ML19312C414
| ML19312C414 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 06/16/1967 |
| From: | US ATOMIC ENERGY COMMISSION (AEC) |
| To: | |
| References | |
| NUDOCS 7912130936 | |
| Download: ML19312C414 (40) | |
Text
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June 16, 1967 U. S. ATOMIC ENERGY COMMISSION j
DIVISION OF REACTOR LICENSING REPORT TO THE ADVISORY COMMITTEE ON REACTOR SAFEGUARD IN THE HATTER OF DUKE POWER COMPANY CONSTRUCTION PERMIT APPLICATION FOR OCONEE UNITS 1, 2 AND 3 REPORT NO. 2 l
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D Note by the Director, Division of Reactor Licensing
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,ig The attached report has been prepared by the Divisica of Reactor Licensing for consideration by the Advisory Committee on Reactor Safeguards at its July 1967 meeting.
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@u10CDAL U$2 @GGLJ ABSTRACT This is the second report submitted by the staff on the Duke Power Company application for Construction Permits for Units 1, 2, and 3 at its Oconee Nuclear Station, Oconee County, South Carolina.
The subjects reviewed in this report include the thermal and hydraulic design, instrumentation, power containment isolation systems, rod drives, the absence of steam line isolation valves, the turbine missile analysis and accident analyses.
As a result of our review of this applicaition as discussed in this report (Items 1 through 6) and in Report No. 1 dated May
- 12) we have concluded that sufficient supporting information in the follow-24, 1967, ing areas has been presented for the purposes of a construction permit.
(1)
Instrumentation (2) Power (onsite and offsite)
(3) Containment isolation systems (4)
That steam line isolation valves are not required on this plant (5) Turbine missile analysis (6) Accident analysis (7) Site (8) Core Design (9) Reactor coolant system (10) Containment (11) Engineered Safety Features (12)
Sharing of auxiliary components between units We conclude that, although the following areas have not been completely resolved sufficient information has been provided for the purposes of a provisional con-struction permit!
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@u JOC0AL U$3 @GLT (1) Research and development items including (a) steam generator tests (b) incore neutron detection (c) control rod drives, (d) hermal and hydraulic design, and (e) emergency core cooling and blowdown analyses.
(2) Other items on which further design information will be required including (a) moderator temperature c 7 efficient, (b) xenon oscillations (c) irradiation survei.'_ ance progran.
and (d) switching arrangement for the emergency systems.
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@JBC0AL U$2 @MJ TABLE OF CONTENTS Page 1.0 Introduction 1
2.0 Revised Containment Design Pressure 1
3.0 Thermal and Ilydraulic Design 2
3.1 Coolant Distribution 3
3.2 Thermal Design Limits 1
3 4.0 Instrumentation 5
4.1 Safety and Control Instrumentation 5
4.2 Reactivity Control 13 5.0 Power 16 6.0 Containment Isolation Systems 22 7.0 Rod Drives 23 8.0 Steam Line Isolation Valves 25 9.0 Turbine Missile Analysis 28 10.0 Accident Analysis 29 10.1 Rod Ejection Accident 30 10.2 Loss of Coolant Accident 30 10.3 Accidental Liquid Effluent Release 32 11.0 Research and Development 33 12.0 Conclusions 35
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@J003AL U$2 @lk7 1.0 Introduction This is the second report by the stGr on the Duke Power Company application for three pressurized water reactors at its Oconee Nuclear Station in Oconee County, South Carolina.
This report deals with subjects which had not been resolved at the writing of Report No. 1.
In addition, a change in containment volume and design pressure has been proposed by the applicant and a new analysis of the containment design capability has been submitted in I
Amendment No. 4 dated May 25, 1967. This amendment also included answers to f
Staff question list No. 2.
Answers to questfons posed by the ACRS at its June meeting have not yet been received. We intend to prepare a supplemental report dealing with the considerations outlined in the question list prior to the July meeting.
2.0 Revised Containment Design Pressure The volume of the containment has been reduced from 2,050,000 ft to 3
1,900,000 ft and the design pressure increased from 55 psig to 59 psig.
The structural design criteria are unaffected by the volume decrease which was caused by a 10 foot reduction in building height.
The thickness of the dome has been increased from 3 ft to 3.25 ft and the wall thickness increased from 3.5 to 3.75 feet.
A parametric analysis was performed by the applicant to determine the response of the containment during a loss of coolant accident. The assumptions were the same as those described in Report No.1 and the same spectrum of hot leg pipe break sizes was evaluated. The analysis resulted in slightly increased margins between design pressure and peak accident pressures, principally due to increased mass of structural heat sinks within the containment. The increase in l
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- available heat sinks does not reflect design changes but rather a more detailed accounting of the heat sinks.
The table below gives the principal results of the revised analysis and original analysis of containment design capability.
Oririnal Revised Containment design pressure 55 psig 59 psig Highest blowdown pressure (3 ft break) 52.9 psig 56.8 psig l
Highest post-blowdown pressure (14.1 ft2 break) 52.7 psig 55.9 psig The capability to withstand zirconium water reaction remains approximately the same for the new parameters. The capacity of the engineered safeguard systems were unaffected by the change.
Our conclusion in Report No. 1 for this application that the containment capability is acceptable remains unchanged.
3.0 Thermal and Hydraulic Design The thermal and hydraulic design bases for the core and thermal design aspects of the fuel rods were reviewed by us and found generally acceptable.
l The B&W design uses conservative parameters and calculational methods and is similar to previously analyzed pressurized water reactors of comparable power level.
We believe that two aspects of the design must be further investigated as research and development items:
(1) the coolant distribution within the reactor must be established for various modes of operation by experimental and analytical techniques and (2) a heat transfer correlation that accurately predicts the DNB (departure from nucleate boiling) heat flux at various positions within the fuel f
bundle must be established.
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- 3.1 Coolant Distribution The coolant distribution within the reactor vessel internals and the asso-ciated pressure drops must be well established.
This is particularly important since this plant differs from other large reactors in having two outlet legs from the upper plenum and four inlet legs which could result in maldistribution in the inlet plenum and the core.
The applicant has stated that a research and development program (Section 1.5.4) is underway to measure flow distribution in the core, fluid mixing in the vessel and core, and the distribution of pressure drop within the vessel.
These tests will be conducted on a 1/6 scale model of the vessel and internals.
In addition, flow distribution, pressure drop, and mixing data vill be obtained with a full scale fuel bundle test assembly and on various models of reactor flow cells.
We have reviewed the development program as described above and believe that the scale nodel testing and the full scale fuel bundle testing are adequate to provide the necessary information and we therefor 2 believe that the proposed program is acceptable.
3.2 Thermal Design Limits The thermal and hydraulic design evaluation presented in Section 3.2.3 of the PSAR made use of the BAW-168 heat transfer relationship to establish that DNB would not be rt+:hed at the 114% overpower condition.
The reactor trip point is 107% power and the maximum overpower of 114% will not be exceeded under
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any forseeable transient conditions. A probability study was included in the analysis as a means of demonstrating the sensitivity of the analysis to the various e
input parameters and to allow an expression of the fraction of the core endangered f
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when-at various hot channel DNB ratios,
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. The nuclear and engineering hot channel factors used in the analysis were
" design" values - vorse than the expected (nominal) values but were not maximized to give an extreme worst case. We also had some concern that the flow rates in the corner and wall channels would be significantly reduced by the friction drag of the perforations in the fuel assembly can.
In response to our concerns, B&W responded informally that the mass flow rate was lower in the corner and wall channels and that the BAW-168 correlation did not fit these cases (the BAW-168 correlation would give DNB ratios of less than 1.0 if directly applied to the i
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corner and the wall cells). B&W stated that the flow areas in these cells had been increased proportionally in the original design to compensate for the lower mass flow rates.
In the formal submittal (Supplement 2, Question 3.2) the DNB ratios presented were derived from uniform bundle burnout tests now underway at B&W.
These were corrected to a non-uniform axial shape by use of single rod data. The burnout ratios obtained in this manner are significantly higher (less conservative) than those obtained by use of the W-3 correlation.
B&W further informed us that axially non-uniform bundle tests were being conducted and would be completed about December 1967.
They stated their pre-ference for obtaining and analyzing all their data before discussing the results of the tests with us in detail. This would be done initially through a seminar-type presentation.
In view of the above, we asked that corner, wall and unit cells be checked using the W-3 correlation with " worst case" hot channel factors. This has been done and the results indicate that the design values could be justified on the basis of the W-3 correlation.
(Supplemen't 4, Question 13.1).
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- We believe that the allowable design heat flux can be designated as a Research and Development item and that the proposed design values can probably be justified on the basis of the B&W work when completed. We have the assurance, however, that the design could be approved on the basis of the W-3 correlation if necessary.
4.0 Instrumentation 4.1 Safety and Controls Instrumentation i
The reactor protection system automatically scrams the reactor to protect the reactor core under the following conditions:
a)
The reactor power, as measured by neutron flux, reached an established maximum limit or the li. nit set by reactor coolant flow.
b) The startup rate reaches an established maximum limit.
c) Certain mismatch conditions exist between reactor coolant flow and the number of pump motor breakers in service, d) The reactor outlet temperature reaches an established maximum limit.
e) The reactor pressure reaches an established minimum limit.
The reactor protection system automatically scrams the reactor to protect the reactor coolant system when the reactor pressure reacl.es an established maximum limit.
The engineered safety features protection system automatically performs i
the following functions to mitigate the effects of a serious accident:
a)
Initiates operation of the core emergency injection system upon detection of low reactor coolant pressure.
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Initates operation of the reactor building cooling systems upon detection of an abnormally high reactor building pressure, c)
Initiates containment isolation upon detection of an abnormally high reactor building pressure.
d)
Initiates isolation of valves which are directly open to the Reactor Building on a high radiation signal.
A schematic diagram of the reactor protection system is shown in Fig ure 7-2 of the PSAR.
The nuclear instrumentation has eight channels of neutron informati on divided into three ranges of sensitivity:
source range, intermediate range, and power range.
The three ranges combine to give a continuous measurement of rea t c or power from source level to approximately 125% of full power, or ten decad es of infor-mation.
A minimum of one decade of overlapping information is provided The physical location of the neutron detectors is shown in Figure 7 10 (PSAR).
The power range detectors are located in four primary positions
, 90 degrees apart around the reactor core.
Three chambers associated with each power range channel are located near the top of the core, at the midplane
, and near the bottom of the core.
The two source range proportional counters are located on opposite sides of the core adjacent to two of the power range detect ors.
The two intermediate range compensated ion chambers are also locat d e on opposite sides of the core, but are rotated 90 degrees from the source range detectors The applicant has stated that the instrumentation will be designed
, built and tested in accordance with the proposed IEEE Standard for Nuclear Power Plant Protective Systems (Rev. 8).
In addition, the applicant will design in accordance with the following specific criteria outlined in Section 7 of the PSAR
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@i10CDAL U$M @El a) No single component failure shall prevent the protection systems from fulfilling their protective functions when required.
b) No single component failure shall initiate unnecessary protective system action, provided implementation does not conflict with the above criterion.
c) All protection system functions shall be Laplemented by means of redundant sensors, instrument strings, logic devices and action f
devices which combine to form the protection channels, h
d) Redundant protection system channels and their associated elements shall be electrically independent and packaged to provide physical separation.
e) A loss of a.c. power to the reactor protection system shall cause the affected channel'(s) to trip.
f)
Equipment is divided between the redundant engineered safety features channels in such a way that the loss of one of the d.c. power busses does not inhibit the systems' intended safety functions. Loss of a.c. to the engineered safety system shall cause the affected channels to trip.
g) Manual trip shall be independent of the automatic trip instrumentation.
h) Preoperational and on-line testing capability shall be provided.
The basic design of the protection system is shown in Figure 7-2 of the PSAR.
The final reactor trip circuit is shown in Figure 3-59.
All reactor trip instrumentation (with the exception of the "startup rate" i
channels) are coincident and redundant. Four independent channels monitor each
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" trip" parameter, and one (and only one) output of each channel controls four I
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The output of these relays are combined (2/4 logic) to operate four circuit breakers which de-energize the two a.c. input circuits feeding the rod drive (d.c.) power supplies.
The circuit break logic is 1/2 X 2:
1.c., a trip results if (at least) one of the two circuit breakers in.one a.c.
line and (at least) one in the other line are opened.
Each a.c. line furnishes power to one of the clutches through diodes which permit testing of the final trip circuits during reactor operation.
1 The power range instumentation consists of four linear level channels origi-
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nating in twelve uncompensated ion chambers.
The gain of each channel is adjust-able, providing a means for calibrating the output against a reactor heat balance These channels combine power, flow, and pump breaker information and effect reactor trip (2/4 logic) under certain conditions.
There are two " flow tubes", one in each primary loop, as shown in Figure 7-11 of the PSAR.
Flow information is measured as a function of pressure drop by four independent sensors at each tube.
The outputs of the eight sensors are combined as pairs such that four independent total flow signals are derived.
Each total-flow signal is fed to one of the four power range channels, thus creating four independent power / flow channels.
In addition, each pump motor breaker has four contacts which are respectively connected to the four power / flow channels and thus each channel receives identical information.
The powe flow channels will initiate reactor trip if:
a) reactor power exceeds 107% full power (F.P.) under any conditions, or b) the power / flow ratio exceeds 1.07 under any conditions, or I
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- d) a sheared rotor) when operating abov an tripping of its break e X% F.P.,
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- e) two pumps are lost as a result the ratio of reactor power to thof tripped pump motor bre
, and to Lae remaining pumps is great e steady state flow correspon 1
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servo action, calling for a reductio proper power / flow ratio, will occur n in power to achieve a when power is below X% F.P.,
and a) one pump is lost due to tripping of its breaker, or b) more than one pumps is lost due to b reactor power (at the instant of breakreaker tripping and th flow corresponding to the remaini er trip) to the steady state The above provisions ng pumps is less than 1.07.
level commensurate with the remai iallow the downward or power to a flow coastdown unless it is n ng pumps as a means of " keeping ahe d" a
of the comparator circuits) that a foregone conclusion (as " judged" by th e various the impending loss-of-flow tra severe to warrant immediate trip nsient is sufficiently Only one of the four nuclear powe the reactor control system r range channels will provida an input t connected all four power range channel. This is a departur o
which s to the servo.
to Section 4.7 of the IEEE Sta d The design now conforms n ard.
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. Our analysis of the power / flow instrumentation indicates that the design conforms to current criteria. There are eight independent flowmeters, four in each primary loop as shown in Figure 7-11.
The flowmeter outputs are connected as four independent pairs (the flowmeters in each pair monitoring a different loop) such that they become four independent (total) flow channels. A flow channel is combined with only one of the power range channels.
Irdependence is further preserved by feeding each of the four power / flow signals to only one of the four logic circuits.
l Those trip circuits which function as a result of abnormal pump breaker operation are designed to be immune to single electrical failures.
Each breaker has four independent contacts which are respectively connected to one of the four power range channels.
Thus, a failed contact will affect only one channel.
However, a mechanical failure within a breaker (e.g., a breaker which failed t o
open even though power to its pump had been interrupted) would not be cancelled by system redundancy.
Our analysis shows that this failure, or any similar failure involving pump breaker trip circuits, does not cc.atitute a hazard inasmuch as the circuits provide only " anticipatory" trip functions and are always backed up the " power / flow > l.07" circuits which would be effective under any conditi ons of pump motor loss.
The startup rate channels are effective only when reactor power is less than 10% of full power.
Above 10% F.P. an operational bypass, actuated by the power range channels through 2/4 logic circuits, removes the startup rate trip function Our analysis indicates that the two startup rate channels are independent and that no single f ailure, including a failre within the bypass removal circuits can pre-their functioning, vent 1
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. There is one set of four pressure sensors and one set of four temperature sensors which trip the reactor on high and low primary system pressure, and high coolant outlet temperature. The logic is 2/4, and the instrument channels are independently connected to the four logic channels in the same manner as the power range channels. One pressure channel also provides a signal to the pressu-rizer pressure controller.
The other three channels will provide trip action on a redundant basis should a common failure initiate a pressure transient and disable the one channel.
This design conforms to the provisions of Section 4.7 of the i
proposed IEEE Standard (Rev. 8).
Operational bypass circuits within the low pressure portions of the protection system will conform to paragraph 4.12 of the IEEE Standard.
The four logic channels have been analyzed and found to be "f ail safe" in the event of voltage loss, immune to single failures, and testable for credible faults.
The " fail safety" is inherent since the channels are tripped when de-energized. A partially or completely failed channel will disable only one "RS" relay. Action of the three remaining channels will open all four circuit breakers at the clutch power supplies. Action of only two of the remaining channels is actually required, and they will open at least one circuit breaker at each power supply. Testing for faults witnin a logic channel vill be revealed when the by-passed contacts do not trip their RS relay when tested. Open circuits are self-revealing.
Short circuits between channels can be detected by tripping the "high pressure" contacts one at a time (these are the contacts located furthest upstream).
For example, a short between channels one and two will prevent RS1 fr /m dropping out when "Hi Pressure #1" is tripped.
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@JBCOAL U$3 @NJ Our analysis of the final trip circuits (Ref. Fig. 3-59) shows that they are fail safe, immune to single failure, and testable.
The loss of one bre~aker in each a.c. line can be tolerated.
Diode fialure, open or shorted, will not prevent trip action. A " hot" short at the positive d.c. line will have no effect since the d.c. system is ungrounded.
The system will be equipped with ground-fault detectors. Loss of a.c. and/or d.c. will cause, or tend to cause, reactor trip.
Testing at power is accomplished by tripping the circuit breakers one at a time and noting the absence of d.c. voltage at the appropriate power supply output just upstream of its isolating diode.
The manual trip switch contacts are in series with the four circuit breaker undervoltage coils. There is no dependence on instrumentation.
We agree with the applicant that his protection system design criteria are acceptable, and that the specific designs which are being proposed conform to these criteria.
Four sets of pressure sensing channels initiate the engineered safety features.
Each set is cc, incident and redundant (2/3 logic). One set initiates the hip,h and low pressure coolant injection systems. These channels operate throug"i amplifiers and bistable devices and are fail, safe in terms of voltage loss. 1vo other sets of three channels respectively actuate the reactor building spray system.
In these channels pressure switches are operated directly - there is no dependence i on electrical power for switch operation.
The remaining set of pressure sensor channels (2/3 logic) initiates reactor building emergency cooling and containment isolation.
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- Contacts controlled by the aforementioned channels are respectively combined into pairs of redundant logic chains which, in turn, control the safety feature systems. This is shown in Figure 7-2e, PSAR.
These chains are testable at power by means of two lights wired across the contacts of each chain such that the tripping of a chain produces an unique response from its lights.
Each redundant logic chain is energized from an independent d.c. power Should a power source be lost, the downstream circuits fail "as is."
source.
However, we believe that with the system redundancy provided this condition is i
tolerable.
The engineered safety features instrument channels do not control the para-meters which they measure; i.e., there is separation of control and safety.
Manual actuation capability is provided.
We agree with the applicant's criteria and believe that the proposed design of the engineered safety features system properly implements these and all other applicable criteria.
The incore instrumentation system provides no automatic control or protection function.
The system is located entirely within containment, thereby precluding the need for isolation of penetrations associated with the system.
4.2 Reactivity Control The control rod drives are being designed in accordance with detailed criteria stated on pages 3-65, 3-66, and 3-67 of the PSAR which can be summarized as follows:
a) " Single failures" shall be limited to one drive.
b)
No single failure shall cause the uncontrolled withdrawal of any rod.
c)
No more than two control groups can be withdrawn at one time.
d)
The withdraw speed shall be -limited so as to exceed 25 percent over-I-
speed in the event of speed control fault.
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@uJQCBAL U$2 @LJ e) Continuous position indication shall be provided.
We agree with these criteria and have performed a failure mode analysis to t
determine the proposed system's degree of conformity to these criteria:
In order to determine the wmt ef fect of " single f ailures" which might not be confined to a single rod drive, we asked the applicant to perform "startup accident" analyses covering the entire spectrum of initial power levels (Ref.
Supplement No. 2, Question 4.9 and answer).
This accident assumes the uncon-trolled simultaneous withdrawal of all rods at maximum design speed, and further assumes that the excursion is terminated only by Doppler feedback and trip action of the power range nuclear channels. The applicant concluded:
"No fuel damage would result from simultaneous all-rod withdrawal from any initial power level."
From the preceding we have concluded that a single failure which allowed an extra rod group to be withdrawn, a situation less severe than the accident analyzed, would not cause fuel damage.
There will be two " speed limiting" features. One is the pulser (or clock)
The other is a which will be designed not to exceed a certain maximum frequency.
" Speed saturating circuit" downstream of the pulser which has the inherent property of not responding to a frequency greater than 125% of rated frequency.
There are two independent analog rod-position sensors at each rod drive, a potentiometer and an LVDT. There are two independent limit switches. In addition, the LVDT's will also generate limit signals. Thus, there are redundant analog and limit position indicating systems at each rod. Each analog signal at a rod can be fed into the individual rod position indicator.
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' Based on our analysis, we.believe that the applicant's criteria conform to our own, that no single failure cas produce an excursion which will breach the protection system, and that the proposed rod drive designs can be built in accordance with these criteria.
Reactivity is also controlled by a permissive system which allows manual dilution of the primary system coolant boron concentration when a particular control rod group reaches the fully withdrawn point. Dilution is automatically terminated when the rod group, driven down by the servo, reaches a prescribed position, or when the integrated dilution flow has reached a preset maximum.
We understand that these circuits will be designed in accordnace with protection system standards.
We agree with the implied criterion that no single failure should prevent automatic termination of dilution, when required.
In summary, we conclude that the applicant's design criteria relating to instrumentation and controls are satisfactory and that the proposed preliminary designs conform to these criteria.
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@u 70C0AL U$$ @ML1 5.0 Power 5.1 Description Each reactor unit vill generate electric power at 19 kv vhich vill be fed through an isolated phase bus to a unit step-up transformer where it vill be raised to 230 kv for Units 1 and 2, and 500 kv for Unit 3.
Two 230 kv overhead transmission lines vill carry power between Units 1 and 2 and the station switchyard which will be connected to the existing Duke 230 kv transmission line by six circuits:
two north to Jocasse, two southeast to Central and (upon completion of Unit 2) two east-northeast to Tiger.
From Unit 3, an overhead transmission line vill carry power between the station and the switchyard which will be connected to Duke's 500 kv transmission network by two circuits:
one to the Lake Norman area and the other to the Lake Wylie area, both being run in a general northeasterly direction.
An autotransformer will tie together the 230 and 500 kv systems at the station switchyard.
In addition, a separate 100 kv line vill be run directly from the gas-turbine generating station at Lee.
Each unit vill have its own 60 MVA startup transformer.
The 100 kv line vill terminate in a transformer at Oconee which vill serve all three units, as required.
Normally, each unit vill supply its own auxiliary loads directly from the generator via ths station auxiliary transformer.
Since each unit is being designed to accept a 100% load rejection, the primary source of power for the auxiliary loads in the event of system blackout vill be the unit generators themselves.
In the event of a unit trip, the power sources will be automatically switched onto t'ae auxiliary busses in the preferential sequence as follows:
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- a) the startup tre.nsformer bus (includes the Keovee Station 230 kv line) b) the other units' auxiliary electrical system (subsequent to the com-pletion of Unit 2) c) the 100 kv transmission line from Lee d) the Keovee Hydro Station 13.8 kv line.
The Keovee Hydro Station vill be located approximately one-half mile from the station switchyard, and will consist of two 70 ige generating units.
Each unit is essentially independent of the other and is pro-vided with its own startup equipment located within separate cubicles within the Keovee control room.
The initiation of startup is accomplished by control signals from the Oconee control room areas. Normal startup of either unit is by operator action while emergency startup is automatic.
Both units are started automatically and simultaneously on either of two conditions: if the external transmission system is lost or if engineered safeguards action is required.
Either hydro can be connected to either of two lines feeding the Oconee Station.
One is an overhead 230 kv line to the station switchyard; the other is an underground 13.8 kv line run directly to a 10 MVA transformer.
Four 125 v.d.c. batteries and six battery chargers will be supplied for Unit 1.
One pair of batteries and one set of three chargers will feed one 250/125 volt bus and the remaining pair of batteries and set of chargers will feed a redundant 250/125 volt bus (Ref. Fig. 8-3, PSAR).
Upon completion of Unit 2, this d.c. system vill serve both units. A third three-vire system vill be installed upon completion of Unit 3 Switching circuits vill permit any d.c. system to serve any unit.
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' Initially, there will be six 125 v.d.c. distribution panels, each of which will receive d.c. power from both three-wire d.c. sources through isolating diodes.
Two more panels will be installed with Unit 3 and will be similarly powered.
Four vital instrument busses (single phase) will be provided for Units 1 and 2, and will be independently energized from static inverters connected to one of the six d.c. distribution panels.
Two more vital instrument busses will be added with Unit 3 These will be powered, through static inverters, from the two additional d.c. panels.
In, addition, there will be three single phase 120 v.a.c. regulated instrument busses.
'Ihese will normally be connected to the 600 v.a.c.
busses of their own units through regulating equipment.
Provision will be made to switch over to the vital instrument busses, if necessary.
5.2 Analysis The staff review included a visit to Duke Power Company's Cowans Fo2d Hydroelectric Station on the Catawba River for the purpose of observing the startup and operation of hydroelectric generators similar to those proposed for the Keovee station.
Upon completion of Unit 1, off-site power will be available from the 100 kv system and from tne 230 kv system which feed power into oconee over separate transmission lines from Jocasse and Central.
An additional 230 kv tie to Tiger will be installed upon completion of Unit 2; and, upon com-pletion of Unit 3, a tie to Duke's 500 kv system vill be installed. All off-site lines will be energized from several power generating stations, and the Duke system is designed to withstand the step-loss of any single generating unit within its network.
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Gu $0C0AL U$$ @Lf Eedundant transformers vill be available to distribute power to engineered safety feature loads.
Transformer CTl (to the 230 kv system) and transformer CT5 (to the 100 kv system) vill be installed with Unit 1.
An additional startup transformer vill be installed with each of the other two units as they are completed, and each transformer vill be able to energize the emergency loads of any unit.
In view of the foregoing, we agree with the applicant that the proposed off-site power sources and associated distribution equipment are sufficiently reliable for the intended purpose.
We cannot, however, determine that these collective off-site sources are immune to the adverse effects of single failures. Fecent blackout experience elsewhere suggests that such immunity may not exist.
Accordingly, and inasmuch as the design and utilization of the on-site power sources are under the direct control of the applicant, we have analyzed the proposed on-site power system on the basis that the single failure criterion must be met.
U on loss of the external grid, redundant voltage and frequency sensing p
devices on each of the 230 kv switching station tusses will initiate, through separate and redundant channels, tripping of all 230 kv switching station
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isolation breakers, closing of all 230 kv switching station power supply breakers and startup of both Keovee units.
They will synchronize and be connected to the 230 kv lines.
one unit vill also feed the 13.8 kv under-ground line.
Shedding of non-essential loads (a requirement because of the limited capacity of the 13.8 kv/4.16 kv transformer) vill be accomplished by circuit breakers with duplicate trip coils energized from different d.c.
busses.
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Upon loss of the external grid and the tripping of a given Oconee unit (caused, for example, by a loss of coolant accident) the emergency power sources vill be automatically switched onto the emergency (4.16 kv) busses of the affected unit in the sequence stated in section 5 1 of this report.
Our analysis indicates that the sequencing system is essential to plant cafety since its failure could leave the emergency busses with no power.
We have been assured that this system vill meet the single failure criterion.
The Keovee hydro units can pick up emergency loads fron black start in t
23 seconds, which is adequate under design basis accident (DBA) conditions.
If tripped off line at full power due to a system disturbance, each unit can pick up full load in seven seconds.
Each unit's voltage regulator is equipped with a volts-per-cycle limiting feature which permits it to accept load at the outset and thus drag the loads up to full speed in synchronism with its own acceleration.
This serves to reduce the time required for the initiation of safeguards system action.
We concur with the applicant that it is a desirable feature.
The hydro plant is started by opening gates which are powered by hydraulic accumulators. Stored hydraulic energy is sufficient for three full opening and closing cycles.
Control circuits for emergency actuation of the accumulators vill be redundant.
A shear pin arrangement within the mechanical portion of the gate drive vill release a jammed or otherwise fouled gate from the others.
The protection system on the hydro plant vill be limited to only those parameters that will prevent generation of power, such as Senerator insula-tion breakdown or loss of field.
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@uf0C0AL 4133 @ E l In tne event both hydro units must be shut down briefly for maintenance emergency power can be made available to Oconee via the 100 kv line which can be isolated from the rest of the grid and kept continuously ener81 zed by one of the Lee station gas turbine generators set aside exclusively f or this purpose.
We believe this merits consideration even though it would allow a temporary non-redundant source of emergency power.We vill at the operating license stage assure that the integrated power systems
, in-cluding the reactor units, vill provide power for accident conditions i
The engineered safety feature auxiliaries are provided with redundan To maintain this redundancy, "le applicant has stated that these auxili aries will be connected to redundant busses such that safety feature auxiliaries performing the same function are connected to different busses Each of these busces is supplied from the redundant h160 volt main feeder busse s which are, in turn, supplied from the redundant sources described previou l s y.
We concur with the applicant in this design approach since it is an effective and simple way of implementing the single failure criterion (Ref. Fig. 8-1 PSAR).
Our review of the station battery system (shown in Fig i
. 8-3, PSAR) indicates that it is redundant and testable.
Voltage at each of the panel-boards, De-A, De-B...etc., is derived from redundant sources feedin g
through isolation diodes such that fai~ture of one source does not aff ect the voltage at the panel board bus.
Loss of voltage at a panelboard bus vill not negate the d.c. system function.
Our review also indicates that no single failure can cause a loss of volta 6e at all vital instrument busses.
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- We have been infomed by the applicant that means will be devised to test the diodes at power, and to determine (also at power) that no battery has become disconnected from its d.c. bus. We concur in these test pro-cedures.
In summary, we agree with the applicant's proposed criteria for the design and implementation of the off-site and on-site power systems, and we further agree with the design approaches to implement these criteria.
I 6.0 Containment Isolation Systems l
The instrumentai on and valve arrangements proposed for Oconee Units 1, 2 e Id 3 have been reviewed and found to conform to current design standards.
(Table 5-3, Fig. 7-2, PSAR; Question 4.8, Supplement No.1).
Since the design conforms to the following criteria for isolation systems, we believe that the proposed design is acceptable.
1.
Lines which penetrate containment and are open to the external atmosphere or to systems designed for less than containment design pressare shall be protected by redundant, automatic istolation valves (check valves are considered
',o be automatic valves) if they fulfill any of the following conditions:
a.
They are connected to the primary system.
b.
They are normally open to containment atmosphere.
Tney are certain to be ruptured in the event of the design c.
basis accident.
Exception:
Lines which must remain open subsequent to the accident chall be protected by redundant valves, one or both of which shall be remote-manual.
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Ventilation lines shall be isolated upon recei t p of " Safety In-jection" signals.
3.
Lines which have a low probability of rupture d urin 6 the accident (e.g., certain secondary system lines) shall be p rotected by at least one automatic valve external to containm ent.
Exception:
Lines which must remain open subsequent to the j
accident shall be protected by one automatic valve or one remote-manual valve external to contain h.
Circuits which control redundant automatic v l a ves shall be redundant in the sense that no single failure shall preclud 7.0 Rod Drives e isolation.
The rod drives originally proposed for the Oc onee units were to be a new design using a nutating disk drive system Due to development problems related to materials used in the nutating drive Duke Power amended the application to previde a drive mechanism utilizin g conventional components.
The drive mechanism now proposed is a rack a d pinion device driven by n
a synchronous steppin6 motor through a worm gear red ucer, undirectional clutch and ma6netic clutch, drive shaft and miter gear set.
The drive is operated in primary coolant up to the magnetic clutch where a buffer seal and rotary seal prevent leakage of primary coolant Tne mechanism is housed in two pressure housin which provides guides for the rack, hydraulic sn bb (1) the rack housing gs:
u er and spring stops for the rack, support for the drive shaft housin g and drive motor assembly and connection to the reactor vessel head, and (2) the d i r ve shaft housing which l
provides alignment a,a support for the rotatin6 d i I
r ve shaft and miter gear set J-
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These housings are designed to the ASME Code,Section III for 2500 psig and 650 F.
'Ihe two housings are joined at the pinion shaft by a bolted, double sealed joint. All gasketed joints are sealed by two Conoseal-type gaskets and pressure testing taps are provided between the gaskets.
'Ihe rack is directly connected to the spider of' the cluster control i
assembly by a tall and socket positive latch mechanism.
The rack is decelerated by a hydraulic snubber assembly contained in the rack housing
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above the pinion shaft.
The snubber causes deceleration of all of the moving parts in the driv. mechanism up to the magnetic and unidirectional clutch.
A bottoming spring vasher assembly is provided at the bottom of the snubber to absorb the end-of-scram stroke impact.
The drive shaft assembly including the miter gear set rotates to drive the pinion shaft.
The drive shaft is actually two shafts supnorted by bearing at their ends and at the center connection to prevent shaft whipping and to assure that the critical speed of the drive shaft is not reached during a scram.
The buffer seal assembly is similar to seals now in use for Consolidated Edison Unit 1, N. S. SAVANNAH, the SM Army reactors, Elk River and 20NUS.
The major departure is that the seal is working in a vertical orientation with a rotating shaft.
The drive motor assembly utilizes a worm gear reducer to prevent torque from being transferred to the drive motor in the event an upward force is applied to the rack. A unidirectional clutch will be provided within the magnetic clutch to provide for drive rundown after scram and to
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This type of device has been employed in the drive mechanisms of the LACEWR reactor to serve the same function.
Normal rod withdrawal and insertion requires that the magnetia clutch be energized.
Scram is accomplished by de-energizing the clutch.
The components of the drive that operate in reactor coolant vill be
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capable of performing their function at 650 F.
The seal water injection to i
the buffer seal a expected to maintain the drive components at a lover temperature.
As indicated in the description the rod drives are an assemblage of components of known characteristics.
Duke has proposed a development program to fully test the proposed design to demonstrate that the specified criteria (Section 3 2.4.3.1, PSAR) are met.
Our review of the proposed design indicates that no unusual problems are apparent.
We agree with the applicant's criteria and believe that the development progr u vill provide an acceptable control rod drive mechanism.
See Section 4.2 of this report for a discussion of the drive control and position indication systems.
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' 8.0 Steam Line Isolation Valves Steam line isolation valves have not been proposed for the present design.
The considerations which could lead to a requirement for steam line isolation valves are discussed below.
(1)
It is desirable to prevent blowdown of all steam generators in the system in case of a steam line break to insure that steam is available to drive emergency feadwater pumps and to provide a heat sink for the reactor.
The blowdown of both steam generators is prevented for the proposed units because the turbine used (General Electric) has stop valves on each line and d7es not, like the Westinghouse turbine, require a cross-tie before the stop valves te attain a constant steam tcu.perature.
In the proposed design a heat sink would be j
provided even if both steam generaters were blown down. This would be accomplished by supplying feedwater at 700 psig by electrically driven condensate booster pumps feeding through the normal feedwater pumps.
(2)
It is necessary to provide a leak-tight barrier in case of a loss of coolant accident with leakage through the steam generator tubes.
A leak-tight barrier in case of steam generator tube leakage af ter a loss of coolant accident is satisfied by the leakage characteristics of the GE stop valves.
We believe that the stop valves in this system can be considered a leakage barrier if the secondary system is tested for leakage integrity in conjunction with contain=
ment leak rate testing. The applicant has indicated that the leakage integrity of the valves will be demonstrated either by pressurizing the secondary system or by opening t'ne secondary system to the containment atmosphere during containment 1
leakage tests.
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The low feedwater inventory of the once-thorugh steam generator causes a decrease in prmiary coolant temperature of less than 60 F even if both steam generators blow down. This, combined with the larger increment of reactivity held by the control rods prevents an immediate secondary criticality.
The applicant's calculations indicate a maximum clad temperature of about 750 F during the trancient and no clad failure is postulated which would result in additional release of fission products to the primary system water. Dese calculations were therefore based on release of fission products contai:wd in the primary system water (1% failed fuel concentration) until the primary system temperature reached 200 F.
The applicant has calculated that a thyroid dose of about 30 rem to the thyroid would result at the site boundary for a complete break of 1 tube. Our calculations indicate a thyroid dose of less than 100 rem for a complete blowdown of the primary system (no plate-out of halogens was as-sumed in either analysis).
The shutdown margin is maintained even with one rod stuck out of the core during primary system blowdown (Question 16.2, Supplemen* No. 4).
Fo example, for a postulated rupture of 125 steam generator tubes coincident with a steam line break the minimum shutdown margin is 1.4% Ak/k with one stuck rod.
It should be noted that even if steam line isolation valves were installed, a break could be postulated on the containment side of the talve which would make isolation of the break impossible.
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' On the basis of the above discussion we believe that steam line isolation valves are not required in the Oconee units since (1) blowdown of both steam generators is prevented after a steam line break (2) the turbine stop valves will serve as a leakage barrier af ter a loss of coolant accident if the steam generator tubes leak and (3) isolation valves are not required to mitigate the consequences of a steam line break which precipitates a primary system blowdown.
9.0 Turbine Missile Analysis 6
In the PSAR, (Section 5.1.2.7.2) the applicant presented a turbine missile failure analysis for a 130% overspeed case.
In Amendment No. 4 (Question 11.1) a revised analysis at 186% rated speed was submitted. The analysis at the higher speed was a result of consultation with the staff, with the turbine manufacturer (General Electric) and with the staff's consultant, J. Proctor of the Naval Ordnance Laboratory.
For the case of "end on" impact of the wcrst missile, part of the last stage The wheel, tendon damage might occur but building integrity would be maintained.
applicant states that two adjacent tendons in the dame or a maximum of three hori-zontal tendons and one vertical tendon in the wall might be damaged by this missile.
These are below the damage limit calculated for loss of function of the building; i.e., more than five adjacent tendons in the dome or three horizontal and three vertical tendons in the wall.
The applicant states that a greater number of tendons could be damaged without building failure.
As an extreme worst case, the above analysis was repeated assuming no energy absorption in the turbine casing. No containment building penetration was calcu-
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lated although as many as five tendons might be damaged in the dome.
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~ Tornado missiles were also considered and result in no significant pene-tration.
Components in the auxiliary building are also protected from turbine missiles below the operating floor and the control room walls and roof would not be pene-trated by a turbine missile.
The applicant has stated that any components located outside protected areas which are vital to shutdown of the reactor will be redun-dant and located so that postulated missiles could not interrupt essential shut-down service.
The modified turbine missile analysis and the tornado miesile analysis was developed in conjunction with the staff and we believe that the approach and assumptions used in the analysis are reasonable. We conclude that the applicant's criteria for protection of shutdown equipment is adequate to protect viral equip-ment from postulated missiles.
10.0 Accident Analysis A number of operational transients were considered by the applicant in Section 14 of the PSAR including rod withdrawal during startup and from power, moderator dilution, and loss of coolant flow and no radiological hazard was found to result. Rupture of steam generator tubes wa' postulated and fission product release through the condenser resulted in doses less than Part 20 limits at the site boundary.
A steam line f ailure was analyzed which resulted in the release of the fissios products contained in the secondary system (which are accumlated due to minor tube leakage in the steam generator).
The doses from this accident were within Part 20:
limits. An analysis of a steam line break coincident with multiple tube failures i
j was discussed in Section 6.0 of this report and resulted in environmental doses u
within Part 100 guidelines.
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. 10.1 Rod Ejection Accident The rod ejection accident was presented in Section 14.2.2.2 of the PSAR The maximum worth and further elaborated in Question 6.2 of Supplement No. 1.
full power was stated to be 0.2% A k/k and the maximum worth of a control rod at at source level 0.5% A k/k. The parametric study presented varied rod worths from 0.1% to 0.7% A k/k for both the full power and source levels.
For the ejection of a 0.2% rod f rom full power the maximum enthalpy in the hot rod was calculated to be 157 cal /gm. The applicant's sensitivity i
analysis (Fig.14-24, PSAE) indicates that ejection of a rod worth 0.6% f rom full power would result in a hot spot enthalpy of about 200 cal /gm.
An ejection of a 0.5% A k/k rod at source power was calculated by ejecting The results of the a 1% rod with the core initially 0.5% A k/k suberitical.
analysis indicate a resultant peak power level of about 39% full power.
Parameters varied in the sensitivity analysis included rod worth ejection Vesscl accidents to obtain an estimate of the margin to failure of the vescel.
failure was estimated to occur for ejection of a rod worth of 2% 4 k/k.
The applicant also stated that core internals would not be damaged by ejection of a 1% rod since no fuel melting was calculated for that rod worth.
We believe that the applicant has performed suf ficient seasitivity analyses to show that the rod ejection accident will not result in intolerable consequencesa 10.2 Loss of Coolant Accident the applicant has proposed a safety in-As discussed in our first report, jection system (including core flooding tanks)
.ich are designed to protect the The core cooling analyses for all core for the full spectrum of break sizes.
break sizes have not yet been compelted and the calculation of blowdown forces i
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' has not been completed but analyses are underway in both cases and will be looked at before the operating license stage, probably on subsequent reactors We believ2 that the present statc for which B&W is the steam system supplier.
of analysis is sufficient to allow issuance of a construction permit for these units.
The applicant has calculated the environmental consequences of this acci-(and we have duplicated the analysis) for the expected course of the acci-dent dent and for a " maximum hypothetical accident" in which 100% of the noble gases I
and 50% of the haolgens are assumed to be released to the reactor containment.
The reactor building leak rate was assumed constant for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at 0.5%/ day As discusse3 in our first report one-half for the duration of the accident.
of this leakage is assumed to pass through the penetration room filters where the halogens are removed with a 90% filter efficiency.
The meteorological model used as a basis for dose calculations is based on It is the drainage of air down the river valley with no loss from the valley.
h assumed the.; Type F diffusion conditions prevail during such times, and that t e wind speed is low.
For the first 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, the wind speed is taken to be 1 m/sec, and the dif-fusion is calculated at the site boundary. At this point, there are two hills which confine the valley so that the total cross sectional area below their tops For a plume uniformly distributed in such (775 feet msl) is 13,500 square meters.
-5 a space, the equivalent X/Q is 7.4 x 10 Doses for longer exposure times are desired at the low population distance A narrow point in the river valley exists 5.8 river miles from about 6 miles.
tion the site boundary in the vicinity of Clemson, S. C., and the valley cross sec
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It is noted that on the way to this point, the Keowce River is joined by a similar sized stream, the Little River. By comparing the valley cross-section at their con-fluence, it can be shown that the air flowing down the Keowee valley is diluted by a factor of about 0.48 at this point.
Using a wind speed of 1.5 meters /sec for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the value of
-6 X/Q in the vicinity of Clemson is 3.16 x 10 For the remaining 29 days over t
j which the dose was calculated, the same conditions are assumed to prevail 35% of
-6 the time, giving a value of X/Q for the thirty day dose of 1.10 x 10 These values of diffusion factors were used as appropriate in the dose calculations for all accidents.
The two hour dose at the site boundary using the above meteorological model is about 250 rem to the thyroid and 2 rem whcle body. The thirty day dose at the low population distance was calculated to be 150 rem to the thyroid by the appli-cant. However this did not include the initial 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> dose.
The total of the
" thirty day" and "24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />" doses would be 220 rem to the thyroid and 5 rem whole body.
10.3 Accidental Liquid effluent Release Section 11.1 of the PSAR discusses the inventory of the radioactive wastes in the facility during operation with 1% failed fuel. A single failure analysis of the waste disposal systems is also presented and shows that a single equipment failure would not result in release of liquid effluent.
The release of activity from a vaste gas tank failure after operation with one percent failed fuel is calculated to be within Part 20 limits.
Question 2.3 of Supplement No. 1 and Question 12.5 f Supplement No. 4 g
discuss available dilution factors to public water intakes in the vicinity and
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@u #0CBAL U$2 @Mid potential doses resulting from an accidental spill of liquid waste. The cases considered are (1) a spill of liquid waste storage after extended operation with one percent failed fuel and (2) a spill of liquid waste storage after a loss-of-coolant accident. The calculations, which utilize conservative dilution factors, show that accidental discharge of operational stored wastes would result in doses below Part 20 limits. The analysis also illustrates that an extended (and unde-tected) release of wastes collected after a major accident must be postulated before Part 100 doses would be exceeded.
The only hazard to the public drinking supply would be af ter a major accident when (it is expected) comprehensive monitoring programs would be undertaken. The doses calculated also assume no corrective action at the public water intakes.
We believe that the analysis presented illustrates the potential magnitude of the problen and the corrective measures which are available and that the accidental release of liquid waste would not result in excessive exposure to the public.
11.0 Research and Development In Section 1.5 of the PSAR the applicant has identified a number of areas in which research and development programs are being pursued. We agree with the appli-cant on the areas in which research and development is required as listed below.
However, items 1, 4 and 5 contain additional considerations beyond those initially specified by the applicant.
(1)
Once-through steam generator Steady state conditions and operational transients will be investi-gated in conjunction with the control system to be used. We believe that l
j vibration tests, including generator response to primary system blowdown,
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(2) Control rod drive unit test The prototype tests outlined by the applicant to be conducted under operating temperature, pressure, flow and water chemistry should provide (Section information on the operability and reliability of the system.
3.3.3.4, PSAR).
(3) Incore neutron detectors The self powered units are currently under test in the Big Rock Point Nuclear Power Plant.
(4) Thermal and Hydraulic Programs The applicant has proposed scaled flow distribution tests on the vessel and internals and rod bundle tests to determine local mixing As discussed in Section 3.0 of this report, we and flow affects.
believe that further work must be done to determine the limiting heat fluxes at various positions within the fuel bundle if the transfer cata.
design is to be based on the B&W heat We also believe that the applicant should include core cooling (5)
Specifically (a) the completion of the in the development program.
analysis of the spectrum of break sizes in the loss of coolant acci-(b) the development of the analytical techniques for determining dent blowdown forces on reactor internals and (3) demonstration that the injection coolant will cool the core including core bypass or formation of a vapor lock.
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' 12.0 Conclusions 12.1 We conclude that the following areas, discussed in this report, have been sufficiently supported for purposes of a provisional construction permit:
(1)
Instrumentation (2) Power (onsite and offsite)
(3) Containment isolation systems (4) That steam line isolation valves are not required on this plant j
(5) Turbine missile analysis (6) Accident analysis In addition to the above items, sufficient supporting information has been sub-mitted (as discussed in Report No. 1) in the following areas:
(1) site, (2) core design, (3) reactor coolant system, (4) containment, (5) engineered safety features and (6) sharing of auxiliary components between units.
12.2 We conclude that, although the following areas have not been completely resolved, sufficient information has been provided for the purposes of a pro-visional construction permit.
(1) The research and development items as summarized in Section 11.0 of this report including our recommendations on the programs.
(2) Those items summarized in Report No. 1 and not included in Section 11.0 of this report:
(a) an acceptable value of the moderator temperature coefficient, (b) xenon oscillations and (c) the ireadiation surveillance program.
(3) Design details associated'with the switching arrangement and redundancy for the emergency power systems.
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. 12.3 On the basis of our evaluation of the Ocon in this report and in Report No ee Units 1, 2 and 3 as discussed
. 1 to the Committee dated May believe that 24, 1967, we there is reasonable assurance that this f operated at the proposed location without und acility can be built and of the public, ue risk to the health and safety t
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