ML20040B815
| ML20040B815 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek, Limerick, 05000000 |
| Issue date: | 08/10/1971 |
| From: | Bush S Advisory Committee on Reactor Safeguards |
| To: | Seaborg G US ATOMIC ENERGY COMMISSION (AEC) |
| Shared Package | |
| ML20040B769 | List: |
| References | |
| NUDOCS 8201260438 | |
| Download: ML20040B815 (10) | |
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125 11E' 1
Appendix B-2 ADVISORY COMMITTEE ON REACTOR SAFEGUARDS UNITED STATES ATOMIC ENERGY Commission I
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-August 10, 1971
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-,e at-i Honorable Glenn T. Seaborg
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Chairman
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U. S. Atomic Energy Commission
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Subject:
REPORT ON NEWBOLD ISLAND NUCLEAR GENERATING ' STATION UNITS NOS. 1 AND 2
Dear Dr. Seaborg:
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At its 136th meeting, Aug'ust 5-7, 1971, the Advisory Comittee on " '
Public Service Electric and Gas Company for a permit to construct
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Reactor Safeguards completed its revieu of the application by the j
the dual-unit Newbold Island Nuclear Generating Station. This project was also considered at the 130th,133rd,134th, and 135th meetings of the Committee on February 4-6, May 6-8, June 10-12, and July 8-10, 1971, respectively; and at Subcommittee meetings on f
June 3,1970 at Argonne National Laboratory, and on February 3, L
March 29, Apri'. 26, June 3, July 7, and August 4,1571 in Wachington, j
D. C.
During its review the Committee had the benefit of discussions l
vith representatives and consultants of.the applicant, the General Electric Company, and the AIC Regulatory Staff. The Committee also l
had the benefit of the documents listed below. The Committee reported the results of its pre-application site review to you in a letter dated September 10, 1969.
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,...a-The station will be located in New Jersey on 530-acre Newbold Island which is near the east bank of the Delaware River about 4-1/2 miles south of Trenton, New Jersey (1970 population - 105,000) and 11 miles northeast of Philadelphia, Pennsylvania (1970 population - 2,000,000).
The nearest population center is a grouping of suburbs in Bucks County, Pennsylvania, known collectively as Levittown (1970 population - 72,000),
l vith its nearest boundary 3.4 miles from the site. The applicant has specified a radius of one mile for the low population ::ene, which had 3
in 1969 a transient population associated with industry of approximately I
1200, and a small resident population which is expected to be about 100 by 1985. The minimum exclusien distance is 700 meters, which extends to the west bank of the Delaware River. As pointed out in the Cnemittee's g
report of September 10, 1069, a relatively high population density is i
associated with this atte.
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8201260438 C20114 PDR ADOCK 05000352 O
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126 Honorable Glenn T. Seaborg' August 10, 1971 r
Each unit includes a boiling water reactor to be operated at 3293 MWt.
_l With respect to core design, power level, and other features of the nuclear steam supply system, the Newbold Island, units are essentially duplicates of the Browns Ferry Units 1, 2 and 3, and Peach Bottom Units 2 and 3.
Waste heat from the station will be rejected to the atmosphere by natural draft cooling towers.
In its report of September 10, 1969, the Committee listed several matters F
which it believed warranted special attention in the design of a plant-for the Newbold Island site. In response to these reco=mendations, the applicant has included in the Newbold Island design several fectures, in addition to those normally provided for boiling water reactor units, to reduce still further the potential for release of radioactivity to the environment. The principal additional features are described below:
J Reactor Building. For cach unit, the conventional steel drywell and suppression chamber primary containment, the fuel handling area and h
spent fuel pool, and the principal components of the engineered safety features are contained in an unlined reinforced concrete building of cylindrical shape with a domed roof. This building is designed to Class I seismic standards and to resist the standard tornado, and mis-siles from this or other sources. The building can resist an internal j
pressure of 2 psig, and inleakage at a differential pressure of 1/4-inch of water will be limited to 10 percent of the bullding volume per day.
A filtration, recirculation, and ventilation system (FRVS) is provided to recirculate and filter the reactor building atmosphere and maintain the building at a negative pressure relative to the outside environment.
Main Steam Lines. A low-leakage, slow-ac' ting, stop valve has been added downstream of the two fast-acting valves in each main steam line, and a i
seal cir system has been provided to further reduce leakage of radio--
activity af ter main steam line isolation. The portion of the, main steam lines containing the isolation valves is enclosed in a Seismic Class I tuane" chamber connected to the reactor building so that any out-leakage following the unlikely event of a design basis loss-of-coolant accident will be treated by the reactor building FRVS before release to the atmos-phere. The entire length of the rain steam lin 4s up to and including the turbine stop valve will be desiped to Class I seismic standards. The main steam lines from the third isolation valve to the turbine stop valve vill be designed and fabricated in substantial accordance vith the require-ments for AEC quality assurance Classification Group B.
In addition, l
seicetive inspection of critical areas of this piping vill be performed during. refueling outages.
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Honorable Glenn T. Seaborg August 10, 1971 f
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e Radiosctive Waste Disposal. The radioactive waste disposal systems include several. features beyond those normally provided in boiling water reactor j
plants. The liquid waste system permits the recycling of equipment and l
floor drain vastes and the evaporation of chemical and laundry wastes before discharge to the envir'onment. The gaseous vaste system provides for the recombining of hydrogen and oxygen, condensing the vapor, hold-up for decay of short-lived isotopes, and cryogenic separation of the noble gases. Krypton and xenon may be stored for periods sufficiently long that krypton-85 becomes the only significant remaining radioisotope. Provisions will be made to utilize non-radioactive steam in the turbine gland seals.
and to process containment purge gases when deinerting. The Comnittee believes that these vaste management systems are capabic of ILmiting releases of radioactivity to the environment to levels that are as low as practicable.
Reactor Vessel Integrity. The applicant has described improvements in the design and fabrication of the reactor vessel. These include redesign of l
the large no:=les to reduce stress concentrations; redesign of the bottom head to reduce the number of welds and improve the capability for in-service inspection; and improved procedures and standards for inspection during fabricacion. The applicant has studied the problems related to i
possible degradation of reactor vessel integrity and has concluded that a nozzle failure or a small break would not impair the integrity of the biological shield, the primary containment, or the reactor internals, and would not affect the ability to cool the core.
In addition, the biological l
shield has been redesigned to increase substantially its ability to with-stand internal pressures, jet forces, or missiles.
Emergency Core Cooling System. The emergency core cooling system (ECCS) has been modified in two ways. The high-pressure coolant injection (HPCI)-
system has been changed to inject water directly to the core through the core spray sparger rather than into the downcomer region via the feedwater
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In addition, the applicant has stated that the steam-turbine sparger.
driven HPCI pump will be modified to the extent feasible to increase the volume of water delivered to the core. The low-pressure coolant injection j
(LPCI) system has been changed to inject water inside the core shroud l
through four scearate vessel penetrations, rather than through the recir-culation lines.
The applicant has stated that these changes provide increased reliability of these systems ar4 reductions in the peak clad temperatures attained in the.unlikely event of a loss-of-coolant accident.
The Cormittee believes that the design changes described above.are suitably responsive to the concerns stated in its letter of September 10, 1969 regarding additional matters which should be considered. for a plant at the Newbold Island site.
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l Honorable Glenn T. Seaborg August 10, 1971
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In the event of an unisolable break of an instru=ent line or a process h
line, reactor coolant will be discharged to the reactor building. Since i
the instrument lines will contain a 3/8-inch flow-restricting orifice in-i Gide the primary Containment, failure of as many as eight such lines will g
not lead to pressures inside the reactor building greater than the 2 psig i
at which it relieves to the environment.
However, failure of a process line, if not isolated in a very short time, could lead to pressures in excess of this relief pressure and significant amounts of reactor coolant would be discharged to the environment. Although the off-site doses from such an accident would be well within the 10 CFR Part 100 guidelines, they would be comparable to or greater than the doses calculated for other less probable accidents. The Committee believes, therefore, that the applicant i
should make design provisions for reducing the quantity of reactor coolant discharged to the reactor building in the event of a process line break.
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The applicant has studied design features to make tolerable the conse -
quences of failure to scram during anticipated transients, and has concluded that automatic tripping of the racirculation pumps and injection of boron could provide a suitable backup to the control rod systed for this type of i
event. The Com:nittee believes that this reci culation pump trip represents a substantial inprovement and should be provided fox the Newbold Island j
reactors. However, further evaluation of the suf ficiency of this approach and the specific means of ieplementing the proposed pump trip should be l
made. This matter should be resolved in a manner satisfactory to the AEC Regulatory Staff and the ACRS during construction of the plant.
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The applicant has stated that a system will be provided to centrol the concentration of hydrogen in the primary containment that might follow in the unlikely event of a less-of-coolant accident. The proposed l
system is not capable of copin < with hydrogen generation rates in t
accordance with current AEC cr:.teria unless the pricary containment is inerted. The re fore, the Cor.ittee believes that th contaitenent should be inerted and that the hydrogen control system should be designed to unintain the hydrogen concentration within accq table limits using the assumptions listed in AEC Safety Guide 7, "Contrul of D :nbustible Gas Concentrations in Containment Following a Loss of Coolant Accident."
Other problems related to large water reactors have been identified by the Regulatory Staff and the ACRS and cited in previous ACRS reports.
The Committee believes that resolution of these items should apply equally to the Newbold Island Station.
The Committee believes that the items centioned above can be resolved during construction and that, if due consideration is given to these items, the Newbold Island Nucicar Generating Station Units Nos.1 and 2 can be constructed with reasonable assurance that they can be operated without undue risk to the health and safety of the public.
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I 129 Honorable Glenn T. Seaborg 5-August 10, 1971 Addi,tional comments by Dr. H. O. Monson, Dr. D. Okrent and Dean N. J.
Palladino are attached.
Sincerely yours, c-13 n 1 Signed by 1
Spencer L E'd Spencer H. Bush Chairman References - Newbold Island Nuclear Generating Station Units Nos. I and 2 1.
Public.Scrvice Electric and Gas Co=pany letter dated February 27, 1970; License Application; Preliminary Safety Analysis Report (PSAR),
Volumes 1 through 5 8
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Amendments Nos. 1 through 5 and Nos. 7 through 9 to PSAR l
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ADDITIONAL COMMENTS BY DR. H. O. MONSON, DR. D. OKPINT AND DEAN N. J. PALLADINO i
Although the large, low pressure, high in-leakage secondary reactor building proposed by the applicant for Newbold Island Units 1 and 2
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represents an improvement over reactor buildings currently employed for BWRs at sites with lower surrounding population densitics, we believe that further improvement is appropriate. The relatively
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scall volume of the steci pressure-suppression type primary contain-ment introduces some crowding of equipment and some attendant problems in the simultaneous accomplishment of full protection against viola-tion of primary containment by possibic missiles, jet forces, and pipe whip, and accomplighment of full, access for in-service inspection.
l Some further protection would te provided against extremely low-probability accidents involving a concurrent loss of primary system integrity and a limited violation of primary containment by the use of a large, relatively high-pressure (of the order of 10 psi, as has been proposed for a BWR at another site having a comparable surround-Lag population density), low-leakage, secondary containment building.
Such a high-pressure, secondary containment, coupled with a pressure-l suppression primary containment, provides a combination which can tolerate a fairly substantial violation ei prbnary containment arising l
from the same event which caused a loss of coolant, as well as further protection against unforeseen events. We believe that this improvement in safety capability is warranted for a more densely populated site like Newbold Island, and recommend that the issuance of a construction permit be contingent on the use of a high-pressure, low-leakage second-ary containment.
.For postulated loss-of-coolant accidents involving small break sizes, l
the high-pressure coolant injection system (HPCI) arranged so as to inject into one of the core spray loops is predicted by the applicant to be highly e f fective in limiting peak clad temperatures to moderate icvels. We believe that for a high power, high-power-density reactor at a site as densely populated as Newbold Island, the applicant should give further consideration to the use of an HPCI system on the second core spray loop. The purpose would be to provide redundancy of this I
means of protection in the event that the single HPCI syscem became ineffective because of failure of an HPCI component or because the l_
accident. arose from tupture of the core spray line into whieS ths i
HPCI injecte. The automatic depressurization system which together
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with the low-pressurc emergency cooling systems constitutes an altersate means for coping with small breaks, albeit by introducing a larger opening, would continue tc serve as a backup.
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J HAME:
LOCHSTET, William A.
l Date of Appointment (s)
(Penn State) 1966-Assistant Professor of Physics l!
Education:
i B.S.
1957 University of Rochester M.A.
1960 University of Rochester Ph.D.
19GS University of Pennsylvania i
Experience:
1957-1958 Graduate Research Assistant, Co=puta.r Center, I
University of Rochester 1959-1961 Graduate Teaching Assistant, University of Pennsylvania 1961-1965 Graduate Research Assistunt, University of Pennsylvania Research with Tandem Accelerator 1965-1966 Instructor, Physics Department, The Pennsylvania State University 1966-Assistant Professor of Physics, The Pennsylvania State University i
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1 William A. Lochstet BIBLIOGRAPHY i
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l J
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C 4
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I
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13'.
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a 14.
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15.
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18.
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29.
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September (1981).
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