ML19320D559
| ML19320D559 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 07/16/1980 |
| From: | Plesset M Advisory Committee on Reactor Safeguards |
| To: | Gilinsky V NRC COMMISSION (OCM) |
| References | |
| ACRS-R-0889, ACRS-R-889, NUDOCS 8007210523 | |
| Download: ML19320D559 (2) | |
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A UNITED STATES
[*' S g gr NUCLEAR REGULATORY COMMISSION W
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QS WASHINGTON, D. C. 20555 July 16,1980 The Honorable Victor Gilinsky Commissioner U.S. Nuclear Regulatory Commission Washington, D.C. 20555
Dear Dr. Gilinsky:
Your letter of June 4,1980 requested the Committee's views on the utility of in-core thermocouples as an aid in the detection of off-normal conditions and for use in accident analysis.
Your request was in the context of instru-
- mentation to follow the course of an accident.
As you noted, the ACRS has long been concerned with assuring that qualified instrumentation should be available to follow the course of a serious acci-dent. This concern has extended to the use of data from core outlet thermo-couples in PWRs.
For example, the ACRS has commented on the _ desirability of thermocouple availability and use in its reviews of the Davis-Besse, Oconee, and Indian Point Unit 3 plants.
More recently, the Committee commented on the use of thermocouples during a discussion with the Com-mission immediately following the accident at TMI (see the attached corre-spondence in which appropriate paragraphs are underlined).
The Committee believed then, as it does now, that core outlet thermocouples can provide a readily available check on certain types of core conditions and
-serve as an additional means for detection of core benavior anomalies.
The Committee also believes that, as practical, such infornation should be made readily available to the operator over the full temperature range of the installed detectors.
The Committee recommends that, as practica', faulty thermocouples should be replaced during refueling intervals.
New PWRs should be designed for thermocouple repl ace.aent.
Core subassemM y exit thermocoupl es or other core outlet thermocouples have not generally been used in BWRs; however, the ACRS favors a careful examination of the feasi-bility of their use in BWRs and the pros and cons of such use.
The - ACRS notes that the NRC Staff has ' developed proposed criteria for installation and use of core outlet thermocouples as part of the Lessons Learned item addressing inadequate core cooling.
NRC Staff efforts in this i
m 'nclude the requi rement specified in ' proposed Regulatory Guide 1.97 for ore exit temperature measurements.
The ACRS supports this Regulatory Guide requirement but suggests that the NRC Staff consider PWR and BWR design differences in its implementation.
The Committee believes that instruments displaying thermoccupl e readings should be readily available in pl ant control rooms, consistent with the philosophy underlying ACRS Generic Item 43:
" Instrumentation to Follow the i
800721043
- t
~The Honorable Victor Gilinsky July 16, 1980 Course. of an Accident".-
The ACRS notes that in the past the availability of core temperature readings has varied. from plant to plant, and offers as
- an example the fact that had the accident at Three Mile Island Unit 2 occurred at Unit 1, it would have been necessary to enter the containment to obtain core. temperature information.
Sincerely, Milton S. Plesset Chai nnan Attachments:
1.
Memo dtd April 18, 1979 transmitting, " Recommendations of the Nuclear Regulatory Commission Advisory Committee on Reactor Safeguards Regarding the March 28, 1979 Accident at the Tnree Mile Island Nuclear Station Unit 2" - Presented Orally to, and Discussed with, the NRC Commiisioners during the ACRS - Commissioners Meeting on April 17, 1979 2.
Ltr dtd November 14, 1973 from H. G. Mangelsdorf, Chainnan, Advisory Committee on Reactor Safeguards to D. L. Ray, Chairman, U.S. Atomic Energy Commission,
Subject:
" Interim Report on Indian Point Nuclear Generating Station Unit No. 3" 3.
Ltr dtd September 23, 1970 from J. M. Hendrie, Chairman, Advisory Commit-tee on Reactor Safeguards to G. T. Seaborg, Chairman, U.S. Atomic Energy Commissior.,
Subject:
" Report on Oconee Nuclear Station Unit No.1" 4.
Ltr dtd August 20, 1970 from J. M. Hendrie, Chairman, Advisory Committee on Reactor Safeguards to G. T. Seaborg, Chainaan, U.S. Atomic Energy Com:nission,
Subject:
" Report-on Davis-Besse duclear Power Station" Referances:
1.
U.S. Nuclear Regulatory Commission: Draft 2 of Revision 2 to Regulatory Guide 1.97:
" Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assass Plant and Environs Conditions During and Following an Accident" June 4,1980.
2.
U.S. Nuclear Regulatory Commission:
"TMI-l Restart Evaluation of Licensec's Compliance With the Short and Long Tenn Items-of Section II of NRC Order dtd-August 9,1979 - Metropolitan Edison Company, et al. Three Mile Island Nuclear Station, Unit 1 - Docket No. 50-289".
U.S. NRC Report NUREG-068U, June 1980.
3.
Letter-dtd May 5,1980 fro,n Bill Lowery, Lawrence Livermore Laboratory, to Gary Holahan,-USNRC, NRR, transmitting, "A Summary of Utility Responses to NUREG-U578, Section 2.1.3.b:
' Instrumentation for Detection of Inade-quate Core Cooling'" University of California, Livennore, California,
.. April 28, 1980.
- 4.. Memorandum for V. Gilinsky from H. Denton,
Subject:
Use of In-Core Thennocouples, dtd November 1,1979.
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UNITED STATES c(
g NUCLEAR REGULATORY COMMISSION 3
j ADVISORY COMMITTEE ON REACTOR SAFEGUARDS
.f WASHINGTON. D. C. 20555 April 18, 1979 MEMORANDUM EUR: Chairman Hendrie Comissioner Gilinsky Comissioner Kennedy Comissioner Bradford Comissioner Ahearne FROM:
R. F. Fraley, Executive Director Adviu ry Committee on Reactor Safeguards Attached for your information and use is a copy of the recommenda-tions of the Advisory Committee on Reactor Safeguards which were orally presented to and discussed with you on April 17, 1979 re-garding the recent accident at the Three Mile Island Nuclear Sta-l tion Unit 2.
R. F. Fraley Executive Director
Attachment:
Recommendations of the NRC Advisory Committee on Reactor Safeguards Re. the 3/28/79 Accident at 'Ihe Three Mile Island Nuclear Station Unit 2 J "-
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April 17, 1979 RECOMMENEATIONS OF 'IHE NUCLEAR REGUIATORY COMMISSION ADVISORY COMMITTEE Qi REACICR SAFEGUARDS REGARDING 'IEE MARCH 28, 1979 ACCIDENT AT
'IEE 'IEREE MILE ISIAND NUCLEAR STATION UNIT 2 Presented orally to, and discussed with, the NRC Comissioners during the ACRS-Commissioners "seting on April 17, 1979 - Washington, D. C.
Natural circulation is an important mode of reactor cooling, both as a planned process and as a process that may be used under abnormal circumstances.
We Comittee believes that greater understanding of this mode of cooling is required and that detailed analyses should be developed by licensees or their suppliers.
S e analyses should be supported, as necessary, by experiment.
Procedures should be de-veloped for initiating natural circulation in a safe manner and for providing the operator with assurance that circulation has, in fact, been established. This may require installation of instrumentation to measure or indicate flow at low water velocity.
Re use of natural circulation for decay heat removal following a loss of offsite power sources requires the maintenance of a suitable over-pressure on the reactor coolant systen.
This overpressure may be
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assured by placing the = pressurizer heaters on a qualified onsite power source with a suitable arrangement of heaters and power distri-bution to provide redundant capability.
Presently operating PWR plants shotad be surveyed expeditiously to determine whether such arrangements can be provided to assure this aspect of natural circula-tion capability.
The plant operator should be adequately informed at all times con-cerning the conditions of reactor coolant system operation which might affect the capability to place the system in the natural circu-lation mode of operation or to sustain such a mode.
Of particular importance is that information which might indicate that the reactor coolant - system is approaching the saturation pressure corresponding to the ~ core exit temperature.
mis impending loss of. system over-pressure will signal to the operator a possible loss of natural circulation capability.
Such a warning may be derived from pressur-
.izer pressure instruments and hot leg temperatures in conjunction with conventional steam tables.
A suitable display of this information
.should be provided to the plant operator at all times.
In addition, censideration should be given to the use of the flow exit tempera-tures from the fuel subassemblies, where available, as an additional
- indication of natural circulation.
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The exit temperature of coolant from the core is currently measured by thermocouples in many PWRs to determine core performance.
Se Committee recommends that these temperature measurements, as currently available, be used to guide the operator concerning core status. We range of the information displayed and recorded should include the full capability of the thermocouples.
It is also recommended that other existing instrumentation be examined for its possible use in assisting operating action during a transient.
We ACRS recommends that op? rating power reactors be given priority with regard to the definition and implementation of instrumentation which provides additional information to help diagnose and follow the course of a serious accident.
This should include improved sampling procedures under accident conditions and techniques to help provide improved guidance to offsite authorities, should this be needed. The Committee recommends that a phased implementation approach be em-ployed so that techniques can ta adopted shortly after they are judged to be appropriate.
We ACRS recommends that a high priority be placed on the developnent and implementation of safety research on the behavior of light water reactors during anomalous transients. The NRC may find it appropriate i
to develop a capability to simulate a wide range of postulated tran-sient and accident conditions in order to gain increased insight into measures which can be taken to improve reactor safety.
The ACRS wishes to reiterate its p' evious recommendations that a high priority be given to research to improve reactor safety.
Consideration should be given to the desirability of additional e
equipment status monitoring on various engineered safeguards features and their supporting services' to help assure their availability at all times.
We ACRS is continuing its review of the implications of this accident and hope to provide further advf.ce as it is developed.
ADVISORY COMMITTEE ON REACTOR SAFEGUARDS UNITED STATES ATOMIC ENERGY COMMISSION WASHINGTON. D.C. 20545 ROY l ' #U Honorable Dixy Lee Ray Chairman U..S.
Atomic Energy Commission Washington,-D.
C.
20545 Subj ect:
INTERIM REPORT ON INDIAN POINT NUCLEAR GENERATING STATION UNIT No. 3
Dear Dr. Ray:
At its 163rd meeting, November 8-10, 1973, the Advisory Committee on Reactor Safeguards completed an interim review of the appli-catio,n of Consolidated Edison Company of New York, Inc., for authorization to operate Indian Point Nuclear Generating Station Unir No.
3.
The proj ect has been previously considered at Sub-1 committee meetings on July 11, 1973, October 10, 1973 and November 7, 1973.
A tour of the facility was made by Committee members on November 2, 1973.
In this review, the Committee had the benefit of discussions with representatives and consul-tants of Consolidated Edison, their contractor, and the AEC Regulatory Staff.
The Committee also had the benefit of the documents listed.
The Committee reported on the application for construction of Indian Point Unit No. 3 on January 15, 1969.
Indian Point Unit No. 3 includes a four-loop Westinghouse nuclear steam supply system with a design power rating of 3025 MW(t).
The. design is similar to that of Unit No. 2 which has a power rating of 2760 MW(t).
The three-unit Indian Point Nuclear Gene-rating Section is located approximately 2-1/2 miles southwest of Peekskill, New York, and 24 miles north of the New York City boundary line.
The Conmittee's report of January 15, 1969, cal 4ed attention to various matters including the following:
consideration of thermal sho ck t.o the pressure vessel in the~unlikely event of a loss-of-coolant accident (LOCA); measures to deal with possible hydrogen concentration buildup in the containment following a LOCA; greater independence-in the on-site power system; main-coolant-(
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- Honorable Dixy Lee Ray ED" 14 EU pump. flywheels as a potential source of missiles; protection against potential effects of a fuel-handling accident; and the possible effects of systematic or common mode failures.
Most of these. items are generic, not unique to Indian Point Unit No. 3.
Acceptable measures have been taken on Indian Point Unit No. 3 4 -
with regard to the on-site power system, hydrogen concentration buildup, and postulated fuel-handling accidents.
Studies are still underway on the potential for missile generation from gross reactor coolant pump overspeed in the event of certain postulated LOCAs; this matter should be resolved in a manner satisfactory to the Regulatory Staff.
It is believed that resolution of the "hermal shock matter can await the development of further information from the Heavy Section Steel Technology Program and other studies.
With regard to anticipated tran-cients without scram, the Committee recommends that the recently announced Regulatory Staff position be implemented for Indian Point Unit No. 3 in timely fashion.
Because there is high power densitylimited operating experience with very large, reactors, the ACRS believes that initial operation should be limited to power levels no greater than 2760 MN(t) and that further review by the Committee is appro-priate before higher pouer levels are permitted.
The Committee believes that, in the consideration of the operation of Unit No. 3 at higher power levels, several factors are pertinent, including the following:
satisfactory experience in Unit No. 3 and other similar reactors; adequate knowledge of fuel perfor-nance; exte t to which an independent confirmation of LOCA-ECCS analysis has been made by the Regulatory Staff; further resolution of. relevant generic matters; and consideration of the possibility of improvements in ECCS effectiveness.
The Committee recognizes that re-evaluation of operating limits may be necessary as result,of possible changes in the accep-a tance criteria for emergency core cooling systems.
The Committee wishes to be kept informed.
1:)
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Honorablo Dixy Lao Rny
', 14 g3 The Applicant stated that he will apply and utilize suitable oquipment to enable periodic testing of the proper positioning of check valves intended to isolate low pressure systems con-nocted to the primary system.
This matter should be resolved is a manner satisfactory to the Regulatory Staff.
Studies are underway with regard to the reliability of the corvice water distribution to the diesel-generators.
This catter should be resolved in a manner satisfactory to the R2gulatory Staff.
The original turbine design has been found by the Applicant to have the possibility of overspeed somewhat beyond the manu-
~facturcr's design condition if the turbine should trip at or near the design power.
The Applicant is preparing design modi-fications to eliminata this condition, and will propose cppropriatc power limitations until acceptable modifications have been made.
This matter should be resolved in a manner ostisfactory to the Regulatory Staff.
The Committee believes that several considerations are appro-priate in the further development of the Technical Specifications,
, na follows:
operating heacup and cooldown pressure-te=perature curves as conservative as practical with respect to 10 CFR Part 50, Appendix G; appropriate baseline inspection and periodic in-service inspection of the steam generator shells; startup of cu idle loop at power; cceeptable cumulative limits on downtime of protection systems and engineered safety features; and con-tinuing availability of care outlet thermocouples.
The Committee also believes that further consideration should be given to augmented use of movable in-core detectors, appropriate in-service inspection of nozzles.in the primary head of the steam generators, and to the detailed specification of administrative controls intended to prevent overpressurization of the reactor vessel below operating temperatures.
Generic problems relating to large water reactors have been identified by the Regulatory Staff and the ACRS and discussed in the Committee's report dated December 18, 1972.
Those prob-loms and additional generic problems identified in more recent ACRS reports should be dealt with appropriately by the Regulatory Staff and the Applicant.
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Honorable Dixy Lee Ray NOV 1 4 1973-The Advisory Committee on Reactor Safeguards believes that, if due regard is given to the items mentioned above, and cubj ec t to satisfactory completion of construction and pre-operational testing, there is reasonable ssurance that indian Point Nuclear Generating Station U it No. 3 can be operated without undue risk to the health and safety of the public.
The Committee believes that operation should be at power levels no greater thaa 2760 MW(t) prior to further Committee review.
Sincerely yours, W
Mangelsdo[
H.
G.
Chairman References Attached 0
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ADVISORY COMMITTEE ON REACTOR SAFEGUARDS UNITED STATES ATOMIC ENERGY COMMISSION WASHINGTON, D.C. 20545 SeptemL e 23, 1970 Honorable Glenn T. Seaborg Chairman U. S. Atomic Energy Com tssion Washington, D. C.
20545 Subjecu REPORT ON 000 SE NUCLEAR STATION UNIT NO. 1 L
Dear Dr. Seaborg:
During its 125th meeting, September 17-19, 1970, the Advisory Comittee on Reactor Safeguards completed its review of the application of the Duke Power Company for a license to operate Unit 1 of the Oconee Nuclear Station at power levels up to 2568 Mi(t). The Committee met with the applicant during its 124th meeting, August 13-15, 1970 and Subcomittee meetings were held on June 23, 1970, at the site and on July 31, 1970' and September 9, 1970, in Washington, D. C.
In the course of the review, the Committee had the benefit of discussions with representatives and consultants of the applicant, the Babcock and Wilcox Cotrpany, the Bechtel Corporation, and the AEC Regulatory Staff, and of stud. of the documents listed.
The Oconee Station is located in a rural area of Oconee County, South Carolina. The nearest population centar is Anderson, 21 miles south, with a population of about 41,000. T1.e minimum exclusion distance for the completed three-unit pcva-statice. will be one mile and the Low Popu-lation Zone radius will be six miles containing about 3,400 people. The water supply for the plant is taken from Lake Keowee which was created by the applicant. The lake and associated recreational facilities are ex-pected to attract a transient population to the area.
The application covers Oconee Units 1, 2, and 3, but this report applies only to Unit 1, which will employ the first of the Babcock and Wilcox two-loop, four-pump, pressurized water reactor, nuclear steam supply systems. The three units are designed to be nearly identical, but some facilities and services are shared in various arrangements. The Committee has reviewed the temporary arrangements necessitated by operation of Unit 1 while Units 2 and 3 are still under construction.
It is believed that the proposed physical measures and administrative procedures to isolate the operating unit from construction activities are adequate.
O dure Mff
'~79 ) D I Sf '747 MutmT 3
-u Honorable Glenn T. Seaborg September 23, 1970 t
The Comittee reported to you on the construction permit application for this power station on July 11, 1967. At that time the proposed operating power was to have been 2452 W(t); the current proposal for operating at powers as high as 2568 W(t) is justified by the applicant, primarily on the basis of a flatter power distribution.
Prior to operation at the higher power level, reactor operation should be reviewed by the Regulatory Staff.
The prestres' sed concrete containment building is similar to those for the Palisades and Point Beach plants which have been reviewed recently for operation.
The Comittee recomends that the applicant accelerate his studies of means l
of preventing comon failure modes from negating scram action and of design features to make tolerabl.e the consequences of failure to scram when required during anticipated transients. As solutions develop and are evaluated by the Regulatory Staff, appropriate action should be proposed and taken by the.
1 applicant an a reasonable time scale. The Comittee wishes to be kept in-formed.
1 The applicant has proposed using a power-to-flow ratio signal as a diverse j
means to cause shutdown of the reactor if emergency core cooling action should be initiated. The Comittee believes it is necessary that either the
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equipment associated with this signal b demonstrated to be able to survive the accident environment for an adequate time or a different, diverse trip signal be employed. This matter should be msolved to the satisfaction of the Regulatory Staff.
The Comittee suggests that developmental techniques, such as neutron noise analysis and use of accelerometers, be considered as an aid in ascertaining displacements, changes in vibration characteristics, and the presence of loose parts in the primary systems.
The Comittee notes the desirability of r.he continuing use of some thermocouples in the core.
The Comittee has commented in previous reports on the development of systems to control the buildup of hydrogen in the containment which might follow in the unlikely event of a loss-of-coolant accident. The applicant I
proposes to make use of a purging technique after a suitable time delay sub-sequent to the accident. Relatively high off-site doses possibly could result folloaing purging of the containment. The Comittee recomends that purging systems be incorporated in the plant but that the primary protection in this regard should utilize a hydrogen control method which keeps the hydrogen concentration within safe limits by means other than purging. The
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Honorable Glenn T. Seaborg September 23, 1970 4
hydrogen control system and provisions for containment atmosphere mixing and sampling should have redundancy and instrumentation suitable for an engineered safety feature; these should be made available within the first two years of power operation. The Comittee wishes to be kept informed of the resolution of this matter.'
The applicant stated that the amount of radioactivity in liquid wastes nor-mally will not be greater than one percent of 10 CFR Part 20 limiting con-centrations after dilution with the minimum flow (30 cfs) below the Keovee dam. Larger flows will have proportionately smaller limiting cuncentrations.
The mean annual discharge from the Keowee dam is expected to be 1,100 cu. ft./
sec. The off-gas system has holding tank and filtering capability and gas re-lease rates are not expected to exceed a few percent of 10 CFR Part 20 limits.
In order to protect against the postulated consequences of the accidental dropping of a fuel element, the applicant has stated that either, he will install filters in the fuel pool building exhaust system, or the equivalent control and protection will be assured by another method. This matter should be resolved to the satisfaction of the Regulatory Staff within the first year of power operation.
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Improved calculational techniques are being applied to the analysis of the efficacy of the emergency core cooling system in the unlikely event of a loss-of-coolant accident.
Interim results appear to be acceptable, but further calculations are needed and some phenomena important to the course of the accident require further study. This matter should be reco1ved in a manner satisfactory to the Regulatory Staff prior to operation at power. The Comit-tee wishes to be kept informed.
The reactor is calculated to have a positive moderator coefficient of reactiv-ity at power which will become negative as boron is removed from the coolant concurrent with build-up of fission products.and fuel burnup. The applicant plans to perform tests to verify that divergent azimuthal xenon oscillations cannot occur in this reactor. The Comittee recomends that the Regulatory Staff follow the measurements and analyses related to these tests.
A conservative method of defining pressure vessel fracture toughness should be_ employed that is satisfactory to the Regulatory Staff.
Other problems relating to large water reactors which have been identified by the Regulatory Staff and the ACRS and cited in previous reports to you should be dealt with appropriately by the Staff and applicant in the Oconee Unit 1 power plant as suitable approaches are developed.
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Honorable Glenn T. Seaborg September 23, 1970 The Advisory Committee on Reactor Safeguards believes that, if due regard is given to the items mentioned above, and subject to satisfac-tory completion of construction and preoperational testing there is reasonable assurance the Oconee Nuclear Plant Unit 1 can be operated at power levels up to 2568 NW(t) without undue risk to the health and safety of the public.
Sincerely yours,
/s/
Joseph M. Hendrie Chairman Additional comments by Dr. W. R. Stratton are presented below:
"The high off-site doses which are stated to accompany the proposed purging operation are based on calculations which include a number of assumptions which I believe to be overly conservative.
It is my opinion that the situation, should
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it ever arise, would be much less severe and that the pro-posed purge system would : ovide adequate protection for the health and safety of the public in this regard and therefore the additional hydrogen control equipment required by this letter is not necessary."
Attachment:
List of References 0077
ADWw. 4Y COMMITTLE'ON ~ REACTOR E JEGUARDS UNITED STATE 3 ATOMIC ENERGY COMMISSION WASHINGTCN D.C. 20545 Augus t 20, 1970 Honorable Glenn T. Seaborg Chairman U. S. Atomic Energy Commission Washfngton, D. C.
20545
Subject:
REPORT ON DAVIS-BESSE NUCLEAR POWER STATION
Dear Dr. Seaborg:
N At its 124th meeting, August 13-15, 1970, the Advisory Com51ttee on Reactor Safeguards completed its review of the application by the Toledo Edison Company and The Cleveland Electric Illuminating Ccmpany for a permit to construct the Davis-Besse Nuclear Power Station. A Subcommittee met to review the project on May 26, 1970, at the site and in Toledo, Ohio, and on August 4,1970, in Washington, D. C.
During its review, the Committee had the benefit of discussions with representatives and consultants of the applicants, the Babcock and.
Wilcox Company, the Bechtel Corporation, and the AEC Regulatory Staff.
The Committee also had the benefit of the documents listed.
The plant will be located on the southwestern shore of Lake Erie ap-p.3ximately' 21 miles east of Toledo, Ohio. The nearest population centers are Toledo and Sandusky, Ohio, each about 20 miles from the site, with populations in 1960 of 379,000 and 32,000, respectively.
The city of Fremont, Ohio, with a 1960 population of about 18,000, is located 17 miles from the site. The minimum exclusion distance is 2400 feet and the low population zone distance is two miles.
Approx-imately 3200 people live within five miles of the site.
Camp Perry, an Ohio National Guard facility, is located on Lake Erie about five miles eas t of the site. This installation is used during a short period each year for target practice with small arms and with 40-mm. anti-aircraft shells armed only with a small destruct charge.
At the Erie Industrial Park, about three to four miles east of the site, Cadillac Gage Company is engaged in testing ordnance equipment firing 120-mm. mortar shells with a maximum range of about two miles. All firing from both locations is directed into restricted areas in Lake Erie.
The applicants have proyided studies which demonstrate that none of the projectiles now being fired from these installations could ATTACHMENT 4 1
Th IIl'h h
E:nerablo.G1cnn T. Sacbarg.
.2-August 20, 1970 penetrate the heavy reinforced concrete structures provided to protect the. essential. portions of the plant. The Committee recommends, how-ever, that the applicants and the Regulatory Staff make suitable arrange-
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masts to be informed of any changes in these activities so tpat their possible effect on the safety of the plant may be evaluated.
An area in Lake Erie about ten miles north of the site is used by air-craft from the Selfridge Air Force Base in Michigan as an Anti-Submarine Warfare practice area and by the Lockbourne Air Force Base at Columbus, Ohio, as an impact area for automatic weapon firing froc aircraft. The applicants have been given assurance by officials of the Department of Defense that military aircraft enroute to or from this area will not be routed closer than ten miles from the site. The Cocsittee believes that this arrangement reduces, to acceptably low levels, the probability of an aircraft striking the plant, but recom= ends that formal arrangements be made to enable the applicants and the Regulatory Staff to maintain continuing awareness of the operational patterns of; military aircraft in this area.
The Davis-Besse plant will include a two-loop pressurized water reactor similar to those for the Midland units except that the internal vent valves have been eliminated by changes in the elevations of the steam generators to obviate their need. Since the proposed arrangement elim
- inates the possibility of coolant flow. bypass through an open vent valve, the Davis-Besse reactor is designed for an initial core power level of 2633 MWt as compared to 2452 MNt for the Midland units.
The applicants stated that it will be possible to anneal the pressure vessel if this shoeld become necessa'ry at some time after operation is begun.
- A suitable preoperational vibration testing program should be employed for the primary system. Also, attention should be given to the devel-opment and utilization of instru=entation for in-service monitoring for excessive vibration or loose parts in the primary system.
The containment consists of a steel vessel surrounded by a reinforced concrete shield buildiag, with the annular space maintained at a slightly negative pressure and the air fro.m this space exhausted thr'ough filters.
This design is sLnilar to that for the Prairie Island, Kewaunee, and Hutchinson Island plants, except that the free volu=e of the steel con-tainment is much greater, nearly three million cubic feet..The Regula-tory Staff should review the containment design pressure to assure that an adequate margin of conservatism exists.
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. jugust 20,;1970
~ginirabia'G13nn T. Sctberg Detailed criteria remain to be formulated by the applicants for the de-sign of the penetrations for the hot process pipes which traverse the In via: of the importance annulus between the two containment barriers.
of these penetrations, criteria 'should be reviewed by the Regula,cory Staff to assure adequate conservatism, and the applicants should arrange for an independent review of the actual design.
The Cdbmittee has commented in previous reports on the development of systems to control the buildup of hydrogen in the containment whichThe might follow in the unlikely event of a loss-of-coolant accident.
applicants are studying various methods of coping with this puoblem, The Committee including purging and the use of catalytic recombiners.
recommends that the primary protection in this regard should utilize a Fydrogen control method which keeps the hydrogen concentration within
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safe limits by means other than purging. The capability for purging The hydrogen control system and provisions should also be provided.
for contain=ent atmosphere mixing and sampling shou 3 have redundancy The Com-and instru=entation suitable for an engineered ss".ety feature.
mittee wishes to be kept informed of the resolution of this catter.
The applicants have stated that they will provide additional evide~nce ob-tained by improved multi-node analytical techniques to assure that the emergency core cooling system is capable of Itaiting core temperatures to acceptably conservative values. They will also make appropriate plant changes if further analysis demonstrates that such c' anges are required.
This matter should be resolved during construction in a manner satisfac-tory to the Regulatory Staff. The Committee wishes to be kept informed.
The Connittee recommends that the applicants accelerate the study of means '
to prevent common failure modes from negating scram action, and of design features to make tolerable the consequences of failure to scram during anticipated transients. The applicants stated that the engineering design
' would maintain flexibility with regard to relief capacity of the primary This matter should systea and to a diverse means of reducing reactivity.
be resolved in a manner satisfactory to the Regulatory Staff during con-struction. The Committee wishes to be kept informed.
The' Committee believes that consideration should be given to the utiliza-tion of instrumentation for prompt detection of gross failure of a fuel Consideration should be given also to the use of core exit element.
thercocouples as an aid to reliable operation and as an additional method of detecting behavior anocalies.
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Honorable Glenn T. Seaborg August 20, 1970*
.The applicants propose batch discharge of liquid wastes following treat-ment. Concentrations of radionuclides in the discharge vill be kept well below 10 CFR 20 limits with positive dilution being provided from several equipment cooling water streams. Plans for operation of waste treatment equipment should be such as to minimize the quantities of radioactivity discharged, ard provisions should be made to achieve rapid dispe sion in the lake.
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.he Davis-Besse plant.
ss The Comnittee believes that the above items can be resolved during con-struction and that, if due consideration is given to these items, the Day.sI esse Nuclear Power Station can be constructed with reasonable B
assurance that it can be operated without undue risk to the health and safety of the public.
Sincerely yours, hd a
3 Joseph M. Hendrie Chairman Reference s:
1.
Letter from Toledo Edison Company, dated August 1,1969; License Application, Volumes 1, 2 and 3 of the Preliminary Safety Analysis Report (PSAR) dolu=e4ofthePSAR,;datedApril 16, 1970 2.
3.
Amendments 1 through 9 to License Application
.