June 1980, re State Radiological Energency Planning; NUREG-0654; and, if still available, the 2nd volune of WASH-1400; flVREG-0093/1 (Jan.1979); EPA Policy Statenent 44 Fed. Re9 2893.(Jan. 15, 1980); Fact Sheet:
The President's Connission on the Accident at Three itile Island (Dec. 7,1979); ilVREG-0623; flVREG-0684; flVREG/CP-0011 ( April 1980); Regulatory Guide 1.101 (nost recent);
' lure'l-510 (Sept.1979); Potassiun Iodide, etc, 45 Fed. Reg.11912 (Feb. 22, 1980); flVREG/CR-1433 (flar.1980); SAflD 80-0981 (flar.1980); Accidental Radioactive Contanination, etc. 43 Fed. Reg. 58799 (Dec.19,1979); Devel op-nent Plan, etc. 44 Fed. Reg. 75344 (Dec.19,1979); flanual of PAG's, etc.,
EPA-320/1-75-001 (June 1979 revision); SAPID-1725; SAtlD-78-0454; (sane as flVREG/CR 1131 ?); EPA 520/1-78-001B ( Apr.1979); NUREG-0553 (Oct.1979);
flVREG/CR-1225 (Jan. 1980); Cost and Funding, etc., 44 Fed. Reg. 64929
('lov. 8, 1979) and 44 Fed. Reg. 66113 (Nov. 16, 1979); and Final Rulenaking, etc., SECY-30-275 (June 3, 1980).
4RC Staff Answer:
Copies of all docunents, with the exceotion of " Fact Sheet:
The President's Connission on the Accident at Three flile Island (Dec. 7,1979)," are available in the Public Document Roon in Charlotte, North Carolina. The iPC Staff has been unable to locate the Fact Sheet referred to.
25.
CESG Interrogatory:
Provide the records in regard to containnent tenperature, wherever neasured, during the first 7 days of the accident at Tl11-2.
flRC Staff Answer:
The record for containment tenperature during the first 17 hours1.967593e-4 days <br />0.00472 hours <br />2.810847e-5 weeks <br />6.4685e-6 months <br /> (following turbine trip) of the TI11-2 accident, as well as an enlarged record of hours om o
g-J ook a N
. S hlih i
l 9-11 involving hydrogen ignition, have been placed in the Public Document Room in Charlotte, North Carolina. The NRC Staff has requested complete containment temperature records for the first 7 days of the accident and will provide these records promptly upon receipt.
27.
CESG Interrogatory:
For the McGuire containments, provide the psig value calculated at yield; at rupture.
Provide any variance information you have in connection with these values.
Provide any infomation you have for the sensitivity of these values to the rate of pressure increase.
Does the NRC have any information of the effect of the shock waves that result from the explosion of a hydrogen /
air mixture in a closed vessel like a containment? Have reflection effects been considered? Has focusin curved (cylindrical section) g of the shock wave by the reflection from the inner surface of the containment been considered?
If such consideration has been given, what are the outcomes?
NRC Staff Answer:
The infomation requested regarding containment capacity at yield and at rupture has been provided in the NRC Staff Response to CESG Interrogatory 5(a).
The pressure values referred to in this response are based on static pressure.
The Staff has not performed any dynamic pressure analysis.
Depending on the time history of the pressure, the response of the contain-ment may ' e negligible, eauivalent to that of a static pressure, or more o
severe tqan that of a static pressure.
The reflection effects of an explo-i sion will change the peak of the pressure time history, but will not affect j
the containment capacity.
l
28 CESG Interrogatory:
Provide the precise dates of issuance of:
the Virgil C. Summer Draf t EIS, CP stage Final EIS, CP stage Draft EIS, OL stage Final EIS, OL stage Supplemental Draf t EIS, NUREG-0534 NRC Staff Answer:
Draft EIS, CP Stage - Sept. 1972 Final EIS, CP Stage - Jan. 11, 1973 Draft EIS, OL Stage - June 29, 1979 Final EIS, OL Stage - Not yet issued Supplemental Draf t '
tEG-0534 - Nov. 10, 1980
- 29. CESG Interrogatory (fourth round of discovery, question 1):
Reference.
NUREG/CR-0400, p. 42, last paragraph.
Item:
Document 1, in entirety if reasonably convenient.
Pp. 180-212 as a minimum concerning PWR/BWR QA/QC procedures.
NRC Staff Answer:
The NRC Staff is unable, without a more precise reference, to obtain the document referred to in CESG Interrogatory 29, "the complete (but undated) write-up of the PWR/BWR QA/QC procedures."
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I f
,1 lN d
n SafWord IsNel I
(Interrogatories 12, 13)
\\/Lv n William V. Johqston (Interrogatory 16)
Sworn to before me this 16th day of January, 1981 W} /d!w l
V
'N&tary Pu p ic /
My Commission fxpires:
M, /, /([h J'
/ '
/def Ja es R. Curtiss
( nterrogatori;+, 7, 15, 25, 28, 29, jections to Interrogatories 4,10) i 1
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I hereby certify that the information detailed above is true and accurate to the best of my personal knowledge
-.k 1-_ _
Thomas A. Kevern (Interrogatories 8, 9, 18, 20, 21, 22,24,24) mb
-au" 4A Ja()es C. Pulsipher
/
(Interrogatories 2, 3, Sc, 5d, 11, 14, 17, 26)
Alw ( k Chen P. Tan (Interrogatories Sa, 5b, 27) b b h, L - 1 &
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riorman LauDen (Interrogatory 1) r
'hM Joh l K. Long (In terrogatory 6)
W Walton L. JensF1 (Objection to Int rogatory 10) b QW
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Lebnard Soffer Ii (interrogatory 19)
'JNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Mater of DUKE POWER COMPANY Docket Nos. 50-369 50-370 (William B. McGuire Nuclear Station, Units 1 and 2)
)
CERTIFICATE OF SERVICE
.s i hereby certify that copies of "NRC STAFF ANSWERS TO CESG INTERR0GATORIES AND REQUESTS FOR DOCUMENTS" dated January 16, 1981, in the above captioned proceeding, have been served on the following, by deposit in the United States mail, first class, or, as indicated by an asterisk through deposit in the Nuclear Regulatory Commission's internal mail system, this 15th day of December, 1980:
- Robert M. Lazo, Esq., Chairman Mr. Jesse L. Riley, President Atomic Safety and Licensing Board Carolina Environmental Study Group U.S. Nuclear Regulatory Commission 854 Henly Place Washington, D.C.
20555 Charlotte, North Carolina 28207
- Dr. Emmeth A. Luebke J. Michael McGarry, III, Esq.
Atomic Safety and Licensing Coard Debevoise & Liberman U.S. Nuclear Regulatory Commission 1200 Seventeenth Street, N.W.
Washington, D.C.
20555 Washington, D.C.
20036 Dr. Cadet H. Hand, Jr., Director W. L. Porter, Esq.
Bodega Marine Lab of California Associate General Counsel P.O. Box 247 P.O. Box 2178 Bodega Bay, California 94923 422 South Church Street Charlotte, North Carolina 28242
- Atomic Safety and Licensing Board Panel U.S. Nuclear Regulatory Commission Washington, D.C.
20555 i
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- Atomic Safety and Licensing Appeal Panel U.S. Nuclear Regulatory Commission Washington, D.C.'
20555 l
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- Secretary.
U.S. Nuclear Regulatory Comission l
ATTil: Chief, Docketing & Service Br.
Washington, D.C.
20555 l
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M James R. Curtiss l
Counsel for NRC Staff 1
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