ML20052F572
| ML20052F572 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 05/07/1982 |
| From: | Diab S Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20052F540 | List: |
| References | |
| NUDOCS 8205130201 | |
| Download: ML20052F572 (10) | |
Text
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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of TEXAS UTILITIES GENERATING COMPANY, Docket Nos. 50-445 ET AL.
50-446 (Comanche Peak Steam Electric (Application for Operating License)
Station, Units 1 and 2)
AFFIDAVIT OF SAMMY DIAB _
I, Sammy Diab, being duly sworn, do depose and state:
Q.1.
By whom are you employed, and what is the nature of the work you perform?
A.1.
I am employed by the U.S. Nuclear Regulatory Comission, Division of Systems Integration, Reactor Systems Branch
("RSB"). A copy of my statement of qualifications is attached to this affidavit.
Q.2.
What is the nature of the responsibilities you have regarding the Comanche Peak Steam Electric Station ("CPSES")?
A.2.
I was the Reactor Systems Branch lead reviewer for CPSES.
In this capacity, I was responsible for the safety review of the CPSES Final Safety Analysis Report ("FSAR") Sections 4.6, 5.2.2, 5.4.7, 6.3 and 15.0, in accordance with the corresponding sections in the Standard Review Plant, NUREG-75/087. Section 6.3 above addresses the design of the Emergency Core Cooling System ("ECCS").
i 9205130201 020507 PDR ADOCK 05000 C
Q.3.
What is the subject matter of your affidavit?
A.3.
I will address the Atomic Safety and Licensing Board (" Licensing Board") questions regarding the eletion of the Boron Injection Tank (" BIT"). The Licensing Board tequested that the Staff pro-vide the following information:
1.
A copy of the " Summary of Meeting on Comanche Peak Design Change and Responses to RSB Questions," by S. B. Burwell dated May 26, 1981.
2.
A description of the system or equipment to be deleted by removal of the BIT.
3.
The status of the deletion of the BIT.
4.
The basis for the deletion of the BIT.
5.
The intended purpose of the system that was being taken out.
Q.4.
Are you familiar with the report entitled, " Summary of Meeting on Comanche Peak Design Change and Responses to RSB Questions,"
by S. B. Burwell dated May 26, 1981?
A.4.
Yes. A copy of that report is attached to this affidavit as.
Q.5.
Please describe the BIT?
A.S.
The BIT is a 900 gallon stainless steel tank filled with 12%
concentration Boric Acid, which is located between the discharge of the centrifugal charging pumps and the injection point into the reactor vessel cold legs. The BIT system is provided with a recirculation loop consisting of a 75 gallon Boron Injection Surge Tank ("BIST"), and two 20-gpm Boron Injection Recirculation
. Pumps ("BIRPs"). This loop maintains the 12 percent boric acid solution in the BIT at a temperature in excess of the solubility limit, and circulates the solution to prevent cold spots and stratification.
In addition the BIT, the BIST and the BIRPs are monitored by instrumentation indicating the system temperature, pressure and flow, and the level in the BIST. Alarms are provided in the control room to alert the operator to any deviations in the system operation. The BIT system w;:ich contains the 12% boric acid solution is isolated on each of the two sides (i.e., the charging pump side and the reactor vessel side) by parallel motor operated valves that are normally closed, and which open on a safety injection signal. All the piping, tanks and pump surfaces adjacent to the high concentration boric acid solution are either heat traced or maintained in heated enclosures.
Q.6.
What system was the BIT part of?
A.6.
The BIT was a component in the high head charging pump system for the Emergency Core Cooling System ("ECCS"). This charging pump system is designed to provide make-up water for core cooling when the reactor vessel (or primary coolant system) pressures remain high.
The system is made up of two centrifugal charging pumps in parallel taking suction through two parallal motor-operated valves from the Refueling Water Storage Tank ("RWST"). The
i i i
RWST contains 450,000 gallons of 2000 parts /million (ppm) j l
boron concentrated water. The two centrifugal charging l
l pumps are designed to provide 150 gpm-each of emergency core cooling water at 2500 psig. That is, the high-head pumping system is designed to provide core cooling at high pressures, l
when lower pressure systems are ineffective. The two centrifugal charging pumps discharge into four 11" lines that inject directly into the four reactor vessel cold legs. is a schematic of the high-head charging system with the BIT. Attachment 3 is a schematic of the systen without the BIT. Attachment 4 is a piping and instrumentation diaaram
("PID") of the high-head charging system with the BIT. Attach-ment 5 is a PID of the high-head charging system without the BIT.
Q.7.
How would the BIT be utilized? What was it's intended purpose?
A.7.
The intended purpose of the BIT was to limit the power increase following a steam line break event.
The BIT would be utilized any time the ECCS is activated by a Safety Injection Signal (" SIS"). The SIS starts the charging
i i
!, I pumps and causes the four BIT isolation valves to open, thus allowing the emergency core cooling water discharged by the l
j centrifugal charging pumps to sweep the BIT inventory into the reactor vessel cold legs, and then into the reactor vessel.
j l
l The following is a scenario that delineates the operation of i
i the BIT. For a steam line break, the excessive RCS cooling leads to the primary coolant shrinkage. As a result the k
I pressure in the RCS drops to the point that SIS is initiated i
4 r
l and the BIT boron-concentrated inventory is introduced into i
j the RCS. The RCS cooling that caused the RCS pressure drop,
)
also adds positive reactivity to the core effecting a power increase.
Since the boron-concentration inventory of the BIT i
i adds negative reactivity to the RCS, the net effect of the BIT l
l 1s to limit the power increase following an excessive cooling l
accident.
I i
Q.8.
What is the basis for the BIT deletion?
j A.8.
The Applicants have submitted for Staff review an ECCS design l
l change deleting the BIT. Thesupportingevaluationsubmitteh by the Applicants for the BIT removal while the BIT provided additional shutdown margins in the form of negative reactivity through the 12% boric acid solution, this additional shutdown
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margin was taken credit for only in the steam line break analysis. With conservative system assumptions in the steam line break analysis the removal of the BIT does not change the ar lysis results significantly, nor does it violate any s
safety criteria. The Applicants calculated, for a large steam j
line break accident without the BIT, a Departure from Nuclear l
Boiling Ratio "(DNBR)" of about 2.5, which is well above the safety limit of 1.3.
Consequently, no fuel failure will result from l
a steam line break with no BIT, according to the Applicants' calculations. However, the Applicants' FSAR conservatively assumed 1% failed fuel initially and an additional 5% fuel l
failing after the steam line break accident for the purpose of maximizing offsite dose calculations. Those doses were l
calculated to be a small fraction of the 10 CFR 100 limits.
1 Since the BIT deletion did not cause a violation of any safety limits, the Staff has found that deletion to be acceptable.
Q.9.
What components or equipment are deleted when the BIT is deleted from the CPSES ECCS?
A.9.
Components between the two upstream isolation valves and the two dnwnstream isolation valves, except for the 4" pipe O
connecting the two sets of valves, are to be removed. The i
upstream isolation valves are either locked-open or removed.
l The heat tracing is also removed.
See Attachments 3 and 5 for l
schematics of the high-head changing system without the BIT.
1 Q.10.
What is the status of the Staff's review and evaluation of the l
BIT removal?
t j
A.10.
The Staff has completed its review of the BIT removal and has i
found it acceptable for CPSES, as stated on page 4-21 of the Safety Evaluation Report ("SER") for CPSES:
The applicant has proposed deleting the concentrated I
boron injection tank (BIT). Although this is a change from previously approved Westinghouse designs, the applicant has shown that removal of the BIT will not result in any unacceptable transient or accident j
analysis results.
l The staff concludes that the designs of the reactivity control systems conform to all applicable regulations j
and are acceptable.
The Staff concluded the CPSES has an acceptable ECCS design as shown in the Safety Evaluation Report ("SER") and Supplement 1 l
l for CPSES. The Staff had previously concluded that removal of j
l tne BIT was also acceptable for the Turkey Point plant, units 3 and 4.
Q.11.
Will the BIT deletion affect the ability of the CPSES to operate safely?
o A.11.
No. As explained in my Answer to Question 8, removal of the BIT will not affect the ability of CPSES to operate Safety.
The Staff's safety evaluation of the CPSES ECCS is shown in the SER and Supplement 1 for CPSES.
Q.12.
What is the construction status of deletion of the BIT?
A.12.
The Applicants advise that it is bypassing the BIT in Unit 1, leaving the BIT unconnected and in place. The BIT will not be installed in Unit 2.
The above statements and opinions are true and correct to the best of my personal knowledge and belief.
1 i
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l i
I SAttMY DIAB i
Subscribed and sworn to me j
this day of May, 1982.
Notary Public liy Commission expires:
I l
~
ST ATEMENT OF PROFESSIONAL QUALIFICATIONS SAMMY S. DIAB
=
4 I an a Nuclear Engineer in the Reactor Systems Branch of the U.S.
Nuclear Regulatory Commission (NRC).
In this position, I am responsible l
t for the technical analysis and evaluation of reactor systems, accidents l
J f
and transients, and applications for nuclear reactor operating licenses.
I have been in my current position since 1980.
[
From 1978 to 1980, I was a reactor systems reviewer in the Reactor Safety Branch, Division of Operating Reactors of the NRC.
In that position i
my responsibilities included:
systems analyses, accident and transient i
analyses, and reload application reviews.
From 1977 to 1978, I was a Nuclear Engineer in the Engineering i
In that Methodology Standards Branch, Office of Standards of the NRC.
position I was responsible for updating and revising the standard review I
I developed Regulatory Guide 1.139, " Residual Heat Removal Guidance",
plan.
and Regulatory Guide 1.141, " Containment Isolation Provisions for Fluid Systems".
From 1973 to 1977, I"was a Nuclear. Engineer with Bechtel Power Corporation, Gaithersburg Power Division, Maryland.
I was responsible i
t for reactor containment pressure and temperature analyses following a spectrum of high energy line breaks, jet impingement calculations, and i
subcompartmen't~ transient behavior.
I developed and used computer codes.
I also modified existing computer codes.
l-S e
e e9 mq_.__
_cpq_ N=_e_
_2 From 1971 to 1973, I was a research assistant with the Nuclear Engineering Department of the Pennsylvania State University.
In 1974 I was awarded a M.S. degree in Nuclear Engineering from Pennsylvania i
State University.
t from 1967 to 1971, I was a researcher with the Egyptian Atomic Energy Establishment.
I received my B.S. degree in Nuclear Engineering t
i from the University of Alexandria, Egypt.
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o, UNITED STATES
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NUCLEAR REGULATORY COMMISSION
/
E WASHINGTON, D. C. 20555
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MAY 2 61981 Docket Nos. 50-445 and 50-466 APPLICANT:
Texas Utilities Generating Company FACILITY:
Comanche Peak Steam Electric Station Units 1 and 2
SUBJECT:
SUMMARY
OF MEETING ON COMANCHE PFAK DESIGN CHANGE AND RESPONSES TO RSB QUESTIONS Summary A meeting was held at NRC Headquarters, 7920 Norfolk Avenue, Bethesda, Maryland on Tuesday, March 10, 1981. The purpose of the meeting was to permit Westinghouse to brief the NRC staff on a proposed design change to eliminate the boron injection tank from the emergency core cooling system, and to discuss information submitted in FSAR Amendments 14 and 15 on boron dilution transients and responses to Reactor Systems Branch questiov.
Attendance at the meeting is listed in Enclosure 1.
Meeting Details The applicant opened the meeting by advising that FSAR Amendment 16 to be submitted March 31, 1981 will show the proposed deletion of the boron injection tank from the emergency core cooling system.
Comanche Peak will be the first Westinghouse reactor to be licensed without the boron injection tank.
However the concept has been reviewed on RESAR 414, Byron-Braidwood and South Texas.
G. Narasimhan, Westinghouse, gave the first part of the presentation on the deletion of the boron injection tank using the slides identified as Boron Injection Tank Removal, Enclosure 2.
This presentation reviewed the impact of deleting the boron injection tank on Westinghouse power reactors in general.
The steam line break is the only design basis accident (Chapter 15) for which credit is taken for the boron injection tank. The presentation concentrated on the response of the nuclear system to a steam line break with and without the boron injection tank.
F. Thomson, Westinghouse, gave the second part of the presentation using the slides identified s Comanche Peak Boron Injection Tank, Enclosure 3.
This presentation revieaed the impact of deleting the boron injection tank on Comanche Peak.
Again, the presentation concentrated on the steam line break accident with and without the boron injection tank.
Comparison of the transient response shows that core power becomes essentially the same with and without the boron injection tank in about 300 seconds.
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The applicant described the removal of the boron injection tank as an operational advantage, using FSAR Figure 6.3-1 Sheet 1 to illustrate the auxiliary equipment needed to support the boron injection tank.
The second part of the meeting started with a presentation by V. Koch, Westing-house, on the design modification responding to the boron dilution transients using the slides identified as Inadvertent Boron Dilution, Enclosure 4.
The presentation tracked the earlier Westinghouse presentation given at the meeting of October 15, 1980. The staff raised questions about the administrative controls on valves within the chemical and volume control system which were resolved in the discussion.
The staff raised questions on how the electrical controls of these valves tie into the instrumentation and control drawings and the impact of control failures on system performance. The applicant will provide additional information on these matters.
The last part of the meeting consisted of a discussion of the applicants responses to Reactor Systems Branch questions.
Of the 51 questions issued in our letter of November 20, 1980, the applicant has responded to 40 questions in FSAR Amendments 14 and 15.
The applicant expects to respond to 9 more in FSAR Amendment 16 at the end of March, and the remaining 2 in April. We discussed the applicants responses to several of the questions.
In most cases the discussion resolved the matter.
The applicant agreed to revise the response to Question 212.96(f) to clarify the consideration on decay and plate-out.
The applicant also advised it is revising the low pressure overpressure protection system.
5 lb /'
es S. B. Burwell Licensing Project Manager Licensing Branch No. 2 Division of Licensing l
Enclosure:
l As stated cc w/ enclosures:
See next page
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i Ere ;+4~= "'ea Dresident and l
General 7.a..awee Te xs s '. ' ' -' : 0:r.:r:tir.; C: ;any i
2001 Brya.. T....
Dallas, Tu-;;
75201
- .'M:h:?:: !. o.cyne!ds, Esq.
Mr. Richard i
- nis ta Debavnica t Liberman Citizens for Fair Utility Regulation 1200 Seventeenth Street 1668-B Carter Drive Washinoton. D. C.
20036 Arlington, Texas 76G10 Spencer C. Relyea, Esc.
Resident Inspector /Lomanche Peak Worsham, Forsythe & Sampels Nuclear Power station 2001 Bryan Tower c/o U. S. huciese kegulatory Commission Dallas, Texas 75201 P. O. Box 38 i
Glen Rose, Texas 76043 l
Mr. Homer C. Schmidt 1
Manager - Nuclear Services
}
Texas Utilities Services, Inc.
l 2001 Bryan Tower Dallas, Texas 75I;;
Mr. H. R. Rock Gibbs and Hill, Inc.
393 Seventh Avenue New York, New York 10001 Mr. A. T. Parker Westinghouse Electric Corporation P. O. Box 355 Pittsburgh, Pennsylvania 15230 David J. Preister Assistant Attorney Ger.aral Environmental Protection Division P. O. Box 12548, Capitol Station Austin, Texas 78711 Mrs. Juanita Ellis, President
)
Citizens Association for Sound Energy 1426 South Polk l
Dallas, Texas 75224 l
Geoffrey M. Gay, Esq.
1 West Texas Legal Services 100 Main Street (Lawyers Bldg.)
j j
Fort Worth, Texas 76102 e
l
i ENCLOSURE 1 ATTENDANCE LIST FOR MARCH 10, 1981 MEETING COMANCHE PEAK STEAM ELECTRIC STATION, UNITS 1 AND 2 Texas Utilities Services, Inc.
NRC Staff i
F. Madden S. Burwell J. Shrewsberry Hulbert Li
- 1. Dunning l
l Westinghouse C. Rossi J. Guttman D. Popp S. Diab G. Narasimhan D. Shum
(
R. Steifler Chang I,i l
F. Thomson V. Koch EG&G I
M. Torcaso S. Bruske Gibbs & Hill i
l S. Kumar i
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_ ENCLOSURE 2 I
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BORON INJECTION TANK REMOVAL O
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BORON INJECTION TANK LOCATED IN THE ECCS CONTAINS 12% BORIC ACID SOLUTION i
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CONSERVATIVE DESIGN FEATURE FROM PAST YEARS SIZING OF THE SHUTDOWN MARGIN BIT AT 12% TO BE COMPATIBLE WITH THEN CURRENT CVCS SYSTEM LOCATION IN ECCS SYSTEM NATURAL AND APPROPRIATE PROVIDES BORIC ACID SOLUTION FOR A STEAMLINE BREAK EVENT a
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STEAMLINE BREAK EVENT SECONDARY SYSTEM PIPE BREAKS OR INADVERTANT OPENING OF SAFETY OR RELIEF VALVE STEAM RELEASE AND INCREASED HEAT EXTRACTION FROM RCS REDUCTION IN COOLANT TEMPERATURE AND PRESSURE t
NEGATIVE MODERATOR TEMPERATURE COEFFICIENT CAUSES POSITIVE REACTIVITY INSERTION REDUCTION IN CORE SHUTDOWN MARGIN POTENTIAL FOR CORE TO SECOME CRITICAL AND RETURN TO POWER 2-5
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STEAMLINE BREAK CONCERNS FUEL AND CORE INTEGRITY CONTAINMENT. INTEGRITY 10CFR20 #1D 10CFR100 DOSE LIMITS 2-15
ACCEPTANCE CRITERIA FOR STEAMLINE BREAK LICENSING CRITERIA MEET DOSE LIMITS FOR SPECTRUM OF BREAKS CONTAINMENT PRESSURE LIMIT WESTINGHOUSE INTERNAL CRITERIA SHOW DNBR > 1.3 FOR SPECTRUM OF BREAKS INSURES DOSE LIMITS ARE MET PERFORM CONTAINMENT ANALYSIS SHOW FOR THE CONDITION II BREAK, I.E., SECONDARY SAFETY VALVE, THE SOM SIZING BASIS IS MET WITHOUi A BIT THE CONDITION II BREAK MAY RETURN TO A LOW POWER LEVEL 2-16
O CONCLUSIONS BIT IS A CONSERVATIVE DESIGN FEATURE STEAMBREAK. LICENSING CRITERIA CAN BE MET WITHOUT A BIT 2-17
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EllCLOSURE 4 INADVERTENi' BORON DillITION CG%NDE PEAK FSAR SECTION 15A.6:
RSS M CVCS OPEPATION M FAIUJRE MEES SYSTEM HARDWARE KDIFICATIONS TO AIERESS REG. GUIDE 1.70 REV. 2 REDJIREPENTS ANALYSIS ASSLPPTIONS AND ESULTS t
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DESCRIPTION OF ACCIDENT INITIATIm 4
RWS/CVCS FAIWE IGES:
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ANALYSIS ASST &TIONS ANALYSIS COES ALL PGES OF OPEFATION AS DEFIED BY STS CONSERVATIVE ASSLfMIONS PADE FOR BORON CONCBERATIONS, BORON WORTHS, SRJTDOWN MARGINS, AND RCS VOLINS DILMION RATE LIMITED BY ADMINISTRATIVE CONIFDLS WHILE RCS IS DEPESSURIZED.
DILUTION PATE LIMITED BY CHARGING Pl W CAPACITY WITH RCS AT PRESSUE l
4-3
ERJEl.ING DIUJTION PECLUDED BY AMINISTFATIE CONTN.'.S WICH ISOLATE RCS FM UNIOPATED MNEP MIER COLD SHJIID#1 MINIM SHJTDOWN MARGIN OF 1% WK l0ST LIMITING MEN RCS ETER LEVEL DPAIED TO MID-N0ZZLE WHILE ONRHR MAEl)P FLOW ESTRICIED TO 150 GPM FOR THIS CASE HOT SHJTDOWN AND HDT STANDBY MINIM SHJTDOWN MARGIN OF 1.6% WK i
10ST LIMITING WHILE ON OE TRAIN OF RHR WITH RCS FILLED #0 ETED MAXIM PERISSIBlf MAEUP FLOW PATE IS IGS GPM 00 ESTRICTIONS ECESSARY FOR CmANCE PEN 0 4-4
STARRP MINIM SUIDOWN MGIN OF 1.6% 4'/K ALL 4 RP'S OPERATIE DIUJTION FLOW LIMITED BY CAPACITY OF TWO OMRGIE PLPPS FULL RCS Rite MIWS PESSURIZER ASSLPE AVAILABLE FOR DILUTION POWEROPERATION DIUJTION FLOW LIMITED TO MAXIM LETDOWN FLDW WWN IN AITmATIC PESSURIZER LEVEL GETR)L DILUTION FLOW LIMITED BY MAXIM 0F 2 0%RGING PTPS IN OPERATION WEN IN MAWAL PESSURIZER LEVEL CONTR)L 00ST LIMITING) 4-5
RESULTS T ANALYSIS COLD S E DOWN, HDT SR HDOWN, W T ST# E Y SOURCE RANGE NIS, BY WAY OF A MICR0 PROCESSOR, It01 CATES DOUBLING OF EUBON FLJJX DUE TO DIllHION ALARi ALERTS TE OPERATOR TO A FUJX DOUBLING CVCS AllTWATICALLY REALIGED TO SWITCH SUCTION FRm VOLLE CONTROL TANK TO RWST (CLOSE VALVES 1-LCV-112B #0 C', OPEN VALVES 1-LCV-112D #0 E IN TE CVCS)
THIS AlfimATIC ACTION MINIMIZES THE APPROACH 10 CRITICALITY
- 0 EGAINS LOST SMDOWN MARGIN l
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4-6
STARTUP OPERATOR INTENTIONALLY DILNES IN THIS tGE 10 GO TD POWER IN TlE EVENT OF AN INAINERTEhT DIlllTION WILE ESCALATIE IN POWER, lliE PLANT TRIPS ON TE LOW SETPOINT OF POWER RANGE HIGHfBJTRONFLlJX OPERATOR HAS > 15 MINJTES FR31 TRIP TO MAMJALLY TERMINATE DIl]JTION AND INITIATE B0 RATION BEFORE CORETE LOSS OF SHJTDOWN mRGIN 4-7
POWER MANJAL E COGOL:
TE EACTIVITY EXCURSION ESULTS IN EACIOR 1 RIP ON DEIBPEFATUE N-16 OPERATOR HAS > 15 MINJTES FM TRIP TO TERMINATE DILUTION ANDINITIATEB0 RATION THIS TRANSIENT BOUM)ED BY UNCONTROLLED E WITHDRAWAL AT POWER TRANSIENT AUTCMATIC RCD COGOL:
OPEFATOR AlfMD TO DILUTION BY E INSERTION LIMIT ALARMS
> 15 MINJTES IS AVAILABLE FM ALARMS FOR OPEFATOR 10 TEFMINATE DILUTION #0 INITIATE B0 RATION l
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CONCUJSIOfE IN ALL MTES OF OPERATION, AN INADERIENT DIURION IS PECUJDED, AlHO-MATICAU_Y TERMINATED, OR TEfMINATED BY OPERATOR ACTION TE EACTOR IS BR00EHT TO A STABLE C0f0lTION WITH TE LOST SMDOWN MARGIN RECAIED, O
4-9
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