ML20010F956
| ML20010F956 | |
| Person / Time | |
|---|---|
| Site: | Diablo Canyon  |
| Issue date: | 09/09/1981 |
| From: | Hubbard R CALIFORNIA, STATE OF, MHB TECHNICAL ASSOCIATES |
| To: | |
| Shared Package | |
| ML20010F955 | List: |
| References | |
| ISSUANCES-OL, NUDOCS 8109150155 | |
| Download: ML20010F956 (31) | |
Text
.-
EXHIBIT l' s
UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE APPEAL BOARD
~
)
In The Matter'Of
)
)
PACIFIC GAS AND ELECTRIC COMPANY)
Docke t Nos. 50-275 0.L.
)
50-323 0.L.
(Diablo Canyon Nuclear Power
)
Plant, Units Nos. 1 and 2)
)
AFFIDAVIT OF RICHARD B. HUBBARD EICHARD B. HUBBARD, being duly sworn, deposes and says as follows:
1.
The purpose of this affidavit is threefold.
First, to es tiaate the elapsed time which is likely to be required af ter i
issuance of a low power operating license to load fuel and to complete the special low power tests at or below 5% of Rated Thermal Power as Pacific Gas and Electric Company has proposed I
for the Diablo Canyon Unit 1; second, to describe the subs tantial fission product inventory that would be created in less than one month of 5 percent power operation; and third, to identify the technical difficulties and increased costs associated with modifying the s tructures, sys tems, and components o f the plant should further modifications be required aftpr fuel has been loaded and operation commenced. A recent statement of my profes-sional qualifications and experience is attached hereto as Appendix A-8109150155 810911 POR ADocK 05000275 o
2.
In preparing this affidavit, I have reviewed PG5E's proposed special low power test program as set forth in the low power license application and as further described in
~
PG5E's safety analysis report provided to the NRC Staff on February 6, 1931.
I also attended, as a consultant to Governor Brown's counsel, all sessions of the recent low power test pro-ceedings which were b71d in San Luis Obispo from May 19 tc May 22, 1981.
Thus, I am familiar with the <?aration of the low power tests as postulated by PG5E and Staff witnesses.
- Further, I have reviewed the actual schedule for fuel loading, initial criticality and zero po,wer tcsting, and low power testing of large pressuri:ed water reactors (PWR's) which have occurred in the pos t-TMI period, particularly North Anna-2, Salem-2, and S equoy ah -l.
In addition, on July 10, 1981, I accompanied NRC
~
Commissioner Gilinsky on his tour of the Diablo Canyon acility.
The results of my review are summarized in the following paragraphs.
A.
INITIAL CRITICALITY AND DURATION OF LOW POWER TEST PROGRAM 3.
During Commissioner Gilinsky's tour of the Diablo Canyon 1
facility, both NRC and PG5E personnel emphasi:ed PG5E's readi-ness to load fuel.
The necessary fuel is presently on site in a building immediately adj acent to the Containment Building.
Further, d.ue to the duration of the licensing process, PGSE I
has had sufficient time to conduct, and in some cases reconduct, ;
=.
_= -
its pre-operational tests as set forth in Section 14.1 of the Final Safety Analysis Report (?FSAR").
Thus, I conclude that Diablo Canyon Unit i equipment is in an advanced s tate of readi-
~
ness to load fuel, and that virtually all preliminary testing 1
such as that described in the FSAR Table 14.1-1 possible prior to fuel loading has been completed.
/
Further, I concluda that PBEE should be able to promptly load fuel once such authoriza-tion is received from the NRC.
l 4.
I estimate that the fuel loading task should be com-pletsd in less than one week elapsed time.
For example, at Salem-2, a Westinghouse-designed PWR similar in design and rating to Diablo Canyon, fuel loading began on May 23, 1980 and was com-pleted on May 27, 1980.
Following fuel loading, the Precritical Test Program of eleven tests, as set forth by PGSE in Table 14.1-2 of the Diablo Canyon FSAR, should require no more than two weeks to complete.
Thus, there is no technical reason that initial criticality could not be achieved within two weeks af ter fuel loading is completed.
Therefore, I conclude that it is reasonable to expect that the fuel.oading and precritical tes t program could be completed in no more than 30 days af ter the -
issuance of a low power test license.
The reactor could be made critical immediately thereafter.
- / A recent Nucleanics Week article indicated that all s teps
~
prior to fuel load will be completed by approximately Augus t 12, 1981 (p.
4, July 23, 1981).
In general, all pre-operational testing will be completed before fuel loading (FSAR, p.14.1-8).
P 5.
The next phase of startup and testing includes ini-tial criticality (i.e., commencement of the nuclear reaction) and testing (oi' the reactor at power levels up to 5 percent of rated capacity).
FSAR Table 14.1-2 summarizes the normal tests which will be performed.
In addition, the scope and duration of the special low power tests were described in detail during the recent low power proceedings in San Luis Obispo.
The Licensing Board, in the Partial Initial Decision dated July 17, 1981, noted at page 24, paragraph 61, that PGSE has proposed a series of eight special low power tes ts.
The proposed tests would probably last for no more than one month and in actuality, as cited by the Board, would perhaps only take about eighteen days (Tr. 10,826-10,728).
Other references to the "relatively few days" encompassed by the proposed low power test program are set forth in the recent decision by the Board a+. page 25 (paragraph 65), page 32 (paragraph 82), and page 33 (paragraph 83).
The re fore, I believe that it is l
reasonable to expect that, absent major problems or absent dis-cretionary delay by PG4E (for ins tance, to conduct some other tes ts), initial criticality can be achieved and low power testing can be conducted in an elapsed time of less than 30 days.
- Thus, assuming a 30-day period for fuel loading and precritical testing, the entire fuel load and tes ting program can readily be completed i
in no more' than 6] days, i
l l
4 I
6.
The reasonableness of a 60-day cycle from license, issuance to completion of the special low power tests was fur-trer confirmed during Commissioner Gilinsky's tour of the Diablo Canyon facility.
In response to a question, the Diablo Canyon Plant Manager, Robert C. Thornberry, stated in my presence that PGSE's current schedules forecast that fuc1 loading, zero power testing, and the special low power test program will be completed approximately 58 days after receipt of a low power license.
Mr.
Thornberry added ihat the schedule might need to be increased if major unanticipated problems were encountered during the test program.
7.
In order to be conservative, I believe it may be appro-priate to add 15 to 30 days to the fuel loading and low power testing schedule to allow t: e for resolution of any routine unanticipated events.
Thus, at the outside, I would expect the entire low power program at Diablo Canyon to take no more than 90 days.
I understand that the NRC Staff recently indicated that the entire program would be completed in 101 days, which I feel is consistent with the schedule set forth herein.
/
8.
The post-TMI experience and the current schedules for startup testing lend f arther support to the preceding conc!u-sions.
The firs t plant granted an operating license in the post-TMI period was Sequoyah-1, which received a low power
- / See Attachment to Trcnscript of NRC Commissioner Briefing of August 27, 1981. -
o license on February 29, 1980.
Fuel loading commenced on March 2, 1980 and was completed on March 8, 1980.
Two major problems thereafter seriously delayed the initial criticality of Sequo-yah-l.
First, in response to ISE Bulletin 79-14, TVA required approximately 60 days to inspect mad rework pipe hangers and s upp orts.
Second, in parallel with the hanger reinspection, TVA conducted a base line inspection of the turbine blades.
The turbine reinspection required 4-5 weeks of elapsed time.
Routine maintenance proble'ms and pre-operational testing resulted in further delays.
Initial criticality was achieved on July 5, 1980.
Following zero power tes ting, the special low power testing program began on July 12 and was completed on July 18, 1980.
9.
The second plant to receive a post-TMI license to load fuel and conduct special low power tests was North Anna-2.
The authorization to load fuel was issued on April 11, 1980 and the low power tes ting was completed by July 1, 1980, an elapsed time of less than 80 days.
10.
The Salem-2 low power license was issued on April 18, 1980.
As set forth in paragraph 4, fuel loading was completed on May 27, 1980.
Initial criticality was achieved on Augus t 2, 1930.
The two months delay between fuel loading and initial criticality was largely due to the need to conduct routine pre-operational maintenance testing and surveillance testing (such as valve operability) which could have been accomplished prior to fuel load.
As presented in paragraph 3, I believe that these pre-operational tests will be accomplished at Diablo Canyon prior to fuel loading. ~Thus, I conclude that the actual duration of the Salem-2, North Anna-2, and Sequoyah-1 fuel loading and low power testing programs is not inconsistent with my conclusions for Diablo Canyon as set forth herein.
B.
FISSION PRODUCT HAZARD
- 11. There is sufficient evidence in the record of the recent low power test proceeding to show that the consequences of a severe accidental release during low power operation would be serious.
The basis for my views are as follows:
First, Table I of the testimony of Applicant's witness, Dr. Brunot, sets forth the fission product inventories which will be produced in the core during the proposed Diablo Canyon LPTP.
The inven-tory of iodine-131, one of the radionuclides which is a signifi-cant contributor to the dominant exposure modes for accidents requiring off-site emergency preparedness, is estimated by Dr.
Brunot as 4,500,000 curies (approximately 1/20th the full power value as set forth in FS AR Table 11.1-4).
In contras t, for the design basis LOCA addressed by the Applicant in the FSAR, only 192 curies of iodine-131 were postulated to be released to the environment in the first two hours.
The corresponding two-hour thyroid doses cited in the FSAR are as follows:
Activity /
Thyroid Doses (Rem) * */
Released
- S00 10,000 Nuclide (Curies)
(Meters)
(Meters)
I-131 27.0 7.3 0.3 I-131 ORG 73.4 19.9 0.8 I-131 PAR 91.8 24.9
1.0 TOTALS
192.2 52.1 2.1 12.
Furthermore, in the Diablo Canyon Emergency Plan ***/
the Applicant has calculated that if the equivalent of 1000 curies of iodine-131 were to be released during e " Site Emergency" sens; class accident, and assuming the design basis meteorological conditions, then the thyroid dose at the plume centerline would be as follows :
Activi ty Thyroid Doses (Rem)
Released 600 10,000 Nuclide (Curies)
(Meters)
(Meters)
I-131 1000 270 12 l
l The preceding relationships between releases and exposures are all based on numbers in the record in the low power proceeding.
l By observation, it can be inferred that the thyroid doses can l
- /
FSAR Table 15.5-12 (attached hereto as Appendix 3).
T*/
FSAR Table 15.5-14 (attached hereto as Appendix C).
TT*/
Emergency Plan, p. 4-5 (attached hereto as Appendix D).
- /
The release potential and significance for a larger class of accidents, the " General Emergency," were not quantified l
by the Applicant in the Diablo Canyon Emergency Plan.
i 1
be scaled approximately linearly with fission product releasei;.
e This relationship is not surprising in that Dr. Brunot states in his testimony that estimated exposure is directly proportional to the core invenfory which could contribute to that exposure.*/
(We believe he mus t be assuming a cons tant release fraction).
Brunot further estimated exposure levels by scaling exposures linearly based on the reduced fission product inventories at LP as compared to the FP operation. **/ Thus, using the Brunot scaling methodology, and assuming release fractions of 1.0 percent or 0.1 percent, the exposures for an accident during the Diablo Canyon LPTP can reasonably be extrapolated.approximately as follows:
Activity Thyroid Doses (Rem)
Released 800 10,000 Nuclide (Curies)
(Meters)
(Meters)
I-131 4,500 (0.1%)
1,221 49 I-131 45,000 (1.0%)
12,211 492 In either of the preceding cases, the potential thyroid exposures appear to be of significant magnitude.
Thus, the next question l
is whether the postulated release fractions are reasonable.
l 13.
The probabilities for nine major PWR release categor-ies (PWR-1 to PWR-9) were developed in the NRC's Reactor Safety Study (WASH-1400). ***/ The event sequences in PWR-1-7 lead to
- /
Bruno t Tes timony, p. 11.
- /
Bruno t Tes timony, p. 12.
I
- /
The dominant PWR accident sequences from NASH-1400 for each of the release categories are set forth in Appendix E which is attached hereto.
1 i
.g_
partial or complete melting of the reactor core while those in the las t two categories do not involve melting of the core.
These severe accidents can be distinguished from design basis accidents in that they involve deterioration of the capability of the containment structure to perform its intended function of limiting the release of radioactive materials to the environ-ment.
In release categories 1 to 3, the event sequences include cont ainmc - t failure by steam explosion, hydrogen burning, or overpressure.
In release categories 6 and 7, t'le dominant con-tainment failure mode is by melt-through of the containment base mat.
The other release categories contain event sequences in which the systems intended to isolate the containment fail to act properly.
The uncertainties in the absolute values of th',
prcbabilities are significant.
The error band for the probabili-ties of some of the event sequences could be as great as a factor of 100 as discussed by Staff witness Lauben in the low power pro-ceeding.
The containment releases postulated in WASH-1400 are described in more detail in Appendix F which is attached hereto.
It is important to note that the magnitudes (curies) of radio-active releases for each PWR category are obtained by multiply-ing the release fractions shown in Table VI 2-1 of Appendix F by the amounts of radionuclides that would be present in the core at the time of the hypothetical accident (for Diablo Canyon LP inventory, see Table I of Brunot testimony).
For example, if one started with the iodine-131 inventory of 4.500,000 curies calculated by Brunot and the release fractions set forth by the WASH-1400 authors, the magnitude of the iodine releases for each of the nine PWR accidents, if it occurred during the proposed Diablo Canyon LPTP, would be as follows:
PWR Activity Release Release Rele ased Cate go ry Fractions (Curies) 1 0.70 3,150,000 2
O.70 3,150,000 3
0.20 900,000 4
0.09 405,000 5
0.03 135,000 6
8x10-4 3,600 7
2x10-5 go 8
1x10-4 450 9
1x10-7 0.45 14.
Several conclusions are obvious.
First, the 1.0%
l release fraction postulated herein is exceeded by a factor of 3 to 70 for WASH-1400 release Categories 1 through 5.
The 0.1%
release is consistent with a Category 6 release occurring during LP operation.
Thus, I conclude that the proposed 1.0% and 0.1%
release fractions are conservative representations of the poten-tial releases. */ Therefore, because of the relatively rapid buildup (half-life of hours to days) of the radioactive isotopes
- /
Indeed, the NRC indicated recently that the possession of as little as 3.3 curies of I-131 constitutes a sufficient amount to be "of potential significant concern in the event of a major accident....." 46 Federal Regis ter 29714 (June 3, 1981). The l
I-131 inventory af ter one month of low power operation of i
Diablo Canyon will be 4. 5 million curies, or more than one million times greater than the JRC's recently s tated thres-hold level of concern..
- - - _ ~
---o-
listed in Table 3 of NUREG-0654*/ which dominate prompt health consequences resulting from postulated accidental releases, I conclude that even at 5% power after less than 30 days the fis-
~
sion products avai1able for release pose a significant potential ha ard.
C.
PLANT CONTAMINATION
- 15. Operation at low power will not only cause a buildup of fission products within the reactor core, making it inaccessi-ble for contact r6 pair and/or modification,but will also cause a spread of radioactive contaminants throughout the primary portion of the steam supply system.
It wi:1 also contaminate certain auxiliary systems such as the Chemicci and Volume Control System, Equipment and Floor Drainage Sys tems, and the Liquid Radioactive Was te Sys tem.
If fuel failures and/cr steam generator tube fail-ures or leaks are experienced, a 1.arge number of other systems, including the turbine, condensate, and other components within l
the Steam and Power Ccnversion System could become contaminated.
Contamination and irradiation of such equipment greatly increases the care required and the time and cost of future modifications that could be required at Diablo Canyon.
I conclude, there fo re,
that it is important that power operation, including low power tes ting, not be permitted until reviews and evaluations that could lead to required plant modifications have been completed.
- / NUREG-0654, Re v.
1 (FEMA-REP-1), Criteria for Preparation and and Evaluation of Radiological Emergency Response Plans and
~
Preparedness in Support of Nuclear Power Plants, Novemoer, 1980. - -.
e r
D.
CONCLUSION
- 16. Based on the foregoing, I conclude: (a) that fuel loading, initial criticality, and low power testing, including
~
the special low power tests, can be accomplished at Diablo Canyon Unit 1 within approximately 60 days, with an outside maximum elapsed time of approximately 90 days, after issuance of the low power operating license; (b) that it is feasible for fuel loading to be completed within one week a'fter issuance of the low power license ; mnd (c) that t:te fuel loading and pre-critical testing portion of the startup schedule should be com-pleted within less than 30 days following issuance of the low power license.and that immediately thereafter initial criticality could be achieved.
Further, I conclude that because of the relatively rapid buildup of the radioactive isotopes which dominate health consequences, even at 5% power the fission products such as iodine-131 available for release pose a signifi-cant hazard.
Finally, I conclude that operation at low power will contaminate some of the facility's components and systems.
This unnecessary commitment of resources creates technical diffi-culties and increased costs associated with modifying the reactor, should further modification be required after fuel has been loaded and power operation commenced.
nh.
4 d
I have read the foregoing and swear that it is true and l
accurate to the bes t of my knowledge.
i kY"'A $,
h-RICHARD B. HUBBARD c7 Subscribed and sworn to before me this
/
day of September, 0
l 1981.
i A-
!,,=;;:/
~ Mgg"if']
NOTARI'7dBLIC
{
O 1
JAMES F LEHMAN
/
3 24 actAAY ruauc cAurca.wa -
SANTA CIAV COUN1Y Commission expires E/
r[j-~
f4 c;mm. exp.r:s AUG 21, 1934 m% # -
- --v w v,
t i
i I
l
. I
APPENDIX A PROFESSIONAL QUALIFICATIONS OF RI CH ARD B. HUBBARD RICHARD 3.
HUBBARD MHB Technical Associates 1723 Hamilton Avenue Suite K San Jose, Calif ornia 95125 (408) 266-2716 E XP E RIEN CE :
9 / 76 - P RESENT Vice-President - MHB Technical Associates, San Jose, California.
Founder, and Vice-? resident of technical consulting firm.
Special-is ts in independent energy assessments for government agencies, particularly technical and economic evaluation of nuclear power facilities.
Consultr.nt in this capacity to Oklahoma and Illinois Attorney Generals, Minnesota rollution Control Agency, German Ministry for Research and Technology, Governor of Co lo rado, Swedish Energy Commission, Swedish Nuclear Inspectorate, and the U.S.
Department of Energy.
Also provided studies and testimony for various public interest groups including the Center for Law in the Public Interes t, Los Angeles; Pu blic Law Utility Group, l
Baton Rouge, Louisiana; Friends of the Earth (F0E), Italy; and the Union of Cc:cerned S cientis ts, Cambridge, Massachusetts.
Provided t e s time.ty to the U.S. Senate / House Jsint Committee on l
Atomic Energy, the U.S.
House Committee on In terior and Insular Affairs, the California Assembly, Land Use, and Energy Committee, the Advisory Committee on Reac tor S af e guards, and the Atomic Safety and Licensing Board.
Performed comprehensive ris k analys is of the accident probabilities and consequences at the B ars eback Nuclear-Plant for the Swedish Energy Commission and edited, as well as contributed to, the Union )f Concerned S cientis t's technical review of the NRC's Reacto r S af ety Study (WASH-1400).
2/76 9/76 Consultant, Project Survival, Palo Alto, California.
Volunteer wo rk on Nuclear Saf eguards I ni tia ti ve campaigns in Cali-fcrnia, Oregon, Washington, Arizona, and Colorado.
Numerous presentations on nuclear power and alternative energy options to civic, g o v e'r n m e n t, and college groups.
Also resource person for public service presentations on radio and television.
A-1 1
5/75 - 1/76 Manager - Quality Assurance Section, Nuclear Energy Control and Instrumentation Department, General Electric Company, San Jose, California.
Report to the Department General Manager.
Develop and implement quality plans, programs, methods, and equipment which assure that products produced by the Department meet quality requirements as defined in NRC regulation 10 CFR 50, Appendix 3, ASME Boiler and Pressur; %s s el Code, customer contracts, and GE Corporate policies and procedures.
Product areas include radiation sensors, reactor vessel incarnals, fuel handling and servicing t*ols, nuclear plant control and p ro tec t ion instrumentation systems, acd nuclear steam supply and Salance of Plant control room panels.
Responsible for approximately 45 exempt personnel, 22 non-exempt personnel, and 129 hourly personnel with an expense budget of nearly 4 million dollars and equipment investment budget of approximately 1.2 million dollars.
11/71 5/75 Manager - Quality Assurance Subsection, Manufacturing Section of Atomic Power Equipment Department, General Electric Company, San Jose, Calif o rnia.
Report to the Manager of. Manufacturing.
Same functional and product responsibilities as in Engagement #1, except at a lower organizational report level.
Developed a quality system which received NRC certification in 19 75.
"he system was also success-fully surveyed f or ASME "N" and "NPT" symbol authorization in 1972 and 1975, plus ASME "U" and "S" symbol authorizations in 1975.
Re s p ons ib l e for from 23 to 39 exempt personnel, 7 to 14 non-exempt j
personnel, and 53 to 97 hourly personnel.
3/70 11/71 Manager - Application Engineering Subsection, Nuclear Instrumen-tation Department, General Electric Co m p any, San Jose, California, i
Responsible for the post order technical in terf ace with architect engineers and power plant owners to define and schedule the instru-mentation and control systems for the Nuclear S team Supply and Balance of Plant portion of nuclear power generating stations.
l Responsibilities included preparation of the plant instrument list with approximate location, review of interface drawings to define l
functional design requirements, and release of functional require-ments for detailed equipment designs.
Personnel supervised included 17 engineers and 5 non-exempt personnel.
A-2
-,,.,-,---r.-
w.,
<.*-3-r,
- ,-,-,y-e+--v---~.,
-,.,-m-,-.--,
9r.,-
..-..---y
,_,.--w-,
-e,e,
-m,#-
,--w--,--vve=
12/69 - 3/70 Chairman - Equipment Ro o m Task Force, Nuclear Ins trumen tation Department, General Electric Company, San Jose, California.
Responsible for a special task fors e reporting to the Department General Manager to define methods to improve the quality and reduce the installation time and cos t of nuclear power plant control rooms.
Study resulted in the conception of a factory-f abricated control room consis ting of signal conditioning and operator control panels scunted on modular floor sections which are completely assembled in the factory and thoroughly tested for proper operation of interacting devices.
Personnel s up e rvis ed included 10 exempt personnel.
12/65 - 12/69 Manager - Proposal Engineering Subsection, Nuclear Instrumentation Department, General Electric Company, San Jose, California.
Responsible for the application of instrumentation systems for nuclear power reactors during the proposal and pre-order period.
Responsible for technical review of bid specifications, preparation of technical bid clarifications and exceptions, definition of material list fo r cost estimating, and the "as sold" review of contracts prior to turnover'to Ap plica tio n En g in ee rin g.
Personnel supervis ed varied frc; 2 to 9 engineers.
8/64 12/65 Sales Engineer, Nuclear Elec tronies Bus iness Section of Ato=ic Power Equipment Deoartment, General Elec tric Company, San Jose, California.
l Responsible for the bid review, contract negotiation, and sale of i
ins trumen ta tio n sys tems and components for nuclear power plants, test reactors, and radiation hot cells.
Also r e s po n s ib le for industrial sales of radiation sensing systems for measurement of chemical properties, level, and density.
10/61 - 8/64 Application Engineer, Low Voltag e Swit chgear Department, General Electric Company, Philadelphia, Pennsylvania.
Responsible for the application and design of advanced diode and silicon-controlled rectifier constant voltage DC power systems and variable voltage DC powet systems for indus trial applica tions.
Designed, followed manufacturing and personally tested an advanced SCR power' supply fo r product in t ro du c tion at the Iron and S t ee l S h ow'.
P roj ec t Engineer for a FC power system for an aluminum pot line sold l
to Anaconda beginning at the 161KV switchyard and encompassing all the equipment to convert the power to 700 volts DC at 160,000 amp'res.
A-3 1
9/60 - 10/61 GE Ro ta tional Training Program Four 3-month assignments on the GE Rotational Training Program for college technical graduates as follows:
a.
Ins t alla~tio n and Service Eng.
Detroit, Michigan.
Installation and startup testing of the world's largest automated hot strip steel mill, b.
Tes ter - Indus try Control - Roanoke, Virginia.
Factory testing of control panels for control of steel, paper, pulp, and utility mills and power plants.
c.
Enginee r.
.Ligh t Military E1ectronics - Johnson City, New York.
Design of ground support equipment for testing the auto pilo ts on the F-105.
d.
Sales Engineer - Morrison, Illinois.
Sale of appliance controls including range timers and refrigerator cold controls.
EDUCATION:
B achelor of S cience Electrical Engineering, University of Arizona, 1960.
Mas ter o f Businass Adminis tration, University of Santa Clara, 1969.
1 P ROFESSIONAL AFFILI ATION :
1 Regis te red Quality Enginee r, License No. QU805, S tate o f California.
Member of Subcommittee 8 of the Nuclear Power Engineering Committee of the IEEE Powec Engineering Society re s p o ns ible for the prepara-i tion and revision of the following 3 national Q.A. Standards:
a.
IEEE 498 (ANSI N4 5. 2.16) :
Requirements for the Calibration and Control of Measuring and Test Equipment used in the Construction and Maintenance of l
Suclear Power Generating S tations.
L l
A-4
l PROFESSIONAL AF FI LI #.TIO N :
( Con td) b.
Installation, Inspection, and Testing Requirements for Class lE Instrumentation and Electric Equipment at Nuclear Power Generating Stations.
c.
IEEE 467 -
- Quality Assurance Program Requirements for the Design and Manufacture of Class IE Instr.umentation and lectric Equipment for Nuclear Power Generating S tations.
I am currently a member of the IEEE Ad Hoc Committee which recommended the issues to be addressed in the development of a standard relating to the selection and utilization of replace-ment parts for Class IE equipment during the cons truction and operation phase.
I am also a member of the work group which 11 prepare this 'p roposed s tandard.
Birth Date:
7/08/37 Marriod; three children Health:
E xc.e ll e n t PUBLICATIONS AND TESTIMONY:
1.
In-Core System Provides Continuous Flux Map of Reactor Cores, R.B. Hubbard and C.E.
Foreman, Power, November, 1967.
2.
Quality Assurance: Providing It, Proving It, R.B. Hubbard, Power, May, 1972.
3.
Testimony of R.B. Hubbard, D.G. B ridenbaugh, and G.C.
Minor before the Unit ed S tates Congress, Joint Committee on Atomic Energy, Feb rua ry 18, 19 76, Washing to n, DC.
(Published by the Union o f Concerned S cientis ts, Cambridge, Massachusetts.)
Excerpts from testimony published in Quote Without Comment, Chemtech, May, 1976.
4 Testimony of R.B. Hubbard, D.G. B rid enb augh, and G.C.
Minor to the Califo rnia S tate Assembly Committee on Resources, Land Use, and Energy, Sacramento, California, March 8, 1976.
5.
Testimony of R.
B.
Hubbard and G.C.
Minor before California State Senate Committee on Public Utilities, Transit, and Energy, Sacramento, California, March 23, 1976.
6.
Tes ti=o ny o r R.B. Hubbard and G.C.
Minor, Judicial Hearings Regarding Grafenrheinfeld Nuclear Plant, March 16 & 17, 1977, Wurzburg, Germany.
A-5
A P_UB LI C ATIO NS AND TESTIMONY:
( Con td) 7.
Testimony of R.3. Hubbard to United States House of Representatives, Subcommittee on Energy and the Environ-ment, June 30, 1977, Washington, DC, entitled, Effectiveness of NRC Regulations - Modifications to Diablo Canyon Nuclear Units.
8.
Tes timony o f R.B. Hubbard to the Advisory Committee on Reactor Safeguards, August 12, 1977, Washington, DC, entitled, Risk Uncertainty Due to Deficiencies in Diablo Canyon Quality Assurance Program and Failure to Implement Current NRC Practices.
9.
The Risks of Nuclear Power Reactors:
A Review of the NRC "Jactor Safety Study WASH-1400, Kendall, et al, edited by R.B.
Hubbard and G.C.
Minor for the Union of Concerned S cientis ts,
August, 1977.
10.
Swedish React'or Safety Study:
Barseb'ck Risk Assessment, MH3 a
Technical Associates, January 1978 (Published by Swedish Depart-ment of Industry as Document DSI 1978:1).
11.
T es timony o f R.B. Hubbard before the Energy Facility S iting Co u nc il, March 31, 1978, in the matter of Pebble Springs Nuclear Power Plant, Risk Assessment:
Pebble Springs Nuclear Plant, Portland, Oregon.
12.
Presentatico by R.S. Hubbard before the Federal Minis try f or Research and Technology (3MFT), August 31 and September 1,
- 1978, Meeting on Reactor Safety Research, Risk Analysis, Sonn, Germany.
13.
Testimony by R.3. Hubbard, D.G. Bridenbaugh, and G.C.
Minor before the Atomic Saf ety and Licensing Board, Septe=ber 25, 1978, in the matter of the Black Fox Nuclear Power S tation Cons truc tion Per=it hearings, Tulsa, Oklahoma.
14 Tes timony of R.B. Hubbard before the Atomic Safety and Licensing Board, November 17, 1978, in the matter of Diablo Canyon Nuclear Power Flant Operating Licens e H earings, Operating Basis Earth-quake and S eismic
.nalysis of Structures, Systems, and Com-ponents, Avila Beach, California.
15.
Testimony of R.3.
Hubbard and D.G.
B ridenbaugh b ef ore the Louisiana Public Service Commission, November 19, 1978, Nuclear Plant and Power Generation Cos ts, Baton Rouge, L o u is ia na.
16.
Testimony of R.B. Hubbard before the California Legislature, Subcommittee on Energy, Los Angeles, April 12, 1979.
A-6
PUB LI C ATIO NS AND TESTIMONY:
( Con td) 17.
Testimony of R.B. Hubbard and G.C. Mino r b efo re the Federal Trade Commission, on behalf of the Union of Concerned S cien tis ts, Standards and Certification Proposed Rule 16 CFR Part 457, May 18, 1979.
18.
ALO-62, Imoroving the Safety o f LWR Power Plants, MHB Technical
' Associates, prepared for U.S.
Department of Energy, Sandia National L aboratories, September, 1979, available from NTIS.
19.
Testimony by R.B. Hubbard b ef ore the Arizona S tate Legislature, Special Interim House Committee on Atomic Energy, Overview of Nuclear Safety, Phoenix, AZ, September 20, 1979.
20.
"The Role of
,t h e Technical Consultant," Practising Law Ins ti-tuter program pn " Nuclear Litigation," New York City and Chicago, November, 1979.
Available from PLI, New York City.
21.
Uncertainty in Nuclear Risk Assessment Me tho do lo gy, MHB Technical Associates, January, 1980, prepared for and available from the Swr. dish Nuclear Power Inspectorate, Stockholm, Sweden.
22.
Italian Reactor Safety Study:
Caorso Risk Assessment, MHB Technical Associates, March, 1980, prepared for and available from Friends of the Earth, Rome, Italy.
23.
Development of Study Plans:
Safety Assessment of Monticello and Prairie Island Nuclear S tations, MHB Technical Associates, August, 1980, prepared for and available from the Minnesota Poll u t io n Control Agency.
24.
Affidavit of Richard B. Hubbard and Gregory C.
Minor before the Illinois Commerce Commission, In the Matter o f an Inves ti-gation of the Plant Construction Prcgran of the Commonwealth E dis on Company, prepared for the League of Woman Voters of Rockford, Illinois, November 12, 1980, ICC Case No. 78-0646.
25.
Systems Interaction and Single Failure Criterion, MHB Tech-nical Associates, January 1981 prepared for and 11 from the Swedish Nuclear hower inspectorate, S tock$olm,able Sweden.
b A-7
s TABLE 15.5-12 CALCULATED ACTIVITY RELEASES FEDM LOCA-DESIGN BASIS CASE (CURIES)
NUCLIDE 0-2 lirs 2-8 lirs 8-24 lirs 24-96 lirs 4-30 Days 1-131 0.27031 0? G.O C.' O 0.0 0.0 l-132 c.;GtSt 02 G.L O.0 0.0 0.0 l-133 0.t?07L 0? 0.0 0.0 0.0 0.0 n
1-134 0.76c.3L 02 C.O O.0 0.0 0.0 1-135 0.5712G 01 0.0 0.0 0.0 0.0 p
1-1310 F< G 0.7340!. 02 0.21100 03 0.5561e 03 0.107 0E C4 - 0. 32 27E 04 p,
1-1320P.C L.F225E O2 0. 67 6? E O2 0.186?E 07 0.'? 40E-01 0. E f 57E-lO m
r:s I
l-133GR6 0.le391 O? O.4314.E 03 0.007EE 03 0.5263F 03 0.5383E O2 s
1-1340RC 0.90471 02 0.2409E 02 0.?O45E 00 0. ? P 11 f -Ot. 0. 76 f SE-31 y
1-13bORG C.1411f 03 0.2e3er 03 0.266eE 03 0.314ni 02 0.lii34e-Oi g
1-131 PAR 0.c17ff 02 0.2713F O '4 0.6451E 03 0.133BF 04 0.4033E 04 a
1-132 PAR O.lO41I 03 0.10650 03 0.2327E 02 0. 1 1 S f. E 00 0.1070E-09
'd 1-133PAF O.20481 63 0.5392E 03 0.1010E 04 0.eS74F 03 0.t-728E O2 l-134 PAR O.1231L 03 0.3066 L 02 0.?SS7E 00 0.3514E-OL O.3331E-31 y
1-13SFAR 0.17 t,4 [. 03 0.3 54 E 03 0.333SE 03 0.3935F 02 0.??93f-Ol KR-P3M O.921,00 63 0.74 E7 E 63 0.8940E O2 0.1154E 00 0.2f78E-12 8
AR-e5 0.t2791 02 0.19130 03 0.5097E 03 0. ll45 F 04 0.9P27F 04 KR-ESM 0.2 h230 04 0. 4 6 f.O E 04 0.2723E 04 0.11*>1E 03 0.1413E-02 KR-87
- 0. 3 F 4 71 04 0.10tAf 04 0.7273F 02
- 0. 5 75. 2 F -0 7 0. 4 5 30E-14 KR-Bf 0.7CADE 04
- 0. 84 84 t.
04 0.238BE 04 0.2720L 02 0.3333E-06 XF-133
- 0. l t.04F 0 5 0.494? F 05 0.1241E 06 0.220fE 06 0.439?E 06 XE -133 M U.4.?tOF O': 0.1212E 04,0.2819E 04 0.376tE 04 0.z$5LE 04 XE-135
(.74021 04 C.165*0 05 0.202PE 05 0.4316E 04 0.1910E 02 AE-13SM O.1S0tt 0 3 0.4137L 01 0.4690E-06 0.7061E-25 0.0 XE-130
- 0. 2 :r 641 04 0.c.551t 01 0.1164E-06 0.125?f-27 0.0
TABLE 15.5-14 (REM)
TilYROID DOSE
'I'dO I!OllR - CONTAINMENT LEAKAGE - DESIGN BASIS CASE 1
1 g
O C151 ANCF FROM P.E L E A SE POINl Hn l
MIC L I D. E
..P ( C M.
I?OUM
?OOOM 4000M 7000M 10000M 20060M 1
1-131 0.73421 bl o.L't15f 41 0.?SSSE 01 0.10800 01 0.44P3r 00 0.3053F 00 0.1??PE 00 ts 1-L3?
6.'3913L 00 O'.251tl 00 0.13 F3 E 00 0.1-71 ! L-Ol O'. 2 6 5 t F-01 0.16?7F-01 0.t547f-02 5' b
1-133 0.455bF,01,0.20?'t 01 0.1611 t- 01 0.6703L 00 0.3093F 00 0.18'5E OG O.7625E-01 Fj l
h 1-134 0.3e41t 60 0.2003f 00 0.11460 00 G.4767E-01 0.12 00E-01 0.134F,F-01
- 0. 5 4 2 7.f- 02 O t;
l C
1-135 0.13OOL (1 (. ! ~- S t i 00 0. 4 56t-F CO O.141?E 00 0. ( 6 23E-C 1 0.f 407F-01 0.?l7'F-Ol n 1 - 1.5 1 0 h G O. I"4 L O? O.1*P1L 02 0.70494-61 0.2933E Ol 0.1353L 01 0.F?o?E CD 0.?33t1 00 $
n I--1320kb O. 817t-f OO O.t?Lbf 00 0.20500 00 0.1202E 00 0.bt4PE-01 0. 34 00E -01 0.13610-01 l 1-133ORG O.1203f 62 e.7734f 01 0. ' ? '. 3 i. 01 0.1770E 01 0.P166L 00 0.5004E 00 0.?cl3E 00 m 1-124 DEL O. 4 L l e 60 0.2'C4L 00 O. I t476 00 0.6646E-01 0.3067E-01 O.1 E.7 5-E -O l 0.7 leo [-07 m 1-1350hc o.3?llf 01 0. 2 ( t h i- 01 0.113cf 01 0.4723E 00 0.2179E OG O.133th 00 0.53720-01 $
1-131PAh O.24 R f (2 (. l t ( 21 02 0.PP1]E 01 0.3666E Cl O.16920 01 C.1037E 01 0.4170F 00 8 1-13?Ftk O.lO??f 01 0.tStti 00 O.3613f 00 0.1503E 00 O.L93CL-01 0.4250E-01 0.1710E-01 1-133PAf< 6.1504. 02 0. 'c I.7 L 01
- 0. t ~- 17 L 01 0.2212E 01 0.1021F 01 0.62 ME 00 0.2b10f 00 I-134 PAL O.56aFE 00 6.?t.30I 00 0.1*+70 00 O.b307E-01 0.3633f-61 0.734cE-01 0.c450E-02 1-13tPAh 0.4014E 01 b.e ' POE 01 0.1419F G1 0.t.904E 00 0.27240 00 0.1660E 00 0.67156-01 01 0.16050 01
_ TOTAL 0.'/$c3F O2 C.I l g,6] 0 2_ o. 31011,( 2 0._1411 E _02, 0. c-511 E_O l
- 0.3 64 0E
APPENDIX D
- (Source :
Diablo Canyon Emergency Plan) 4.1.3 Site Emergency 4.1.3.1 Description The Site Emergency action level. reflects conriitions where there is a clear potential for significant releases, such releases are likely, or they are occurring, but in all 4
cases where a core meltdown situation is not indicated based on current information.
Because the possible release associated with a Site Emergency is significant, care must be taken in alerting offsite authorities to distinguish whether the release is merely potential, likely, or actually occurring.
Response of offsite authorities will be guided initially by this-determination.
4.1.3.2 Release Potential and Significance The Site Emergency class includes releases up to 1000 Ci of I-131 equivalent and/or up to 108 Ci of Xe-133 equivalent.
Assuming design basis meteorological conditions, the maximum Site Emergency release would produce the following i
doses due to direct exposure to.the plume centerline:
00WNVIND WHOLE BODY DOSE THYR 0ID DOSE DISTANCE ASSUMED FROM Xe-133 FROM I-131 (m)
(y[Q)(sec/m)
(mrem)
(rem) 3 800 (site 5.3 x 10-4 6000 270 boundary) 10000 2.2 x 10-5 250 12
. (edge of l
LPZ) l 16000 1.2 x 10-s 140 7
(10 mile zone)
As can be seen, such a release occurring with unfavorable meteorological conditions would certainly require that protective measures be taken on the site and in the downwind sectors throughout the plume exposure Emergency l
Planning Zone.
However, even in the case of a maximum release, it is likely that offsite doses would be much i
lower than those tabulated above due to such factors as l
more favorable meteorology and the ef fects of sheltering.
j The appropriate near term response for such an occurrence l
is to make an assessment of conditions ~ as they actually exist and take action based on this assessment, as it discussed below.
(
4-5
YddBNOIX H
)soni.oc :
MYSH-lr00' gpu uodop(
1VB1'3 S-E dMH CONINYNI YDOIG3NI. S3Dn3ND3S As
- H3I'3Y53 DY13OCIn3S
.mvu avuoc.us 3...
o n 13 1
t t
5 9
4 8
6-
,g
'34-9,gg 70 9 wek,
40-2 v-9 V
yf**
vt-A VU-D p
p
,g p
taf0 tetO tet0,
f810
- te 8810,
tsto ts10 8810 VJ-D 79-t vwh,
gro.f,
vuJb wwb,
1'v893 1TY 1510
- 10 1810,
ts10,
1810 tut 0,
- 10
,p
-10 t 10 28t0 v0J.3 vJ.9 T
~ yraJ.A.11 ts10,
tat 0*tS 11 VD*D yp.9 11 4810,
6810
-6
,g
,g
,g g
v dase'etttat**
tstO ts10,
tut 0 1810,
tut 0,
tut 0 ts10,
ts10 1 10
$ 0-3,,
$.9 3
I -
9 3C-9 S R-9
$ CJb 8 C-D
$ 0-1 -H 1
9
,g yts10,,
9 9-D t
'I' 10 t *I'* ti t
6 t
'I' 9 tI' I
gtst0*tt 8810 9810 I
tstO 9 ub,,
S C-9 8 eh 9 ub,,
3 6-t @
5 30> g t
y b
1
-6 3
g y,
ts10 1810 tet0 tst0 8810 get0 g
MJ*A
$ J*f f MJb S,tst0, ts10-0 t
B Jk tDst0=fO g
9 tat 0 3 0=t,,
5 0-9 t
t
=10 ts10 p
,g
,g 9
,I dangegtt ttte s tat 0,
ts10 2810 tsiC este,
9*t0 9810,
tut 0, ts10 8 6-A,,
$ cb
$gCD-5 5 0-f,,
8 6b 8 eb,,
S,S-D *tO t
p g9510,
g ts10 tet0 taf0 ts10 ts10 gat 0
$ Jk 3 uJ-A*
I kh S N.g g.3 g m.3 g[ts10*8 I9s10
- I 1810'
!stO 9810 1810 swvir rv
,23gp
,tg,9
,tj,9
,r,t ut O jp
-6
-10 aM
=t s
ts10 tstO 1810 I
-9,
3[ts10,
S 0-0 g
g 6810 w
- g g.9 g/gatO' I6s10 '
,g
,g
,g
,g t e Jee'ei t t tf *
- tst0 tuf0 tat 0,
teto tat 0 8510,
ts10 I
u3 =p(
e31 v*D s *3 tatC,gg ts10,gg tatC,g fs10,g 5712401 ASff34 A=t tut 0*tt vt'J4nuS
- d-e 3,,
12 18t0
,gg g
p g
-4
,g e sa.eee1111Tes tat 0 1810 ts10 t a t0 1810 tst0,
1810 IeJNW#431N$
51811WS 17131 9
,,99
)3M23E 447A3(
- A p
p a gaogeetitttes pste tsfO
- tst0, tsf0 red"s10,
Mf..I,
Mb,,
- to ANTW. *3, g MH M -9 t. *O, t
tstO ts10 tst0 ts10 9510,
4svws12xA Ms..p ne-e,,
aab-t
,g, Aseb,
sAsnA. A tute,,
tst0 8810 ts10,
ratse-e '
anret.>
- tsfO*
1s10*
,g p
p
,g
,g A d288'e1113188 ts10 t810 ts10
&810 tste ts10,
1810 lI( t m atoN CJ v*". T73!33MA 520 nan 31$ d3v e717vif 374130#A m2civre p
p p
p p
1$04 hvTnaf 6810 0510 tst0 5810 tat 0, 9810 9810 ts10 9810
,g 1?ws ecneG p
,g p
,g ts10,
ts10
)lt AmfI(
- 6*t0,
se to 9s10 esio ts10 t'10,
1810 64HW 50o31 p
,g p
,g
,g 16M AWnI(
6810 8810 D8'0 5810,
- 10,
ret 0 f810,
ts10,
tet0
&qe isagegittttes )02 eesu setsege stteleJA jes tep saset 1Jee tuf ive 3 )ti ett 9331psul sebaeussa tte ce u,epteW wo tr e
.. im. m mivni.uteni e.tsa>*s a-e se an ante owt ita sta. tes anerfa sne m..eew
- .t......ie6. 4 e 1ti 4 t>dA 10 1YEI'3 S-E ON JOMfCMIND dYD3 H-I
KEY TO PWR ACCIDENT SEQUENCE SYMBOLS A - Intermediate to large LOCA.
3 - Failure of electric power to ESFs.
B* - Failure to recover either onsite or offsite electric power within about 1 to 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> following an initiating transient which is a loss of offsite AC pcwer.
C - Failure of the containment spray injection system.
Failure of the emergency core ccoling injection system.
D Failure of the containment spray recirculation system.
F Failure of the containment heat removal system.
G Failure of the emergency core cooling recirculation system.
H K - Failure of ths reactor protection system.
Failure of the secondary system steam relief valves and the auxiliary feedwater system.
L M ~ Failure of the secondary system steam relief valves and the power conversion system.
Q - Failure of the primary system safety relief valves to reclose after Opening.
R - Massive rtpture of the reactor vessel.
S - A small LOCA with an equiv'alent diameter of about 2 to 6 inches.
5 - A small LOCA with an equivalent diameter of about 1/2 to 2 inches.
2 T - Transient event.
V - LPIS check valve failure.
Coht ainment rupture due to a reactor vessel steam explosion.
o S - Containment failure resulting from inadequate isolation of containment openings and per,atrations.
Y - Containment failure due to hydrogen burning.
6 - Containment f3ilure due to overpressure.
C - Containment vessel melt-throug:
KEY TO TABLE 5-2 s -
%-2
- ?
APPENDIX F (Source:
WASH-1400, Appendix VI)
Sution 2 Releases from Containment 2.1 GENEPAL REMARKS A large pertion of the work of the Reactor Safety Study was expended in determining the probability and magnitude of various radioactive releases. This work is described in datail in the preceding appendices as well as Appendices VII, and VIII.
In order to define the various releases that might occur, a series of release categories were identified for the postulated types of containment failure in both BWRs and PWRs.
The prcbability of each release category and the associated magnitude of radioactive releases (as fractions of the initial core radioactivity that might leak frem the containment structure) are used as. input data to the consequence model.
In addition to probability.and release magnitude, the parameters that characterize the various hypothetical hccident sequences are time of release, duration of release, warning time for evacuation, height of release, and energy content of the released plume.
The time of release refers to the time interval between the start of the hypothetical accident and the release of radioactive material from the containment building to the atmosphera; it is used to calculate the initial decay of radioactivity. The duration of release is the total time durine which cadioactive material is emitted into the atmosphere; it is used to account for continuous releases by adjusting for horizontal dispersion due to wind meander.
These parameters, time and duration of release, represent the temporal behavior of the release in the dispersion model.
They are used to model a " puff" release from the calculations of release versus time presented in Appendix V.
The warning time for evacuation (see section 11.1.1) is the interval betwsen awareness of impending core ' melt and the release of radioactive material from the containment building. Finally, the height of release and the energy content of the released plume gas af fect the manner in which the plu' e would be dispersed in m
the atmosphere.
Table VI 2-1 lists the* leakage parameters that characterize the PWR and BWR release categories.
It should be understood that these categories are composites of numerous event tree sequences with similar characteristics, as discussed in Appendix V.
2.2 ACCIDENT DESCRIPTIONS l
To help the reader understand the postulated containment releases, this section i
presents brief descriptions of the various physical processes that define each release category.
For more detailed information on the release categories and the techniques empicyed to compute the radioactive releases to the atmosphere, the reader is referred to Appendices V, VII, and VIII.
The dominant event tree sequences l
in each release category are discussed in detail in section 4.6 of Appendix V.
PWR 1 This release category can be characterized by a core meltdown followed by a steam l
explosion on contact of molten fuel with the residual water in the reactor vessel.
l The containment spray and heat removal systems are also assumed to have failed and, therefore, the containment could be at a pressure above ambient at the time of the steam explosion.
It is assumed that the steam explosion would rupture the upper l
portion of the reactor vessel and breach the containment barrier, with the result I
that a substantial amount of radioactivity might be released from the containment in a puff over a period of about 10 minutes.
Due to the sweeping action of gases generated during containment-vessel melttnrough, tne release of radioactive materials j
would continue at a relatively low rate thereafter.
The total release would contain 2-1
s.
m approximately 70% of the iodines and 40% of the alkali metals present in the core at the time of release.1 Because the containment would contain hot pressurized gases at the time of failure, a relatively high release rate of sensible energy from the containment could be associated with this category. This category also includes certain potential accident sequences that would involve the occurrence of core melting and a steam explosion af ter containment rupture due to overpressure.
In these sequences, the rate.of energy release would be lower, altaough still relatively high.
PWR 2 This category is associated with the f ailure of core-cooling systems and core malting concurrent with the failure of containment spray and heat-removal.. systems.
Failure of the centainment barrier would occur through overpressure, causing a substantial fraction of the containment atmosphere to be released in a puff over a period of about 30 minutes.
Due to the sweeping action of gases generated during containment vessel meltthrough, the release of radioactive material would continue at a relatively low rate thereafter.
The total release would contain approximately 70% of the iodines and 50% of the alkali metals present in the core at the time of release. As in PWR release. category 1, the high temperature and pressure within containment at the time of containment failure would result in a relatively high release rate cf sensible energy from the containment.
PWR 3 This category involves an everpressure failure of the containment due to failure of containment heat, removal.
Centainment failure would occur prior to the commencement of core melting.
Core melting then would cause radioactive materials to be released through a ruptured containment barrier. Approximately 20% of the iodiles and 20% of the alkali metC s present in the core at the time of release would be reieased to the atmosphers.
Most of the re} sase would occur over a period of about 1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />.
The release of radicactive mate ial from containment would be caused by the sweeping Since action of gases generated y the reactica of the moltan fuel with concrete.
thsse gases would be init...lly heated by contact with the melt, the rate of sensible energy release to the atmosphere would be mcderately high.
PWR 4 This category involves f ailure of the core-ccoling system and the containment spray injection system af ter a loss-of-coolant accident, together with a concurrent failure of the containment system to properly isolate.
This would result in the release of 9% of the iodines and 4% of the alkali metals present in the core at the time of release.
Most of the release would occur continuously over a period of the containment recirculation spray and heat-removal systems 2 to 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />.
Because would operate to remove heat from the containment atmosphere during core melting, a relatively icw rate of release of sensible energy would be associated with tnis category.
PWR S This category involves f ailure of the core cooling systems and is similar to PWR release category 4, except that the containment spray injection system would operate to further reduce the quantity of airborne radioactive material and to initially suppress containment temperature and pressure.
The containment barrier would have large leakage rate due to a concurrent failure of the containment system to properly aisolate, and most of the radioactive material would be released continuously over a period of several hours. Approximately 3% of the iodines and 0.9% of the alkali matals present in the core would be' released.
Because of the operation of the containment heat-removal systems, the energy release rate would be low.
The release fractions of all the chemical species are listed in Table V 2-1.
I The release fractions of iodine and alkali metals are indicated here to illustrate the variations in release with release category.
j r
l
.,~
a, PWR 6 This category involves a core meltdown due to f ailure in the core cooling systems.
The containment sprays would not operate, but the containment barrier.would retain its integrity until the molten core proceeded to melt through the concrete containment base mat.
The radioactive materials would be released into the ground, with some leakage to the atmosphere _ occurring upward through the ground.
Direct leakage te the atmosphere would also occur at a low rate prior to containment-vessel.meltthrough.
Most of the release would occur continuously over a period of about 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />.
The release would include approximately 0.08% of the iodines and alkali metals present in the core at the time of release. Because leakage from containment to-the, atmosphere would be low and gases escaping through the ground would be cooled by contact with the soil, the energy release rate would be very low.
PWR 7 This category is similar to PWR release category 6, except that containment sprays would cperate to reduce the containment temperature and pressure as well as the amount of airborne radioactivity. The release would involve 0.002% of the iodines and 0.001% of the alkali; metals present in the core at the time of release.
Most of the release would occur ever a period of 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />.
As in PWR release category 6, the energy release rate would be very low.
PWR S This category approximates a PWR design basis accident (large pipe break),, except that the containment would f ail to isolate properly on demand. The other engineered,
safeguards are assumed to function properly.
The core would not melt.
The release would involve approximately 0.01% of the iodines and 0.05% of the alkali metals.
Most of the release would occur in the 0.5-hour period during which containment pressure would be above ambient. Because containment sprays would cperate and core melting would not occur, the energy release rate would also be low.
PWR 9 This category approximates a' PWR design basis accident (large pipe break), in which only the activity initially contained within the gap between the fuel pellet and claading would be released into the containment.
.he core would not melt.
It is assumed that the minimum required. engineered safeguards would function satisf actorily to remove heat from the core and containment. The release would occur over the 0.5-hour period during*which the containment pressure would be above ambient.
Approximately 0.00001% of the iodines and 0.00C06% of the alkali metals would be released.
As in PWR release category 8, the energy release rate would be very low.
BWR 1 This release category is representative of a core meltdown folicwed by a steam explosion in the reactor vessel. The latter would cause the release of a substantial quantity of radioactive material to the atmosphere. The total release would contain approximately 40% of the iodines and alkali metals present in tha corr at the time of containment failure. Most of the release would occur over a 1/2 hour period.
Because of the energy generated in the steam explosion, this category would be characterized by a relatively high rate of energy release to the atmosphere.
This category also includes certain sequences that involve overpressure failure of the containment prior to the occurrence of core melting and a steam explosion.
In these sequences, the rate of energy release would be somewhat smaller than for those discussed above,, although it would still be relatively high.
W
.?
w.
c.
o swr 2 This release category is representative of a core meltdown resulting from a transient avant in which decay-heat-removal systems'are assured to fail. ' Containment over-prsssure failure would result, and core melting would follow. Most of the release would occur over a period of about 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />.
The containment failure would be such that radioactivity would be released directly to the atmosphere without si,gnificant r3tention of fission products. This category involves a relatively high rate of snsegy release due to the sweeping action of the gases generated by the molten mass.
Approximately 90% of the iodines and 50% of the alkali metals present in the. core would be released to the atmosphere.
SWR 3 This release category represents a core meltdown caused by a transient event accompanied by a failure to scram or failure to remove decay heat.
Containment failure would occur either before core melt or as a result of gases generated during the inter-action of the molten fuel with concrete af ter reactor-vessel meltthrough. Some fission-product retention would occur either in the suppression pool or the reactor building prior to release to s the atmosphere. Most of the release would occur over a period of about 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> and'would involve 10% of the iodines and 10% of the alkali matals.
For those sequences in which the containment would f ail due to overpressure after core melt, the rate of energy release to the atmosphere would be relatively high.
For those sequences in which overpressure failure would occur before core melt, the energy release rate would be somewhat smaller, although still moderately high.
5WR 4 This release category is representative of a core meltdown *with enough containment leakage to the reactor building to prevent containment f ailure by overpressure. The qucntity of radicactivity released to the at=caphere would be significantly reduced by normal ventilation paths in the reactor building and potential mitigation by the secondary containment filter systems.
Condensation in the containment and the action of the standby gas treatment system en the releases would also lead to a icw rate of energy release.
The radioactive material would be released from the reactor building or the stack at an elevated level. Most of the release would occur over a 2-hour period and would involve approximately 0.08% of the iodines and 0.5% of the alkali metals.
l SWR S This category approximates a BWR design basis accident (large pipe break) in which only the activity initially centained within the gap between the fuel pellet and cladding would be released into containment.
The core would not melt, and containment leakage would be small.
It is assumed that the minimum required engi sered safe-guards would function satisfactorily.
The release would be filtered and pass through t
the elevated stack.
It would occur over a perisd of about 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> while the
~8 containment is pressurized above ambient and wo 21d involve approximately 6 x 10
-7 of the iodines and 4 x 10
% of the alkali metals.
Since core melt would not cccur and containment heat-removal systems would operate, the release to the
_mosphere would involve a negligibly small amount of thermal energy.
.y O e D.
g.
.=a
. ' -.'o'.
-3
'o
= =. b.b a.
.R * * ~ ~ ~ ~. e
- 2.*. * ~ a 3
o.
o.
j
=
m..e
- f. $ 8, e IhIhhh
.$ *. 3.5h e o e......
e e e..
.-e
-==
e e 0
.e
'. ' b = b 2 2 2.
~*
1 R
R~~~~~
~
o. e e.....
.e o..
.'s
.-~--
~!.
e....
S. '. '.
I
- 6 4 ci 4.- b'o 'e
'e e
i
.444..-
. e...
m H
i v
w g
y
- .a.
a.
- * - '.'.'o-
...'o'.
Q T
.-o t ?--a
- q 1
o. o o.....
o.. =
y e. -.
a<
7 U
m m.
.i a a m.. O m m
.p
'. '.- '.-.- b
. 'e. '. '.-
M 4
p 7
w
{
j
,.<=.
- H O
l a
~
g
""t
?.
..L a
e e e g
f
=
i e.e e. e.
e.=.=.---
- e. e. e. 4 e
., ]
e ee e a e e
-- - e.
g 3
e~=
e e
2 S,
E
~
9 e
ja B
Z W
21 M
- 2) 2...*.....
.I s
.-a 3
22 B
1 a a a a a 1
s s s s s s i
a
.3
- 2
{
w a:
2
- 4. :.
je s3 g
L.
M 8:-
4 1
- P 3
- 2'
....g gj 13 L
8 4-0
- 3.. 2 gg O
14
..f.; j W
f
% :J g
- L s
- s
- i.
r
- 1..
i o
.9 g.
W 1=5
- a444;a11
- 444 1 3.g
...o l
M a*..
- 1
$ *3 :. *?..
I:
8' i
i w
I-~ J..* 28 l
a f:
.1. 12:.3:
- 3..
[
?.
W 1 -
o
- s...,.
t&j.:gL
- ~
O 1!:
.L ti
- 13.3 1
tS2 J"5:
x t:
y g
ddd da s**
4sidd
$3:J'.II" 1.
g}f a
2
,1E*0.*L I ~I r 2
- j. 2 e.
o gj f:Ab ".
m f:
8
- 5. ~. u a
.'. b 1 2*
3*.ll I.
'. b '.e
'o '. b '. b 6
'e 6
M 3.i.
3 e.
N e
e a e s e e e e o e a e e a I
.e e
a la eeo ea eo e e
- eo o.
- g..., -.
e
.=
2*-
t
~
.i:.lill. i e e
l M
!(===....e
.5-175 12.f a
g,
.a
.I m
2CIII=......LLLLLLL i i i i 1 13*---
23 46 v