ML20125B962
| ML20125B962 | |
| Person / Time | |
|---|---|
| Site: | Monticello  |
| Issue date: | 11/14/1969 |
| From: | Moore V US ATOMIC ENERGY COMMISSION (AEC) |
| To: | Boyd R US ATOMIC ENERGY COMMISSION (AEC) |
| References | |
| NUDOCS 9212100230 | |
| Download: ML20125B962 (20) | |
Text
{{#Wiki_filter:e o
\ hk NOV 141969 Roger S. Boyd, Assistant Director for Beactor Projects, DEL I 11110 : Saul Levine, Assistant Director for Reactor Technology, DEL NotTHERN STATES POWER COMPANY, MONTICEL14 WUCLEAR CENERATING PLANT, 1911T 1, N DJ. 50-263; SAFETY ANALYSIS The safety analysis relating to the Protection System and Emergency Power System is transmitted for inclusion in the report being The pre-pared for consideration by the ACRS at the December meeting.
paragraphs of the report listed below identify unresolved problem areas or areas which require confirmatory information: Subiect_ Paraaraoh < 1.1.1 Flow-Biased Flux Reactor Trip 1.1.5.1 ARS - Pressure Interlock 1.1.8 Standby Sas Treatment System 1.1.10 Single Failure Criterion 1.4 F.nvirossental Testing Original signed ti Voss A. Moore ' T. A. Moora, Chisf Instrumentation and Power Er-852A Technology Branch DRL 1&PT5 TAI Division of Reactor Licensing
Enclosure:
Safety Analysis cc w/eac1: Distribution: F. Schroeder 'M - huppl. D. Haller DRL Reading D. Vaasa11o AD/RT Reading 1&PTB Reading bec: S. Levine
* ~~ I R. DeYoung V. Moore T. Ippolito DAD RT, AD/RT orncr > __ RL11&PTB.., ...D.RL:1&PIB _
RDe,Y ,ung,,,,, . _ e,__ _ suRN ate PPo ..91ese... .V. .M. .o. .. .. . om > 11/M./.6L . 11/1f169 .11)).f.]69.. .. 11//7 /69 mm ac us(n...un ,, _ ,_ ,,m , _ , _ , . 9212100230 691114 PDR ADOCK 05000263 A PDR i ___
NOV 141969 IONTICELLA NUCIAAR ENERATING STATION INIIT 1, DOCERT No. 50-163 EVALUATION OF THE INSTRUMENTATION, CONTROL AND AUKILIART ELECTRIC POWER SYSTEME
- 1. Introduction
)
- The instrumentation, control and auxiliary electric power systems have been evaluated against the Commission's General Design Criteria (GDC) and - for the Proposed IEEE Criteria for Nuclear Power Plant Pro-1 L taction Systems (IEEE 279) dated August 28, 1968. A comparative review was made with the Dresden Nuclear Power Station, Unite 2 & 3, the design of which was evaluated in our Esport Number 2 to the_ ACRS, dated August 18, 1969. The reactor protectics instrumentation and control systems as well as the instrumentation which initiates and controls the engineered safety features were found 'to be functionally the same.
However, the auxiliary electric powersystems were found to be unique to each application. During the course of our evaluation, we had several meetings with the applicant and his representatives. Two of these meetings were devoted to the review of elementary diagrams. In addition, a visit was made to the site on September 17 and 18,1969, for the purpose of observing the physical arrangement and installation of the instruman-tation, control. and auxiliary electric power systems.
' NOV 141969 Monticello Unit 1 2 Our repou is limited to the SWR generic problem areas first identified in the Dresden, Units 2 & 3, evaluation, areas of destga i
for which new information is available, and design features unique to Monticello. Specifically these areas ares i
- a. 9WR Instrumentation Generic problem Areas
- b. Reactor protection system and Engineered Safety peature 1 (ESP) Installation Criteria
- c. Auxiliary Electric power Systems
- d. Environmental Testing of Electrical Components and 1
instruments 1.1 SWR Instrummatation probles Areae 1.1.1 Flow-Biased Flus teactor Tris Six Average Power Range Monitor (APEM) instrument chmanals are provided for continuous monitoring and indication of reactor power and to supply trip signals to the reactor protection system. As a result of our expressed concera, the applicant has modified his design to provide reactor trip signals which are automatically biased by reactor coolant recirculation-flow. The design modifications differ from our understanding of those proposed in Dresden, Units 2 & 3 and those shown in elenantary diagrame for Oyster Creek, Unit 1, and Wine Mile Point. The difference is that in the Monticello design the fixed high flux reactor trip is being removed whereas la the other designs this trip remained in effect. Our concern with tha Monticello design _
NOV 141969 Montieello Datt 1 3 is that the instrumsatattom d ish monitors rectreulation flow and provides the flow bias input to the AFEN chaamals is not inusame to . single failures. , he applicant has stated that his design is such that these failures will not negate the ability to trip the reactor et the fixed 1201 high flux level. Further, he stated that except for the flow bias instrumentation, the design meets the requiressats of IREE 279. h is exception is the same as that identified and accepted in Dresden, Nine Mile Point and Oyster Creek. The applicant will submit elemen-tary diagrams of his design to assist us in confirming the suitability of these modifications. We anticipate no problems in this regard. 1.1.2 Rod Block Monitor (RSM) he REN is designed to initiate a rod block under the worst per-mitted bypass and detector failures to prevent local fuel damage during single rod withdrawal errors starting with any permitted power and flow condition. We RBM consists of two channals of instrumentation which are effective only during rod selection and movement above 301 power. The Dresden Units 2 6 3, review identified' single failureswhich would preclude rod block action een required. he consequence of failure of the REN is evaluated and documsated in the Dresden, Units 2 6 3, application and'is applicable to Monticello. his evaluation shows that during reactor operation with certain limit-ing control rod patterne, the withdrawal of a designated single rod .
..m. 2 .- m . ;,...._ m -- -- _.. _ _ - __ - _ _ . - - . . .~
a 4 NOV 141969 Monticello Omit 1 eould result in one or unre fuel rods with IEMFR's toes than 1.0. 4 As in the case of Dresden, the judansat was made (1) that the 11mittag rod patterns are unique, (2) that during operation with such patteras,. , testing of the REN system prior to withdrawal of such rods and on a daily basis thereaf ter will provide adequate assurance against ingroper rod withdrawals. The applir wt has agreed to the sans conditions and has included the requirement in the Technical Specifications. 1.1.3 Containment Sorav Actuation The containment spray eyetem consists of the same components as the LPCI plus the additional valves and piping required to direct cooling water into the containment spray headers. These- copats
! ere arranged in two loops and the controls for each loop are located in the same logic matrix associated with its corresponding LPCI loop. The admission valves for each loop are manually tattiated by a switch in the control room. The remote manual controle - for these valves are interlocked so that opening is not possible unless primary containment pressure is above 1 peig and reactor water level' inside the core shroud is above 2/3 core height. Our review of these. inter-locks revealed that single failures could result la the loss of both-containment spray loops as well as prevent termination of the flow in one loop when high containment pressure is reduced. Additionally. .
a single failure could result in the initiation of a single loop _ prior to the core being adequately covered (Failure -of the. 2/3 core shroud level interlock). l
~
Monticello Unit 1 5 NOV 141969 The applicant has revised the desian of this instrumentation. Our review of the elementary diagrams of this revised design revealed that:
- 1. Initiation logic has changed from two out-of-two per loop I to a one-out-of-two taken twice logic per loop. This corrects the problem concerning initiation of the containment spray system.
- 2. Single failures remain which would prevent a single loop from being shut down either manually or automatically.
- 3. Single failures remain which would permit the initiation of a loop with the water level in the core below the 2/3 core shroud level.
- 4. This revised circuitry did not include the provision for test-ing the sensors.
Items 2 and 3 above are not required to be corrected since these failures do not constitute any danger to the public health and safety. The applicant has agreed to incorporata- test jacks, test lights, and alarms in his design of containment stray initiation sensor circuitry. We conclude that the design is satisfactory for this application, 1.1.4 Diversity in Initiation of Core Surav and LPCI The inputs for initiating the sore spray system and its functionally i
Monticello Unit 1 6 NOV 141969 redundant and diverse counterpart, the LPCI system, are derived from signals which are a direct measure of the desired variables, he pumps for these systems are started by diverse (high containment pres-sure or low water level) eignals. D e admission valves of both systems, however, are operated by the same non-diverse but redundant reactor low pressure signals. Wese low pressure signals also compromise the independence of the core spray and 1.PCI systems. De applicant has agreed to provide equipment divers-cy in the form of two different types of pressure sensing devices. His design change is the same as that proposed and accepted durinq J review of Dresden, Units 2 & 3. 1.1.5 Auto-astief system (Amsl 1.1.5.1 Pressure Interlock he applicant has revised the design of the ARg to provide sa interlock which will prevent automatte initiation unless the low pres-sure core cooling systems are available. B is interlock function receives input signals from sik pressure switches monitoring the discharge pressure of the six pumps (two core spray pumps and each of the four LPCI pumps) such that one switch monitors one pump. De circuitry is arranged such that the operation of any one switch (one pomp operating) will permit the automatic initiation of the Agg. A rsview of the elementary diagrams revealed that single failures in - l L
. ....~.,...- . - - .. ,- - - - ~ - - . . - - - . - - - - - - - - - -- - - --
(
~ NOV 141969 Nonticello Unit 1 7 this intsrlock function will not prevent autoestic initiation. How-ever, single failures will result in automatic initiation with mone
- of the low pressure core cooling' systems available. Additionally, the design does not incorporate provisions for the testing of these
! added sensors. 1 The applicant states that the design of this pressure interlock satisfies his design criteria in that single active component failures will not negate the purpose of this function. He has been advised that we require that this pressure interlock be capable .of performing its intended function in the presence of any single failure -(single failure as defined in IEEE 279), and that the design incorporate provisions for sensor testing. '1his problem remains unresolved. 1.1.5.2 ARS Manual Initiation The applicant has modified the design of the ARs to make manual initiation h=mme to single comumpent failures. To provide this capability, provisions were made to automatically and independently provide each reitef valve with an alternate soure of-125 vde control power upon loss of the preferred source. This is accomplished by the provision of a preferred 125 vde source monitor. relay for each valve. Loss of the preferred power source._will de gnergiae.the relay, open the circuit from the preferred source and thea . complete the circuit to the alternate-source. This same modification has been incorporated in each of the two automatie initiation logic matrices.
l 1 e i S^ , Monticello Unit 1 8 NOV 141969 Our review of the elementary diagrams did not reveal any area I~ which causes us concera except to ensure that the relay maatters are i in fact the type which function to " break before sake." he applicant has subsequently confirmed that the proper relay is being used in this ! design. l ne justification and/or reasons for using single cosoosent failure criterion as a basis for this design is addressed in section 1.1.11 of this report, i 1.1.6 Testability of snaineered safety Feature Instrumentation l We design of the engineered safety feature circuitry did not 5 include provisions for non-aisbiguous periodic testing. The applicant has provided in the design of the emergency core cooling system (ERR, Core spray, RPCI and ARs) instrwesatation, permanently installed i test jacks, test lights, and alarms. D ese added features will facili-tate periodic testing, and reduce the need for using clip leads or i to disconnect wiring, sven with these features, non-embiguous periodic testing remains heavily dependent on written procedures. Dese procedures are scheduled to be completed just prior to core load-ing. However, preliminary procedures will be available in time for preoperational testing. We will request the Divistom of Compliance I to review Op;se procedures to assure that all credible random faults are detectable during periodic testing. t l
.- - - . . - - - - - . - _ - . - - -_~
Noaticello Unit 1 9 NOV 141969 1.1.7 anactor Buildt== (RS) ventilation Isolation isolation of the R3 ventilation system and Laitiation of the. Standby Oss Treatment System (SOTS) can be actuated by (1) the radia. tion monitors located in the ventilation exhaust plenua, or (2) the area monitors located abow and to the side'of the refueling pool. Two monitors are provided in each of these areas. Isolation of the RB ventilation system and initiation of the SGTS occurs on one of-two upscale trip or two downseale trips from either set of monitors, our review of the elementary diagrams reve'aled that_the monitors located in the RR vent exhaust plenum as well as those located over the refuel- 1 ing pool are inunune to single failures and are capable of being tested during operation. This design is'similar to that proposed and accepted in our Dresden, Units 2 & 3, review. 1.1.8 St 9 by Gas Treatment System (Sots) The SOTS consists of two separate and-redundant full capacity filter / absorber /faa units. The major components are shown in Figure 5.3.1 of-the F8AR. This system.is provided to. maintain'a_seall negative pressure (0.25 inches) in~ the reactor building under isolation __ conditions to minimise grosed level release of airborne radioactivity. This system is- initiated by either low water level, high contatmasat pressure, or high radiation signals. The radiation monitors provid-ing these signals are described in section 1.1.7 of this report. l
,r.,,- -
m , _ - - - , , -w+.e,-n-%, ,e, , - . - %,--w-- -w,,,,w.,.,-
i ~ Monticello voit 1 lo- . NOV 141969 - Our review of elementary diagrame reveals that although eseh l / l J t equipment-train is physically and electrically separate from the i other, the control instrumentation is not-independent. In the proposed design, an equipment chata is dependent upon the fatture ' of its redundant counterpart for initiation and operation. The applicant was advised that the lack of independence creates undue i vulnerability to single failures. Further, he was advised that the f l j j design should be changed to satisfy the same basic principles of independence required in the reactor protection eyeten. The applicant stated that the design will be changed to _ assure that the control circuitry which detects the failure of the first j i equipment train and actuates its redundant counterpart will meet f t IEEE 279. We will advise the Committee orally of the ~ schedule for l i submission of the design revisions for our evaluation. 1.1.9 confirmation of WPCI Operation I The NPCI system is not individually designed to meet- the single failure criterion. It is, however, functionally redundant to the Automatic Relief System (ARS) in conjunction with the core spray or-
~
LPCI systems. Although our review of the contro1 and protectics
- circuitry of the RPCI has indicated that the design is capable of meeting its functional requirements, we believe that a confirmatory 4
field test of'this system is necessary. The applicant has agreed to 9
, - . . , - -- - - - +-,-,--v ---v.al* wa,-- a- ~---m.a
1 i 11 3 4 1969 Monticello Onit 1 - p r j perform preeperettomat tests with actual or simulated signals to l verify its fumettomal capability. We have sonotuded that this system -- i
' is acceptable.
i i 4 1.1.10 Sinale Feiture Criterion t ' The applicant lists in the design basis for engineered safety I feature initiati= and control instrumentation the requiremsat that no sinale_ couponent failure shall prevent a protective action. Thel f i definition of single component failures, as we interpret-it, does s I not agree with the single failure criterion defined in IEEE 279. -The applicant has been requested to identify _all reactor protection and
- ( engineered safety features instrument systems to which the single
~l t
j component. failure applies and to provide justification for taking l exception to IEEE 279. The applicant has agreed to submit this I information in a forthcoming amendment. Further, we are led to 1 believe that all protection systems,-when one considers functional l n If thia redundancy, seet the single failure criterios of IEEE 279. f is casiirmed .in the forthcoming amendment, this aseter _will be = i resolved. Otherwise each system will have to be re % valuated to determine its effect .on public health and safety. We will report 4 orally to the Committee in this regard, 4 i-i i
. . . ,_ _ i
;r,, . . _
4
~
Monticello Unit i 12 N I 41969 1.2 Reactor Protection Systes (RPS) and t==ineered Safety Features f") lasta11stion Criteria The applicant has documented his criteria for the installation - of the RFS and EST. We conclude from our review of these criteria that if properly implemented, the probability of loss of redundant channels from a single cause such as fire will be acceptably low. These criteria include identification of safety related circuits and' counponents from like items not related to safety. i Our site visit, however, revealed that the physical and elec-trical installation of the RFS instrument sensor located in the i turbine building differed considerably from those sensors located j
! in the reactor building. In some instances, the installation of
' these sensors does not satisfy our understanding of the applicaat's design criteria. The RPS sensors located in the turbine building are those which monitor for condenser low vacuum, control valve fast closure, turbine first stage pressure, main steam line low pressure, and turbine stop valve closure.
Of the aforementioned sensors, only the installation of the con-trol valve fast closure sensors is judged unacceptable since we cannot condlude that'the installation is tr=saa to a single cosmion fault or event. Additionally, provisions are not included in the design to permit testing of these sensors during operation. The ] l l I l _ . . - ._ - . . . . _ _ . . . - . . ~ - _ . . - - . . . . _ . .
i
- NOV 141963 13 i
Monticello Unit i I sensors monitoring the action of the control valves are four pressure switches sensing the turbine acceleration relay oil pressure. Loss of oil pressure on this relay results in closure of the control valves. hese sensors were observed to be mounted in an enclosure somewhat l smaller than one cubic foot in size and located at the front end of i , the turbine. Within this er. closure, the four switches are mounted Two on a vertical steel plate of about 5/8 inches-in thickness. l switches are mounted on each side of the plate. We cables and sensing { j lines to these sensors were not installed; hcwever, it is evident i they would all have to be within inches of each other, f
; i The applicant has agreed to modify the installation of these l l l
l
' sensors to provide greater assurance against their failure from a i
! single common event and to provide a means to peruf t testing of these i
l i j sensors during power operation. We will request the Division of I compliance to assure that this installation as well as the installa-tion of the other sensors located in the turbine building satisfy h l j the stated criteria and are satisfactory. f l ! 1.3 Auxiliary glectric Power systems l-
;I
! i 1.3.1 Offsite Powet
' offsite power for the Monticello Unit will be supplied from the
}-,';
plant 345 kv and 115 kv switchyards. Power is supplied to the switch-f l yarda via multiple 115 kv and 345 kv linas over multiple rights-of-way. I ! l l l
Montice11e Unit 1 14 NOV j 4 7969 l 1 d Each of these lines independently has the capacity to supply sufficient 1 The power for safe shutdown or the engineered safety features loads. j j incoming auxiliary power requirement to provided from the Northera 1 ! I States Power network and from the Upper Mississippi Valley power Pool, the United Power Association and the Interconnected Systems Group. i Studies were conducted by NSF to determine the network characterie-ties and the system stability. From these studies the app 1 tant l concluded that loss of the Monticello Unit (460 W) could be tolerated and that the spinning reserve (% 618 W in 1970) and the quantity of kinetic energy of the prime movers and generators would minimise l i the system disturbance. Thus, the loss of the Monticello Unit should not result in the inability of the grid to supply power to the station. Initially the offsite transmission systems will be terminated at the substation switchyarda in ring bus configurations; later, the i I switchyards will be converted into a breaker-and-one-half arrangement. We have concluded that either breaker arrangement is adequate in that a single failure cannot negate the ability of the grid to supply off-site power to the engineered safety feature loads. The switchyard breaker controls are not specifically designed 4 to meet the ABC General Design Criteria (GDC 39). A single battery system is used to supply the control and power requirements for the actuation of- the switchyard circuit breakers. Our review of this
-- - + g- -m-. , , , , , ~ , y -qv- p,,-g--
. ... .., ~ . . . . - . , , _- .- ... . . -- - . - . -- - --.- . . .
i 1
- 1 heaticello Unit 1 13 NOV 141969
! design reveals that loss of voltage in this battery system is alarmed-i in the acetrol room and that operating personnel are tratand to
-ef fect local (at the switchyard) circuit breaker operations in accor-
- dance with an approved procedure. We conclude from the aforementioned-
! i i
l- i provisions and the applicant's stated maintenance procedures that ' i $ k i there is adequate assurance against undetected failures and that .l i t i repaire can be readily effected in this system. As a result, we l i ; i judge that the addition (backfitting) of a redvadant switchyard j i battery system would not add significantly to the puhtic health and
- eafety and is _ not required for this app
- %ation. -
i ! From the substation switchyards, two overhead lines, one at 115 kv L [i I and one at 13.8 Lv,are run to the station transformers at the reactor 1 i ( i' building approximately 1300 feet away. Two transformers connect the I 115 kv and 13.8 kv lines into the station's two 4.16 kv essential buses. Power is supplied by the reserve transformer during startup and !
~
! I shutdown and by the unit auxiliary transformer during normal operation. I I j Upon unit trip, na automatic fast transfer te- the reserve transformer l i will occur. Inability of this transformer to supply auxiliary power i i results in an automatic transfer to the'second reserve transformer.
}
This latter source is of smaller capacity but to capable _of providing l-sufficient power to assure safe shutdown or supply a11' engineered . safety feature loads. i l. I' l l 4 ?
<y .. - _.
.. - . _ . . m., , , _ , . . _ , , , _ _ . _ . _ , _ , . , . _ , . . , , . _ . _ . . . . . , , _ . , . .
~
Monticello Unit 1 16 NOV 141969 We have concluded that, because of the capacity and redundancy provided and the relative independence of the redundant power sources, the offsite power system meets Criterion 39 and is acceptable. 1.3.2 Onsite Power The design of the auxiliary power system utilizes the split bus concept. The 4160 auxiliary buses are in eight sections. Two of these buses, and their associated 480 volt load centers, supply power to the essential loads. The remaining buses supply power to all other plant services. No provisions are made to automatically connect j redundant buses togethe upon a loss of power. -This assures that a fault in one system will not be propagated to its redundant counter-part. Two diesel generators provide power for the two essential buses, each bus having its own source of power.. Each diesel generator is rated at 2500 kw (continuous), and 2750 kw (2000 hour). Both units start automatically, and are ready to accept load af ter ten seconds, upon initiation by either of the low reactor water level, high contain-I ment pressure, or loss or potential loss of offsite power signals. The diesel . generators are independent with respect to physical location (separate Class I rooms), cooling water, air start systems, control and sequential' loading circuits, and fuel supely. Makeup to er.,.h unit's day tank (8 hour supply) is from a single fuel oil storage tank which has suf ficient fuel capacity for one week's operation of one unit at full
_ _ _ _ _ _ _ _ . _ _ . .. _ _ _ - _ . . . _ . . _ _ _ ~ . _ _ . Monticello Unit 1 17 NOV 141969 i power. gedundant fuel pumpe are provided to pump fuel from the f, storage tank to the day tanks of each diesel. 4 i ne required loads, after two hours in a E3A and loss of off-- site power condition, total 1527 km (design) and 2745 kw (maximum) for each diesel generator. S e applicant has stated that, even though the continuous rating is exceeded, the 2000 hour and 30 minute ratings provide adequate margia for the intermittent and short tie, loads l especially if load diversity factors are considered. W ile we da not agree that design loads should exceed the continuon rating, it is J our judgment that, since tha automatically enerQsed loads are within the continuous rating and the other loads are under operator control, this is not sufficiently significant to re. quire a backfit of larger 1 { capacity diesel generators. Additio *ily, it is our understanding that the applicant's operating procedure will require load sharing l l l l l beyond the two hour accident period to further assure diesel loads
' remain within the continuous rating.
D ree (250 v. 125 v, and 24 v) d-c power systems are provided. Dese systems are insulated from ground and each is provided with a j ground detection system to annunciate the first ground. All batteries I are mounted on racks designed to withstand the maximum earthquake. Each system, as well as the radundant components within each system, is physically and electrically separated from the _others. We conclude i that, since a loss of any d-c bus does not result in the loss'of any protection functica, the design of the d-c power system is adequate. i I
. NOV 141969 Nonticello Unit 1 18 As originally designed, the redundant 125 y battery systems were interr.,anected by automatic transfer switches. This design compromised independence and was susceptible to failure from single common events.
The applicant has agreed to provide manually controlled switches in lieu of the automatic switches. We conclude that this change provides further assurance that no single failure in the 125 v battery system will result in the complete loss of any protection function and is, therefore , considered adequate. We conclude that, because of the capacity and redundancy provided and the ralative independence of the redundant features, the onsite power system meets Criterion 39 and is acceptable. 1.4 Environmental Testina A study was made by the applicant to determine whether the e1*c-l trical equipment used in the reactor protection and engineered safety i features could perform their design functions in an accident environment. The electrical equipment located in containment thet must function consists of a-c electric motor-operated valves with their associated operators and electrical cabling, and solenoid actuaton for main steam isolation and ARS v.1ves. The applicant has stated that quali-fication testa have been satisfactorily completed on prototype pieces I -. of equipment and a summary report will be submitted. We conclude that l- l l
Monticello Unit.1 - 19_ NOV 141969 -{ these equipment are satisfactory for use,in this; application provided the test report does not reveal areas of concern. -: We will report orally to the Cosmaittee in this regard. The instruments inside containment that must function are' limited to the sensors used for the reactor waler level measurements. Test results included in the FSAR show that the sensors remain operable and maintain their required accuracy during and subsequent to rapid depressurisation of the vessel. We conclude that this instrumentation is satisfactory for this application. The appitcant has proposed a program for assuring that Class I instrumentation meets seismic requirements. Our review of the criginal prograu plan submitted in the FSAR found it to be incoglete in scope . since such vital Class I systems -as Standby Gas Treatment, Contaiwant '
-Isolation, and Emergency Electric Power are- not included. The appli-cant has re-examined the scope of the program to include all Class I systems.
The applicant has stated that this program will be coupleted' for the General Electric supplied systems by December. 31,1%9, and -for-the balance of plant systems by . - The applicant will be required to correct any problem-that is judged significaut-to the public health and safety prior to cosassacament of commercial operation. .
' _ --_. ____ ______.___.__._______-_-___m_m_-___m_ _m_.._.____-.a _._m . _ _ _ _ _ _ _ _
__.-}}