ML20125D151
| ML20125D151 | |
| Person / Time | |
|---|---|
| Site: | 05000130 |
| Issue date: | 03/12/1964 |
| From: | Lowenstein R US ATOMIC ENERGY COMMISSION (AEC) |
| To: | |
| Shared Package | |
| ML20125A538 | List:
|
| References | |
| FOIA-85-54 NUDOCS 8506120262 | |
| Download: ML20125D151 (86) | |
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'e UNITED STATES
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ATOMIC ENERGY COMMISSION
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% b} # NORTHERN STATES POWER COMPANY DOCKET NO. 50-130 i )
i PROVISIONAL OPERATING LICENSE License No. DPR-ll
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- 1. This provisional operating license applies to the controlled recirculation boiling water reactor owned by the Northern States Power Company (hereinafter referred to as " Northern States") . The reactor which is part of the Path-finder Atomic Power Plant is located approximately five and one-half miles northeast of Sioux Falls, South Dakota. The reactor is described in the licensee's application for operating license dated March 30, 1959, and
==aad= ants thereto dated July 6, 1959. August 7, 1959, November 5, 1959, November 20, 1959, December 18, 1959, August 24, 1960, November 7, 1960, ,
January 17, 1961. May 22, 1962, June 12, 1962, February 22, 1963. April 24, ,
1963. May 15,1963, May 29,1963, June 11,1963, June 18,1963, August 14, 1963, August 28, 1963, October 21, 1963, October 24, 1963, October 29, 1963, October 30,,1963, December- 18,.1963,'and February 6, 1964, (hereinafter collectively. referred to as "the application"). ,
- 2. Subject to the coaditions and requirements incorporated herein, including the Technical Specifications hereto, the Commission hereby licenses Northern States:
l A. Pursuant to Section 104(b) of the Act and 10 CFR 50, to possess, use ,
l and operate the reactor as a utilization facility. :
B. Pursuant to the Act and 10 CFR 70, to receive, possess and use in operation of the reactor at any one time:
(1) 800 kilograms of contained uranium-235; i
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(2) 128 grams of plutonium encapsulated as two 1-curie and one I 6-curie plutonium-beryllium neutron sources.
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C. Pursuant to the Act and 10 CFR 30, to receive, possess and use in operation of the reactor at any time.
(1) 10,000 curies of antimony-124 as an antimony-beryllium neutron source; (2) Three sealed sources of cobalt-60 not to exceed 100 milli-curies each for calibration of film badges and instruments and for the testing of radiation monitors and measurement of liquid levels in tanks and p.ipes; B506120262 850214 PDR FOIA PAYB5-54 PDR k ,
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'(3) 300 microcuries of cesium-137 to be used as a laboratory standard- ,
(4) 50 microcuries of iron-59 to be used as a laboratory standa-d; (5) 100 zierocuries of strontium-90 to be used as a laboratory standard; (6) 300 microcuries of cobalt-60 to be used as a laboratory
> stanAmed; (7) 0.002 microcuries of americium-241 for calibration of instruments; (8) 50 millicuries of krypton-85 for calibration of gaseous activity monitors; (9) 300 microcuries of chromium-51 in a solution of Crcl3 to be used as a laboratory standard.
D. Pursuant to the Act and 10 CFR 30, to possess, but not to separate such byproduct material as may be produced by operation of the reactor.
- 3. This license shall be deemed to contain and be subject to the conditions-specified in Section 30.32 of Part 30, Sections 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70, Title 10, Chapter 1 CF1, and to be subject to all applicable provisions of gha Act, and to the rules, ,
regulations and orders of the Commission, now or hereafter in effect, and to the additional conditions specified below: l
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A. Northern States shall not operate the reactor at power levels in i excess of oce (1) megawatt char ==1 i B. Northern States shall not install the proposed nuclear - +,-ter l' fuel in the reactor without prior written authorisation by the Commissioc. .
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C. Teh=4 cal Specifications The Technical Specifications coarmined in Ap"4= "A" hereto are hereby incorporated into this license. Except as otherwise permitted by the Act and the rules, regulations, and orders of.the Commission, Northern States shall operate the reactor in accordance with the Technical Specifications. No changes shall be made in the Technical Specifications unless authorized by the Consission as provided in
- 10 CFE 50.59.
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- Records In addition to those otherwise required under this license and i applicable regulations, Northern States shall keep the following records:
i (1) Reactor operating records, including power levels and I periods of operation at each power level; i
s (2) Records showing radioactivity discharges into the air or water beyond the effective control of Northern States as f measured at or prior to the point of such release or ;
discharge; .
I (3) Records of radioactivity levels at both on-site and off-site !
monitoring stations; (4) Records of emergency shutdowns and inadvertent scrans including the reasons therefor; (5) Records of safety system component tests and measurements performed pursuant to the Technical Specifications; (6) Records of maintenance operations involving substitution l or replacement of reactor equipment or components; (7) Records of all facility tests and measurements performed.
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E. Reports In addition to reports otherwise required under this license and applicable regulations:
(1) Northern States shall make an innadiate report in writing to the Commission of any indication or occurrence of a possible unsafe condition relating to the operation of the reactor, including, without implied limitation:
(a) Any substantial variance disclosed by operation of the reactor from the performance specifications set forth in the Hazards Summary Report; (b) Any accidental release of radioactivity, whether or not
.resulting. in. property damaget or. personal injury or >
exposure above permissible limits.
i (2) Within 60 days after (a) completion of initial core loading !
and associated critical testing and (b) completion of Phase II of the Power Operation Test Program Northern States shall l submit a written report to the Commission of the results pertinent to safety of the tests and operations conducted, !
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including a description of changes made in the facility '
j design, performance characteristics and operating procedures. -
(3) Within 30 days after the completion of six months of operation ,
of the reactor (calculated from 'the date of completion of Phase d' II of the Power Operation Test Program), and at the and of each six-month period thereafter Northern States shall submit a written report to the Commission which summarizes the following:
) (a) Total number of hours of operation and total energy I generated by the reactor; l (b) Number of shutdowns of the reactor with a brief i
- explanation of the cause of each shutdown; '
l (c) Operating experience including levels of radioactivity in principal systems; routine releases, discharges, and shipments of radioactive materials; a description of tests performed in the reactor; and the results of any test analyses completed during the period in the reactor including results of tests required by the Technical Specifications; a saunary of experiments conducted; number of malfunctions in the control and safety systems with brief explanations of each; and a discussion of data obtained relating to superheater operation; (d) Principal maintenance performed and replacements made ;
in the reactor and associated systems including a report l l on various tests performed on components of the reactor :
i and associated systems;- l J
(e) A description of the leak tests performed pursuant to ]
the Technical Specifications and the results of such >
tests including a description of any necessary corrective !
measures taken to meet the requirements of the Technical Specifications for assuring the specified containment leak tightness; (f) Significant changes made in operating procedures and in plant organization; -
(g) Radiation levels recorded at both on-site and off-site monitoring stations.
- 4. Pursuant to Section 50.60 of 10 CFR 50, the Commission has allocated to Northern States for use in the operation of the reactor 758.4 kilograms l of uranium-235 contained in uranium at the isotopic ratios specified in j the application. Estimated schedules of special nuclear material trans-fers to Northern States and returns to the Commission are contained in Appendix "B" attached hereto, which amands the allocation contained in 1 -
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Conistruction Permit No. CPPR-8. Transfers by the Conniasion to Northern States in accordance with column (2) in Appendix "B" will be conditioned upon Northern States' return to the Cossaission of material substantially in accordance with coltaan (3) of Appendix *T".
i' 5, this license shall be effective as of the date of issuance and shall i expire eighteen (18) months from said dates,3:nlees. extended, .forcsood -
l cause shown, or upon the geslier issuance of a superseding operating ,
6 license. .
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FGL THF. ATCMIC ENERGY Cated1SSION Original signed by R. Lewenstein.
Director Division of Licensing and Regulation Alteohnents:
- 1. Appendix "A"
- 2. Appendix "B"
. Date of Issuance: MAR 1 % 1964 t
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4DEEDIE "B" E
ptRTERN STATES 70HE C(RIFANY FECTISIONAL OPERA 1ING LICENSE
$ Este-ted seh-d.d e of Trm==fara of Snacial Nuclear Ma'a-tal fr w the *- 4seios to Northern States -
Power Compaar (ESP) and from E8F to the CommiM i
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Date of Treasters free Esturns by Net Yearly Camalative '
Transfer ABC to NSF NSF to ABC Distribution Distribution (Fiscal Year) Eilograms U-235 Kilograms U-235 Eilograms U-235 Eilosrams U-235 '
Cold Hot , l Thru 10/18/63 490.0 - - - 490.0 19648- 80.6 195.2 -
(114.6) 375.4 i 1965 243.8 13.1 41.9 188.8 564.1 ;
i 1966 243.8' 53.2 - 190.6 754.'8 1967. ,
154.0 15.4 219.3 (80.7) 674.1 1968 .' 187.0 M.2 103.8 (51.9) 623.r'
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61.1 135.3 758.4 ., ,
t 1970 . 87.1 17.4 18$.0 (118.6) . 639.8 1,681.8 348.9 4M.1 -
'03rd and 4th guartare NY I M4 only .
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APPENDIX "A'.' .
. NORTHERN STATES POWER COMP.'ANY. .
PA1HFINDER ATOMIC POWER PLANT.
TECHNICAL SPECIFICATIONS .
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APPENDED TO PROVISIONAL OPERATING LICENSE DPR-11,
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. TABLE OF CONTENTS' -
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~1.0 Introduction 1-1 1.1 Scope 1-1 ,
1.2 Definitions 1-1 2.0 Site 2-1 2.1 Location 2-1 2.2 Exclusion and Restricted Areas 2-1 2.3~ Principal Activities -
, 2-1 3.0 Reactor Containment 3-1 3.1 Containment Vessel Design Parameters 3-1
- 3.2 Containment Vessel Dimensions 3-1 3.3 Construction 3-1 3.4 Penetrations 3-1 3.5 Reactor Building Spray System 3-4 3.6 Containment Requirements 3-4 3.7 Containment Testing 3-5 .,
4.0 Reactor and Power Systems Equipment 4-1 4.1 Reactor Vessel 4-1 4.2 Primary Coolant System 4-3 4.3 Safety Relief Valves 4-4 4.4 Reactor Power Operating Cooling '4-5 4.5 Reactor Shutdown Cooling .
4-5 4.5 Reactor Emergency Cooling 4-5 4.7 Operating Requirements 4-6 4.8 Biological Shield 4-6
.,_ 4.9 Power System Equipment and Associated Facilities, 4-7 __ _
5.0 Reactor Core and Controls 5-1 ,
5.1 Boiler Control Rods 5-1 5.2 Superheater Control Rods 5-1 5.3 Poison Shims . 5-1 i 5.4 Core Composition 5-3 5.5 Sources 5-5 i 5.6 Principal Calculated Thermal, Hydraulic, and Nuclear Characteristics 5-5 ,
5.7 Principal ~ Core Operating Limitations 5-8 '
5.8 Control Rod Drives 5-10 5.9 Control Rod System 5-11 5.10 Boron Injection System Design 5-13 .
5.11 Boron Injection System 5-14 '
i-5.12 Reactivity Coefficients ~
5-14 l 5.13 Reactivity Additions During Core Alterations 5-14 i l
5.14 Reactivity Additions During Power Operation 5-15 l
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Page 6.0 Plant Safety and Monitoring Systems 6-1 6.1 Reactor Safety System 6-1 .
6.2 Reactor Safety System Operating Requirements 6-9 6.3 Plant Monitoring Systems 6-17 6.4 Radioactive Waste Disposal Systems 6-21 7.0 Operating Procedures . 7-1 -
7.1 Basic Operating Principles . '
7-1 h 7.2 Procedural . Safeguards 7-2 7.3 Pro-operational Testing 7-3 7.4 Initial Core Loading and Critical Tests 7-4 7.5 Power Operation Test Program 7-6 7.6 Nomal Operation 7-10 7.7 Refueling Operation 7-12 7.8 Maintenance 7-13 8.0 Research and Development Program 8-1
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1.0 INTRODUCTION
1.1 83 1.1.1 These Technical Specifications set forth operating limits and requirements and principal design features which affect safety of the Pathfinder Atomic Power Plant.
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'1.2 Definitions ,
The following terms are defined to clarify the intent of the various
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provisions given within th,ese Technical Specifications.
1.2.1 Power Operation - is any operatics with the reactor vessel closure bolted in place when reactor' criticality is possible. Reactor criticality is to be considered possible if the core is loaded
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'with a quantity of fuel equal to 'or greater'than a critical mass of fuel, power' is 'available to the control rod drive motors and
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more than one rod is ,latclied'to'its, drive.
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1.2.2 Refueling 0peration - is any operation with any of the reactor vessel clos _ures open during.which either, core alterations am being made, or other, operations which. might 'directly or indirectly increase the reactivity of the core -am in progress.
1.2.3 Shutdown - is 'any conditica not covend"by Power Operation or Refueling Operatica except that Cold Shutdown shall be specifi-cally defined as indicated in Section 1.2.4 below. I 1.2.4 Cold Shutdown - is a react'or condition involving either no fuel
- -in the reactor;or.a condition meeting the following requirements:
(a) The control rods 'are fully inserted in the core, an'd their power circuit is locked by means of the key switch in the off position to prevent withdrawal. The key to the switch must be in the possessica of the Shift Supervisor or higher plant management.
(b) The reactor coolant system is at atmospherie.psessure.
1.2.5 Seram - is any automatic or manual action which de-energizes the magnetic clutches on the boiler control ro4 drives and causes run-in of the drives for the boiler and superheater control rods until the associated rod has reached its bottom limit.
1.2.6 Main Steam Isolatica Scram - is a reactor shutdown as in 1.2 5 above with closure of the' main steam isolation valve and the main steam isolation bypass valve.
1.2 7 Runback - is any automatic or manual action which drives in all i boiler control rods until correction of the condition which initiated the action. -
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1.2.8 Imaksas Rate - is defined as the percent of the emtained stacephere (weight basis) which esespes per der (23 hours2.662037e-4 days <br />0.00639 hours <br />3.80291e-5 weeks <br />8.7515e-6 months <br />) under4 the denned pressure ecaditions through any leaks in the contain-ammt building and its compements including any extensica of the contat====t boundary and all isolation valves and their associated Piping.
2.0 g 2.1 Loestion 6 [
The Pathfinder reactor plant shall be loested near the Big 81ous River, l approximately 5-1/2 miles northeast of the center of the city of l Bioux Palls, and 2-1/2 miles west of the town of Branden in Minnehaha j l County , South Dakata. i l
2.2 Exclusion and Es'stricMd' Areas i
The distance from the centerline of the reactor building to the i boundary of the exclusion area shall be'at least 2250 feet. N distance ! i from the centerline of the zwaetor building to the boundary of the
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restricted aswa shall be at least 135 feet. The restricted area which includes the cooling tower, switch yard, and other equipment shall be
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enclosed by a fence.
t 2.3 Principal Activities h principal activity carried a within the exclusion area shall be the operation of the -reactor and associated power generating equipment.-
Other activities which may be' carried m at the site shall be centrolled by Northern States- Power Company and may include maintenance of i buildings, roads, grounds, and equipment; transmission and distribution
-of- power; and operation of an Atomic Information Center.
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s 300 REACTOR 00NTAINMErf Rosator containnsat shall consist of an externally insulated cylindrical steel vessel, hominafter reformd to as the containment, vessel or enclosum. The insulation shall be appresimately one inch thick with a waterproof exterior coating. The integrity of the insulation shall be sufficient to maintain the reactor building metal temperatum above 55'F during power operation conditions.
3.1. The reactor contaf n=nt shall enclose the reactor, the recirculation shielding, and other components arranged as shown in Figures .3.3 loops,h and 3. of ACEP 5905 dated January 15,1962. Electrical . circuits and hydreolic and pneunatic lines' used for the purpose, of controlling the reactor or cetrolling or actuating _ safety and emergency systems shall be separated from all high pmssure piping by substantial structural fleatums so that rupture,of a high pmssure pipe would not impair the function of the control systems. ,
3.2 Containment vessel-Desian Paraineters 'shall be as follows:
Internal pressure, psig 78 External pressure, psig 7.3 Temperature (Coincident with design internal pressure),'F 3k2 Minimum building temperature, 'F 10 Wind load, psf 30 Snow load, psf 35 Earthquake factor, C (UBC code) 0.05 Permissible air leakage rate at 78 psig at ambient temperature, percent per day of contained atmosphere (including penetrations) 0.2 Diameter, feet 50 Height, feet 120.5 Approximate free volume, cubic feet 145,000 3.3 construction ,
The principal material of constructim shall be SA-212 Grade B, l i
firebox quality steel produced to SA-300 specifications. Design, i construction, and testing shall be in accordance with ASMC thfired !
Pmssum Vessel Code, Sectica VIII, as modified by the applicable l 6 nuclear code as of 1960 for the conditions specified in Section 3.2. )
i The parent metal and weld metal shall have an NDE temperature of less than 25'F.
- 3.h Penetrations Penetraticas through the reactor building shall conform to.the following i specifications:
3.b.1 Electrical Cables l Penetrations shall be prcwided for a maximum of eighty-five I electrical cables. About seventy-two of these penetrations i shall be used initially and the remainder shall be spans for 3-1 I
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future use. The electrical penetraties 'shall be fianesionally,. as j shown in Allis-Chalmers Drawing No. A-87219-D dated January k 1963.~
3.4.2 Slates Pinina-
. - . . Penetrations shall be provided for a maximaa of 19 system piping lines. The piping and instinnent line penetrations shall be 1selded sleeves varying in siameter from 1-1/4 to 36 inches, with the lines rigidly;velded ,to the sleeves except where temperature
, or pipe movessats preclude' rigid connections, in which case h bellows seals shall be provide,4., I All piping' penetsstions shall
}- be designed with adequate. reisiforcement and support to prevent j . esponsion er reaction from a ruptured pipe from causing damage i to the penetration er.4he'_ containment vessel. The stresses ,
- resulting from sach reactions shall not exceed the allowable (
l stresses for the materials .used in the penetration and contain-. j ment building. Each . valve in the inlet and outlet of the '
ventilatica system shall be provided,with' individual actuators.
. Valves which sene as isolatim valves shall be located with
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respect to the containment vessel as shown actionstically in i
Figure 2.1 of ACIP 5905 dated January ~15,1962. 'All such valves which may be open during operaties requiring contain-
- ment integrity shall be designed to permit remote manual closing of the valves from the cetrol room. The following ,
, isolation valves lsheil'$ designed to be actuated automatically l
- and manually from the control. room. The mode of operation of these isolation valves 'shall. be as follows
i Penet'ratiois ,f- Mode of Operation
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! a. Beactor Waiteiir Purification air to open, spring
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to close
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. b.--Becirculation Pumps Seal . -~ r air to open r spring-~ --- - -- '
Imak-off' to close
- c. Shield Pool Cooling Outlet air to open, spring
- . to close
- d. Main Steam !
Main Steam Isolation Valve electric to open and close -
d'* Main Steest Isolation Bypass air to open, spring Valve - to close
- e. Heating Steam dir to open, spring to close !
- f. Safety Valve Discharge air to open, spring l
- to close '
i j g. Sump Pump Discharge air is open, spring !
to close. !
j . h. Safety Valve Drain -
air to open, spring i to close
- i. Ventilation Inlet and Outlet (see Sec. 3.k.6) i
- j. Instrument and Service Air air to open, spring .
to close .
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i 3.k.3 Personnel Ace.as There shall be a moraal airlock (sise 7 ft high x 3 ft vide) i and an emergency airlock (size 30" diameter) for personnel ;
access and egress. Each airlock shall withstand a 78 psig building pressure. Each airlock door shall open inward toward the reactor building. Airlock doors shall be mechanically l
interlocked so that aly cae door can be opened at a time. The (
emergamey airlock shall be provided with a means for closing !
the inside door from the outside of the reactor building.
k Shafts or other novable mechanical devices penetrating the :
airlocks shall pass through packed fittings which provide a '
seal between the inside and the outside of the reactor building.
3.k.k Equipment Transfer The equipment transfer door shall be eleven feet in diameter and secured by a bolted 0-ring gasketed joint arranged so that pressure inside the contaf n=ent building tends to compass the gasket.
3.k.5 Fiael Transfer h fuel transfer tube shall conneet the shield pool to the fuel storage pool. A gate valve shall be used to close the transfer tube.
3.k.6 Isolation Yalve Operators y
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All system piping isolation valves except the main steam isolation bypass valve shall be designed to close faster than the main steam isolation valve which shall close in less than 19 seconds. h main steam isolation bypass valve shall have a controlled closure time of about 2 minutes; backup control devices shall initiate instantaneous closure within 2-1/2 minutes.
The main stena isolatim valve shall have an electric powered valve operator. The electrical power shall be available from the emergency power system.
Both of the main steam line valves shall close without operator attention upon Main Steam Isolation Scram signals as described '
in Section 6.1.k. ,
hre shall be two valves in each ventilation duct. One valve ;
in each duet shall be nitrogen operated in both dimetions. I h nitrogen shall be supplied by two 1000 cubic inch I accumulators at a minima pressurw of 400 psig. h second ;
valve in each duct shall be air-tc> open and spring-to-close.
N ventilation valves shall close in less than 15 seccada. -
h inlet and outlet ventilation valves shall close autcssatically I upon Reactor Building Isolation signals as described in Section '
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1 3 5 Practor Building Spray System i
I The reactor building spray system shall be designed to spray water in the reactor building at a minimum rate of kT gym with 78 psig pressure
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in the nactor building. The" system shall be designed with the !
following features:
- a. Number of Sprar Nozzles 1
- b. Nossle Pressure, psia 118
- c. System Actuation Automatic, Manual Backup '
- 4. Signal Used to Actuate - - High Catain=nt f Yessel Pressure
! e. Signal Trip Setting Not mon than 6 psi abore atmospheric
- f. Water Supply Circulating Water System The automatic actuatica system shall include a time delar device allowing a time delay between actuatico signal and opening of the valve of not more than five minutee. . Manual acetrols shall be provided which allow opening and closing of the supply valve at any time.
3.6 contaf n=nt Integrity Jieauirements i
3.6.1 containment integrity provisions shall include the following:
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l (a) Maintenance of the nactor building so that the leakage rate shall not exceed 0.2 percent per day of the contained atmosphere, at an internal pressure of 78 psig at ambient temperature. -
(b) Sealing of all building access and equipment transfer ports including: the equipment. door and fuel transfer valve.
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(c) Maintaining all systems for automatically closing contain-ment building penetrations in operating conditica.
(d) Maintaining temperature of the steel shell of the
! containment above 55'F."
l l (e) Preventing all access to and egress from the containment l' building except through airlocks in which the door interlocks are operable. l (f) Maintaining all emergency power supplies, monitors, and automatic emergency equipment associated with the contain-ment building and building sprar system in operating L
condition.
3.6.2 Whenever primary sys --a pressure exceeds 250 peig and fuel is ::
in the reactor, whenever fuel is in the reactor and any control rod is withdrawn, or whenever any component is being handled in proximity to irradiated fuel within the reactor building, the i ccatainment integrity provisims specified in this section shall be in effect; except that whenever (1) irradiated fuel ,
i is being handled with the primary system at atmospheric '
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pressere and with control rods disemnected from control rod drive mechanisms the gate valve in the fuel transfer tube may be opened ame (2) whenever the feel involwd la any type of opera,- I ties defined by this sectiam shall have been irradiated to less than 100 megawatt days per metric ton _ of uranium the provisions
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of subparagraph 3.6.1(4) shall. net be required.
i 37 containment Testina ;
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i 3.T.1 Proenerational can64=At Initial Testinz l (a) The reactor buildkas sinsll be initially tested by a posissatic pressun test at 125% of the design pressure
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(acminal 97-1/2 pois test pmssuio}. ,
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(b) A leak-detection' test shall'be eenducted at a pressure !
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not less than 5 peig for all reactor building welds, all 1 shell penetrations, gasketed joints, and isolation valves !
l using t' n e soap-bubble method, or the Halide leak detector. l l
All discernible leaks a venhed by these test methods !
shall require apair and retest.
(c) The initial integrated leakage rate test shall be conducted i at 1005 ef the design pressure. i (d) The nazimum allowable leakage rate of the Beactor Building i et desia pressure shall not exceed 0.2% of the contained air in twenty-four hours. The actual measured leakage rate derived from the test coedneted with air shall be corrected for the contat====t conditions postulated by i the-maximum desi p accident. '
! (e) The -accuracy ofdholleakageIrate asasuring system, in-any j
! test, shall be verified by superiaposing a contmlled L leakage rate (measured through a gas flor meter) upon the i existing vessel leakage ratei or by other means of-j equivalent accuracy, and continuing the t4st a sufficient ,
l period of time to measure the composite leakage rate.
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! (f) The tests required by Sections 3.T.l(b) and (e) shall be l conducted after all construction work affecting contain-ment leak tightness is complete and within ene year i peeseding1mitial operations requiring catainment i integrity as defined in Sectica 3.6. '
3.T.2 containaset Periodie Testina (During Operation Below 1 mr(th)
(a) All penetrations [an&7sasketed closuzes shall be subjecte'd to a leak detection test either separately or as a building ;
test, at a pressure not less than 5 Psig every year, using ;
the soap-bubble technique or other methods at equivalent ;
eensitivity.
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~(b)'~ Equipment transfer door shall be subjected to similar leak l
, detection test following each closure prior to plant startup. '
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(c) All airlocks shall'. be subject to leak detection tests every six months.
(d) All isolation valves with direct communication to outside atmosphere, including the valves for the ventilation openings and building vacuum relief, shall be subjected to a leak detection test at a pressure of no less than 5 psig every four months using soap bubble technique or other methods of equivalent sensitivity. The normal mode of valve
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operation shall be employed to close the isolation valves prior to the performance of, the leak detection tests.
All automatic controls and instrumentation associated with these isolation valves shall be tested every three months.
(e) Allcontainmedtisolationvalvesandcheckvalvesconsidered as necessary for containment but not included under 3.7.2(d) and the automatic' valves of the spray system shal1 ~be tested every year or at each major refueling, whichever occurs sooner, to verify the opers bility of the valves.
The automatic controls and its instrumentation associated with these valves shall be tested at approximately quarterly intervals. -
i Defective operation .shall require repair and retests.
All discernible leaks revealed (from tests of (a), (b), (c),
and (d)) subsequent to the preoperational tests shall require repair and retests. !
' ' ~ ~" -
[' (f)' "Integfated' leakage' rate ~ tests oOthe Reactor Building
~
shall be performed at least once every 18 months.
( In order that the integrated leakage rate test be representative of the "as is" condition of containment building, no preliminary preparation of the leak tight condition of the containment building shall be performed .
which would influence the resuts. of the scheduled I integrated leakage rate test. Closure of the isolation valves of the Reactor Building penetrations for the purpose of the tests, shall be effected by the normal means provided for operation of valves.
Leaks detected in the containent boundary, or in the isolation valves directly prior to or during the test which require repairs to enable the integrated leakage rate testing to proceed, may be repaired provided such repairs are reported as part of the record of the leakage rate results.
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The integrated leakage rate test any be conducted at 100% i of design pressure. The corneted leakage rate shall not !
exceed the nazimum allowable _ leakage rate specified under l 3 7.l(d).
- l In the event the naminum allowable leakage rate is exceeded l as determined at any time by the test usults of 3.7.2(e),
l a stest shall be made following repairs of leaks in the containment building boundary.
A proposed schedule and specification for periodie leakage '
rate retesting for operatim above 1 Mw(th) shall be submitted to Division of Licensing sad Regulation for approval prior to operation above 1 Me(th).
(g) The time periods specified in 'this Section 3 7 2 shall be the maxiann interval of. time elapsed between successive tests. These ' tests'shall be performed initially after all construction work affecting the function to be tested is complete and within three maths preceding initial.
operations requiring containment integrity as defined in Section 3.6; provided that tests also specified under Section 3.T.1 may be performed initially at times required by that section.
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4.0 Reactor and Power Systems Equipment h e reactor and power system equipment shall consist of the reactor vessel, the primary coolant recirculating system, the shutdown cooling system, emergency cooling s'ystem, pressure relief system, the primary steam and
- associated power producing equipmedt and the interconnecting piping and i valves. his system shall be arranged as shown in Drawing 43-500-997 in ACNP-5905 dated January 15, 1962. . his section will specify the in-portant mechanical design features and operating limits of variables affecting these systems. -
4.1 Reactor Vessel h e reactor vessel shall be designed, fabricated, installed, and tested in accordance with Section VIII and II of the ASME Pressure vessel Code as of 1955 as modified by Special Code Cases 1270N, 1271N, and 1273N.
4.1.1 Desian Features shall be~as follows: -
Nominal Length, overall, inches 433 Nominal Inside diameter, inches 132 i Nominal Wall thickness, excluding 3 (except head =
cladding, inches {
2 3/8) l Cladding thickness, minimum inches 1/4 (except head = !
~-
1/8) i Design pressure, psig 700 ,
j Design temperature, 'F 500 . > .
Approximate Initial nil ductility 10 '
transition temperature, 'F i 4.1.2 Principal Materials of Construction shall be as follows:
Component Material Specification Vessel shall and Steel ASTM A212 Grade B- i head !
I Flanges and Nozzles Steel ASTM A-105 Gr 11, ASTM A155 Gr EC-70 i Cladding Stainless Types 304L and 309 steel Head Stude and Nuts ,. Steel ASTM A437 l
4.1.3 Reactor Vessel Penetrations !
he penetrations of the reactor vessel and reactor head j shall include only those listed below: ,
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- Penetrations Quantity L.D.(inches) Location
! 1. Nossles to Recirculation Pumps 3 20-1/2 Below Core 2 Nozzles from Recirculation Pumps 3 20-1/2 Below Core
- 3. steam Outlet Nozzle 1 16-1/2 Below Supher 4
._.: Paedw6cer Nozzle' . .
1 9-3/4 Below Core
~5. Boiler Core Instrument Nozzle 1 6-3/8 Above Core
- 6. Lower Liquid Level Nozzle 1 .2-3/8 -Below Core
- 7. Control- Rod Drive Nossles 20 3-1/2 Vessel Head
- 8. gaperheater Core Instrument 1 1-1/2 Vessel Head Nossle -
- 9. Upper Liquid Level Nossle 1 1/2 Vessel Head 4.1.4 Vessel closure I he top head closure shall be a bolted flange with a i 7 foot-7 inch clearance diameter. We seal shall be made j by two corrugated, type 316 stainless steel jacketed, soft j iron core gaskets. The space between gasket ' rings shall be connected to a leak monitoring system. .
4.1.5. vessel support i
n e reactor vessel shall be supported by.five columns 2 fabricated of high strength T-1 steel. D e vessel shall i be held in a centered po'sition by.four keys welded to the I
, vessel wall and engaged in channels fastened to the concrete shield.
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- 4.1.6 vessel Internals '
i
- l The vessel internals shall consist of 412 superheater fuel assen-j blias and 96 boiler: fuel elements, boiler fuel boxes, super- -
! heater structural assembly, control rods, shroud structure, and structures which support and retain fuel elements.
These vessel. components,shall be as shown in Figure;.1.1 of
~
ACNP-5905, dated January 15, 1962. h e core region shall
, be surrounded by a light water reflector region approximately 32 inches thick.
4.1.7 Control Rod Guidance h e. boxes which contain the boiler fuel elements'shall form the guide channels for the cruciform control rods. Wese channels shall be fabricated to provide.0.113 inches clearance between the control rod and the guide ch==nal when the con-rod is centered in the guide channal. -
The superheater control rods shall move in guide tubes.. % ere 'l shall be 0.065 inches between the guide tubs'and the control rod when centered. -
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] 4.1.8 2eactor vessel and steam Line semples l !
There shall initially be at least fiftees capsules Tarme=J l in the vessel Waf =i== specimene of the reme*=r vessel ; ,
asterial. The capsules shall be locatad is semple holders j i
f which are part of the steam eeyarater assemblies. The samples shall be -located at tuo radial df ar==ema from the [ ,
core. The funer position shall provide an accelerated ,1 ;
test cf radiation effects. E i i
$ Three capsules will'be removed during the first year of f equivalent full power eyeration for *===fmaties. The results I l
of these ====*==rione shall dete =4=n the schedule for re- y .
noval of the other samples. { l A test program shall be condected to evaluate adverse effects of neutron irradiation of main steam line material. 5 i-4.1.9 operatinz Esquireme .rs !
l The rate of change of temperature in the reactor vessel l vall, flanges, and sozzles shall not ezceed 200*F per hour )
as dece mined by at least 2 thermocouples in the reacter j ;
vessel wall. Reactor vessel pressurisation in excess of i 201 of normal operating pressure shall not be a11 sued to occur at temperatures below the ==w" - established mil ,
dactility transition temperature plus 60*F. A deta=4==*f as ,
of the shift in the mil doctility trsasition temperature -
shall be made at least once each year. '
4.2 Primary Coolsac Systmas 4.2.1 Recirculation loops t
a The reactor recirculatica system shall consist of three loops, i ea:h comeminis= a vertical mined flow centrif W pamp. The { ,
p = p shaft sealing arrangement shall be designed to prevene !
loss of recirrn1m**= water daring all operating ecoditions. i seal water shall be s-applied by either or both of two positive displacement gland seal booster y mps. >
i 4.2.1.1 Piping i !
The recirculation piping shall be designed, fabricated j and inspected in accordance with ASA 3 31.1 1955 Code i
, for Pressure Piping and applicable suelear case rulings.
The design pressure and temperature shall be 700 psig i and 5008 F respectively. The inlet and outlet piping from the reactor to the three recirculation p g s shall be 22 <=ches 0.D. x 5/8 f ach thick As1M-A155 Cr. EC-70 carbon steel internally clad with 1/8 inch thick type >
, 3MI, stainless steel.
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. g 4.2.1.2 valves Butterfly valves shall be located at the suction and discharge of each of the recirculation pumps. Se
', valve operators shall,be of the electrical motor.
gear drive type which fails in position on loss of i power. - he pump discharge valve operators and con- 1 trols shall be designed to limit total rate of flow increase to less than 455 spe per second. (his shall correspond to a mawi== reactivity addition l
> rate of less than 5 cents /sec.) .
De discharge vilves shall'be interlocked to prevent closure to below 45 percent open when the associated- !
Pump is running. l Control interlocks shall be' designed to prevent con-trolled increase of recirculation flow coincident with control rod withdrawal. ;
Operation of pep discharge valves may be used to control recirculation flow. Reactivity addition by t recirculation flow control shall not be continued for more than 10 seconds in any one 20 second interval. !
4.2.1.3 Pumps t-The recirculation ymps shall be designed to deliver 21,600 gym against si 71 foot head. sach pump shall be driven by a 400 HP induction motor. Re pump '
casing shall be made of ASTM A-351-57T grade CF-8
~
stainless steel.
4.2.2 Biological sl$ielding Esch recirculation pump and motor shall be in a separate shielded ! ;
compartment in the reactor building. .
4.2.3 Reactor n e volume of water in the reactor at normal operating level shall be about 13,000 gal. W e volume of water in the reactor when the reactor is flooded shall be about 20,000 gal.
4.3 Safety Relief valves:
Number 4 Type Spring-Loaded, Bellows sealed Maximum Setting of First Valve, Psig '600 -
Maximum Pressure Setting of Remaining Valves, Psig 620 .
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Safety Relief Valves: (Cont'd.) i l
Design Capacity per Valve, Founds per Bour 177,500 ,
Nossle Area, Square Inches 6.38 i Rupture Disc Design Pressure, Psi 250 1
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4 d Reactor Power Operation Cool'ina:
Coolant Material Domineralized water f Type of Cooling System Forced Recirculation f l
, System Pressurisation -
- Boiling Unter I, Mini - Loops Operating Concurrently (or Equivalent) 1 Nimmber of Passes Through Core 1 Plow Direction Through Boiler Core Upw.ard 4.5~ Reactor Shutdown Coolina:
Design Pressure, (Standby Cooler) 150 Psig (Shield Pool cooler) 50 Design Temperature, (Standby cooler) 350 ,
- F (Shield Pool Cooler) 212 !
Number Pumps 2 l Number Esat Exchangers Available -2 Heat Removal Capacity (Standby Cooler) 19.5 x 106 Beu/Er Heat Removal Capacity (Shield Pool cooler) 1.5 x 106 Btu /Er 4.6 Reactor Emergency Coolina:
Emergency Condenser:
Type Shell and Tube .
Minimum Capacity, Btu /Hr 17 x 106 Minimus Cooling Time Available 24 from Water Storage, Hours Coolant Shield Pool Water Design Pressure of Shell, Psig 55 Design Temp of Shell, 'F 300 I Design Pressure of Tubes, Psig 660 l Design Tony of Tubes, 'F 900 Actuating Signal Isolation scram
.i Maximum Time to put System in 30 see I Full Operation Following Signal, Seconds .
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4.7 Operating Requircaents _ .
During all reactor Power Operation in excess of 1 Mf(th) the j minimum recirculation . flow shall be at least (1) 297 gpm per '
I Mf(th), or (2) flow equivalent to design flow of one recircula-tion loop. l
'The maximum operating pressure and temperature shall be 700 psig and 500*F. The controlled rate of change of temperature in the reactor vessel shall be limited to 200*F/hr. All other com-ponents in the system shall be capable of following this 3
temperature change rate.- The safety relief valves shall be set appropriately for all planned reactor operating pressures so
~t hat the allowable pressure (pluse 10%) in the reactor system is not exceeded. The emergency condenser shall be operable and ready for service at all times during power operation.at levels aboveIMw(t). The shutdown cooling system shall be operable and ready for service during refueling operations if required for decay heat removal.
The primary coolant shall be sampled and analyzed at least daily.
during periods of power operation. The primary coolant shall be analyzed whenever the conducitivity increases unexpectedly.
The following are absolute limits which if exceeded shall necessitate reactor shutdown. Corrective action shall necessarily be taken at more stringent limits to minimize the possibility of ,
these absolute limits ever being reached. ?
Conductivity (Micromoho/ca) maximum 5 maximum transient
- 10 pH (Lower and Upper Limits) 4.0 and 10.0 Chloride Ion (ppa) maximum 0.1
. . . _ . _ . maximum . transient *- .. _ _ _ - . _ . 1. 0. . ._._-- . .._..._ _ .
Iodine (microcurie /al) 20 Boron, ppm, maximum except during 100 experiments below 1 Mw(th) 4.8 Biological Shield -
The reactor vessel shall be located in a cavity formed by the bio-logical ~ shield of standard stone aggregate concrete approximately 10 feet thick.
There shall be about a one-fdot air space between the concrete shield and the reactor vessel. Air shall be circulated in this
' space. Cooling coils in the concrete shield shall be designed to prevent the shield temperature from exceeding 180*F.
'Conducitivity and chlorideconcentration is expected to increase temporarily .
cftsr startups from cold shutdown. The time delay before the transients reach -
thsir peak will depend on the flow rate. The maximum transient values for p ccnducitivity and chlorides here stated are the maximum permissible and cpply only to a period not to exceed 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after reaching ~20%" rated. power subs:quent to a cold shutdown, i. g ,
4-6 l l
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j he top of the reactor shs11 be shielded by water in the shield pool. There shall be enough water in the shield pool to protect personnel working over the pool. l 4.9 Power System seuipment and Associated Facilities !
. i
! l 4.9.1 Electrical System l
(a) Auxiliary Power 6 Se auxiliary 'powe'r system shall be the normal source h of power to the plant under Power Operation, Refueling, '
. and Shutdown conditions. f i
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Por continuity of station auxiliary power, the system design shall consist of four independent sources of power; (1) S e 13.8 kv generator with voltage stepped down for station service, (2) the 115 kv system avail-
, able through a reserve station auxiliary transformer, (3) an automatic starting standby 125 kw diesel-generator and (4) an emergency 500 kva manually switched inter-connection outdoor with a local 12.5 kv distribution ,
line. The station a-c service system shall be divided I into three voltage classes: 2400 volt, 480 volt and j 120/208 volt.
The 2400 volt system load shall;be divided between two l independent busses which may be manually transferred i ,
from the 13.8 kv stepdown transformer to the 115 kv stepdown transformer or vice versa.
. .. . , The,480 volt system loa.d,shall,also be divided between two busses normally independent and fed from separate 2400/480 volt 1200 KVA, stepdown transformers. One of these transformers shall be fed from one'of the 2400 volt busses, the other transformer shall be fed from the other 2400 volt bus. In case of loss of either of these two sources to the 480 volt system, a 480 volt bus tie breaker shall automatically close to feed both 480 volt busses from the remaining source.
The station 208/120 volt system load (consisting
- primarily of lighting and fractional horsepower motors) l
! shall connect to a single bus normally fed from one of l the 2400 volt busses via a 2400 - 120/206 volt 300 KVA,
- i stepdown transformer with provision for automatic transfer j to a second stepdown transformer fed from the other 2400 volt bus. i 47 :
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(b) Emergency Power
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The diesel generator shall provide emergency 480 volt.
power for essential auxiliary equipment in the svent of a complete 480 volt station auxiliary powsr failure. ,
. The diesel generator shall start automatically after loss of power to both of the 480 volt buses.
The diesel generator loads shall *oe as follows:
i Boron Injection System Standby Lighting Turbine Auxiliary Equipment Reactor Building Spray Control
. Motor-Generator and associated reactor safety circuits The station battery shall supply power through the motor-generator to the circuits which are normally supplied by the 480 volt bus. The amargency A.C. and D.C. loads shall include:
Switchgear ,
Radiation Monitoring Containsent Building Isolation control Nuclear Instrumentation ,,
Annunciators Main Steam Isolation Valve Control Rod Controls .: -
Pertinent Recorders Waste Controls
%.s ....,;- m-... Reactor Building T ,ransmitters Turbine Controls Reactor Pressure Control l The station battery shall have sufficient capacity to l ,
carry the emergency D.C. load for at least 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. '
The diesel generator shall be tested monthly for load carrying capacity and automatic operation.
The motor-generator set shall be tested we,ekly for automatic operation.
s A 500 kva emergency backup source connection shall be available to the 480 volt bug during power operation.
4.9.2 Main Condenser i
! The main condenser shall have sufficient capacity to condense l full reactor steem flow while the turbine is being bypassed.
l The condenser design features shall be as follows:
4-8
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-Type Cross Flow Surface Condenser with Deaerating Hot Well Condensing Surface Area, Square Feet 67,500 Design Condensing Pressure, Inches Hg Absolute 1.5 Btu per Hour at 1.5 Inches 3
Hg Absolute -
451,250,000 Air Ejector Capacity 187.5 lbs/hr air and vapor 4.9.3 Turbine By-pass Control System (Dump Valve)
(a) Design Features shall be as follows:
~
Flow Capacity at 500 psig, Pounds 700,000 per Hour Maximum Time, Full Valve Stroke, Approximately Seconds -
0.5 4.9.4 Condensate and Feedwater System Two,1450 gpm, full-capacitrcondensate pumps of conventional design shall be provided to pump condensate from the con-denser through the feedwater heaters and filters to the ,
suction of the reactor feed pumps. l Two- full-capacity fi1ters, designed to each pass -1400 gpm,- ----"
j shall.be provided for removal of solids (Turbine-condenser system corrosion products).
There shall be two,1500 gpm, horizontal, centrifugal, motor driven reactor feed pumps. .Feedwater shall then pass through ,
a feedwater control valve, high-pressure feedwater heater, and check valve.
4.9.5 Main Steam Piping The main steam line shall be a 16-inch schedule - 60 pipe fabricated from A-335 Grade P-11 steel. The design pressure and temperature of the main steam line shall be 600 psi and 825*F. All connections to the main steam line shall be con-sistent with this pressure and temperature rating. h 4.9.6 Cooling Water System
- The cooling water system shall utilize circulating water to remove heat from the following pieces of equipment
- -
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Diesel Generator Oil Coolers Rotary Vasuus Fusy Gland steam Condenser Reactor purification Cooler Pool Coolers .-
Two cooling water pays, each rated at 15.00 syn, shall be provided to circulate water through the above equipment.
> 4.9.7 Fire Protection system-There shall be hose stations, estinguisher stations, and overhead sprinklers la the plant and hose stations outside ,
the plant. All areas of the plaat shall be covered by at least one station. Se hose stations and sprinklers shall be connected to a fire protection header.
Water shall be supplied to the hidder by two cooling water pumps, a diesel fire pump, and a well pump as backup.
Dere shall be chemical and CO2 extinguishers in the plant to supplement the water system.
4.9.3 ventilation system The induced draft fans shall exhaust air to the stack from
- the reactor building, turbine building, and the fuel
! handling building.
3 The following air exhaust ducts shall contain absolute i filters while in operation:
p . . . , x, . . . _ . , . ,_. .. .. .
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- 1. Condenser off-gas to I.D. plenum.
- 2. Condenser vacuus pump discharge to turbine b1dg. vent exhaust.
l 3. Decontamination room exhaust fan dis-charge to I.D. pleans.
- l 4. Hot chemistry lab exhaust fan discharge l
- to atmosphere.
i Theadministrationbuildingairexhauststotheaccessible area of the turbisie buildias.
l l (a) Turbine Buildina Turbine building ventilation air shall be delivered [
to the operating floor,' equipment rooms,'and shops.
Turbine building ventilation air shall be exhausted from the accessible clean area to the shielded area, t
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The turbine building ventilation air shall be either 100% outside air or a mixture of outside air and re-circulated air from the accessible area of the j turbine building. l l
(b) Fuel Bandlina Buildina l l
The fuel handling building ventilation system shall be divided into three sections: (1) The main venti-lation section shall deliver air to all sections of
> the building except the flash tank vault. (2) A separate exhaust fan shall draw the air from the de-ea=e==i== tion area as required. (3) The air for the flash tank vault shall be supplied through a. duct from the shielded area of the turbine building. ,
The ventilation air for the fuel handling building shall be 100% outside air with no recirculation.
4.9.9 Instrument and service Air System .
Instrument and service air shall be supplied by two air compressors, each rated at 317 scfm. Instrument air shall also pass through a dryer. -
4.9.10 Fuel Bandlios !
The fuel handled or stored as described in this section ;
shall be limited to fuel as described in these specifica- ;
tions.
. (a), Fuel Transfar _ _
Irradiated fuel elements and control rods shall normally ,
be transferred between the reactor shield pool and the fuel storage pool through a transfer tube located near the bottom of the pools. New fuel shall be transferred either through the transfer tube or through the per-sonnel airlock.
(b) New Fuel Storaae ,
New fuel shall normally be stored in steel racks in the storage vault. The boiler fuel racks shall be arranged ;
in three double rows with a 23-inch aisle between double :
rows. A maximum of 32 new boiler elements may be in ,
temporary storage in the reactor building during Refueling *
, Operation. [
The mart === calculated flooded K,gg for the boiler fuel storage shall be less than 0.80.
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The maximum calculated looded'K.gg for the superheater storage shall be less than 0.70.
Movement of new fuel elements fron,the storage area for refueling or inspection shall be under qualified supervision. !
The new fuel storage vault shall contain a maximum l of 156 boiler elements and 550 superheater elements. l There shall be approximately 173 lbs. of UO2 in each boiler element and approximately 0.33 lbs. of UO2 in each superheater element.
The boiler fuel element racks shall have 26 spaces for fuel elements separated by 3/8-inch. thick steel partitions.
There shall be 5 superheater fuel racks. Each fuel rack shall have 10 tiers. Each tier shall have 11 spaces for superheater fuel. he minimum spacing of superheater fuel shall be 1-5/8 inches center-to-center. The tiers shall be approximately 8 inches apart.
Amaximumof64superheaterelementsmaybeinNem-porary storage in the' reactor building during Refueling Operation.
The door of the new fuel storage vault shall be locked, except when fuel handling activities or inspections require that it.be unlocked, and the key shall be under strict. supervisory control.
(c) Irradiated Fuel Storage ,,
The fuel storage pool shall be 21 feet deep in the fuel storage area and 29 feet deep in the cask * -
loading area. Were shall be 9 double rows o'r 262 storage spaces provided by steel storage racks securely anchored ta the concrete pool. Each space shall be able to contain one boiler element, a basket of up to 16 superheater elements, sources, or other materials. De racks shall be arranged in double rows separated by approximately 7-3/4 inches.
We channel between-rows shall be too narrow to pass more than one boiler element at a time.
. I Neighboring fuel elements in the rows shall be sep-arated by 3/8-inch thick stainless steel. We calculated maximun reactivity for a new 3.2% enriched I l
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( g spike element passing between the racks at the most reactive location shall be 0.82 or le'ss. The calcu-laced reactivity of a basket of 16 superheater elements shall be less than either 2.2% or 3.2% enriched boiler elements.
Temporary storage for 8 boiler elements or superheater baskets shall be available in two fuel transfer boxes in both the fuel storage pool and the shield pool.
ne calculated maxisman gg, of these boxes with new 3.2% elements shall be 0.76 or less.
W e fuel storage. area shall be monitored for radiation whenever fuel is handled. The monitor over the storage pool shall be set to actuate local and control room alarms.
4.9.11 Turbo-Generator The turbine shall be a tandem-compound,. double-flow unit.
The turbine steen seals shall be designeI to prevent radio-active leakage to the turbics room. Non-radioactive steam shall be supplied to the seals by the gland steam evaporator or a fossil fueled gland steam generator.
The generator shall be a direct coupled, hydrogen cooled generator rated as follows:
88,235 kva 85% pf 3600 rpm 3 phase 60 cycle 13,800 volts r
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5.0 Reactor Core and Controls' The reactor core shall be as described in this section. The location and arrangement of such-components shall bit as shown in Figure 1.1 of ACNP 5905, dated January 15, 1962 and Figure 1 of these specifications.
The method of positioning each particular component within the core and the design features of each component shall be as specified in sub-sections 5.1, 5.2, 5.3, 5.4, and 5.5. :
The nuclear characteristics including the reactivity characteristics, ,
control rod drives, and boron injection system are also specified in t this section. ,
5.1 Boiler control Rods There shall be 16 boiler control rods,.of irudiform shape located in the core as shown in Figure 1. Each boiler control rod shall be cruciform in shape as shown in Figure 1.17 of ACNP 5905, idated January 15, 1962. The rod shall bajof all welded construction and shall be as described below Poison Material in Rods 2.0% Natural Baron in 304 SS Noseinal Active Length, Inches 72 Nominal Width, Inches 10-7/16 Nominal Blade Thickness, Inches 1/4 5.2 Suoerheater Control Rods There shall be four superheater control rod support yokes, as shown on
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' Figure 1'.19 of'ACNP 5905 of January'15, 1962, located in the core as ' ~
shown in Figure 1. The superheater control rods shall be as arranged in Figure 1.18 of ACNP 5905, dated January 15, 1962, 6 shall be as described below Number of Rods per Assembly 12 Poison Material in Rods 2.0% Natural Boron in 304'SS
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Nominal Active Length Inches 72 , .
Nominal Rod Diameter, Inches 3/4N Cladding 304L SS i Nominal Cladding Thickness, Inches 0.060 5.3 Poison Shias Poison shis's may be used for reactivity adjustment and initial physics requirements. Up to 70 shims may be placed between boiler boxes. These shims shall be prevented from moving sideways or upward by slots in the I
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top and sides of the boiler boxes and the hold down structure. The poison shims shall bet -
Material 0121. Natural Boron in 304 SS Nominal Length,'. Inches 74 Nominal Width, Inches 2-25/32 Nominal Thickness, Inches - . 0.10 ,
5.4 Core Composition The data presented in this section consists of design features of the fuel which shall make up the physical composition of the core as arranged in Figure 1.
Total lhamber of Boiler Assembly locations % ,
Number of 2.2 w/o U-235 Boiler Fuel Assemblies, Max. 96 Number of 3.2 w/o U-235 Boiler Fuel Assemblies, Max. 32 :
Nominal Total Weight of U-235 in 96 - 2.2 w/o Fuel Aspemblies, Eg. 145 Nominal Total Weight of U-235 in 64 - 2.2 w/o Fuel .
Assemb119s and 32 - 3.2 wife Assemblies, Eg. 167 Max. Number of Superheater Fuel Elements in Core 412 Enrichment of Superheater Elements, w/o U-235 93 Nominal Total Weight of U-235 in 412 Superheatgir Fuel
. Elements,.Es. , . . ,
50 5.4.1 Boiler Fuel Assembiv The'bo11er fuel assembly shall be as in Figure 1 3 in a ACNP 5905 dated January 15, 1962, with the fuel rods as spect-fled belows (a) G metal Nominal Square Bundle Outside Dimension. 4.735 Inches Geometry [FuelRod' Array 9x9 Number of Rod Sections per Rod 4 Fuel Rods per Bundle 81 f Number of Tube Sheets per Assembly s 3 I.
5-2 I
i ,
Fig. CORE PLAN AND CHAMBER LOCA*. 15 .
= Control Rod Nutr.bors
~
~ = Number of Chamber
- Locations Available ll '
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BF BE BD BC BB BA
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hree startup channels - Two shall be diametrically opposite Two linear channels shall be diametrically opposite Two Power channels shall be diametrically opposite i Diametrically opposite shall be: N-S, NE-SW, E-SW, NE-S, N-SW I 5-3 . . . . . _ _ _ _ - - - .
- I 1 . . .
1 l
l 1
(P Fuel Rod Claddina (Di - stons at Room Teasersture)
Haterial Zircaloy-II Clad Thickness of Upper Sections, Inches 0.026 i CIM* Thickness of Lower Sections, Inches dB <
. (c) Fuel Rod (Dimensions at Room Tencerature) s Rod Diameter of Upper Section, Inches 0.367 Rod Diameter, Lower Sections, Inches O.408
- Fuel Pellet Diameter, Upper Sections, Inches 0.310 Fuel Pellet Diameter, Lower Sections, 0.348 Inches Fuel Pellet Density, sm/cc 10.41 5.4.2 Suoerheater Fuel Elements The superheater fuel elements shall be as arranged in Figure 'L.5 of ACNP 5905, dated January 15, 1962. The fuel tubes shall be as specified, belows Fuel Tubes Outer 0 D. Inches .839 Outer I D. Inches .769 Inner 0 D Inches .630 Inner I D, Inches .560 UO in Cermet, w/o 17.5 2
Poison Tube O D. Inches .467 i l
Poison Material, w/o Natural Boron 0.35 i Carbide in Alumina Cladding Material 316L SS Cladding Thickness Fuel Tubes, Inches .0075 Poison Tubes. Inches .028 Up to 72 superheater fuel assemblies without boron in the poison pin may be used for flux shaping.
5.4.3 Boiler Fuel Boxes
~ Th'e'bo~iler fuel" Boxes shall'be'as*Yd116ws? ""~ -- '*~ '"~~"~ ~
Material Z1Tcaloy-II Number of Single Boxes 32
. 5-4 4_ -
e - - - - - - _,. . . - - , v. . ,, -
s . p 4
Number of Quad Boxes 16 r Wall Thickness'(Nominal), Inches .1 56 5.5 Sources 5.5.1 Initial Start-uo Source During initial fuel loading and low-pressure testing, one 6-curie plutonium-beryllium neutron source shall be used.
This source shall be positioned in a superheater process tube.
At all times during.the use of this source, the maximum thermal flux shall be limited to 5 x 10 01 av at the position of the source.
5.5.2 Doeratina Source Type Antimony Beryllium Quantity 1 Location Superbgeter Minimum Initial Strength 5 x 10 n/sec.
(a) Physical Description The beryllium annulus shall be part of a special inner insulating tube of the superheater. The antimony rod shall located in the center of the beryllium cylinder. The dimensions of the assembly shall be as follows: i Total Length of Insulating Tube, ' Inches"-/19411/32 Insulating Tube O.D., Inches 1 Total Length of Antimony, Inches 76 7/16 Length of Beryllium Annulus,, Length 73 3/8
- 5.6 Princloal Calculated Thermal Hydraulic and Nuclear Characteristics ;
The core shall have the following calculated design parameters. The operating limitations upon the core are given in Section-5.7. - - -' '
i 5.6.1 Principal Calculated Thermal and Hydraulic Characteristics of the Core Loadina at Full Power (a) Core Power at Rated Steam Flow, MWT 190 (b) Boiler Peaking Factors (To be applied to Heat Flux) '
Radial 2.00 Axial 1.35 Fuel Loading 1.03 ,
Fuel Eccentricity . 1.25 l
~ Combined 3.47 i
i
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5-5 ,
e 6
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(c) Heat Flux and Fuel Center Temperature at 190 W(th) Maximum Steady-State Power l Max. Boiler Heat Flux, BTU /hr ft 447,000 Max. Superheater Heat Flux, BTU /hr ft 219,000 Boiler Cladding Surface Temperature, F 514 Max. Superheater Cladding Temperature, "F 1270 5
Max. Linear Heat Ceiwration Rate of Boiler Fuel, kv/ft 14.0 Max. Superheater Fuel Temperature, F 1280 (d) Burnout Ratio,* Minimum 1.9 (e) Max. Fuel Cladding Stress, Psi (55% of Yield Strength) 12,790 (f) Average Core Power Density, Kw/ft 1280 (g) Total Recirculating Flow Rate Max., spa 65,000 (h) Boiler Flow Rate, Percent of Total 89.7
. Pccircul .atin?. Flov ,Iirte , .
(i) Superheater Flow ilate, Percent of Total Recirculation Flow Rate 6.0 (j) Core Inlet Conditions Inlet Velocity:
Maximum, Ft/Sec. 14.2 Minimum, Ft/Sec. 12.5 Inlet Subcooling,- Btu /Lb 3.8 u
(k) Reactor Boiler Core Pressure Drop'at '
60,000 gpm Flow, Psi 13.6 (1) Boiler Exit Bulk Temperature at 190 M(th), F 489 (m) Superheater Exit Bulk Temperature ** F 725
- Calcukatedat600psigbyusing
, 9 B0 "H -4.12
= 0.0536 103
.g -
i This correlation, with the exoonential derived from laboratory data has been biased by a factor of. 75 to cover all data points.
- The superheater exit bulk temperature shall increase to a maximum at a core life of approximately two months and then decrease. The maximum shall not be greater than 7500F.
5-6 6
4 . p I
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(n) Steam Volume Fraction Average Core Exit Void Fraction .43
' Maximum Exit Void Fraction .56 ;
Average Void Fraction Over Core Length .31 (o) Minimum Core Inlet Pressure, psig 586.4
.. i 3
5.6.2 Principal Calculated Nuclear Characteristics of the Core ,
The calculated physics parameters shall conform to the j values tabulated below '
' l (a) Minimum Nena'tive Temperature Coefficients (delta kegg/ F) i
' Temperature Coefficient i
~
~
68'F -3.0 x10 f
~
68 F at end of itfe -1.2 x 10 !
~
450 F -8.1 x 10 '
450 F at end of life . -4.0 x 10" ,
(b) Minimum Nenative Void Coefficient (delta k egg/1% eg)hvoids)
Reactor Power Delta k/1% v -
Hot, zero , .3 x 10 ~3
~ ~
50% power -1.0 x 10 100% power -1.6 x 10 ~3 (c) MinimumNenativeUO2(Donoler) Coefficient (deltakgh/%) e Reactor Power Delta k/oF Cold, zero voids -1.22 x 10-5 Hot, zero voids -1.1'5 x 10-5
'50% Power ' 0.97 x'10-5 100% Power -0.92 x 10-5 (d) Mjnimue' Pressure Coefficient (delta'k egg/ psi).
I t Pressure -At 20% Power At 100% Power _
400 psi +3.3 x 10-5 +7.5 x 10-5 600 psi +1.5 x 10-5 +4.6 x 10-5 l 800 psi +.9ts x 10-5 +3.3 x 10-5 5-7 !
.p f .
(e) Calculated Core Reactivity Cold, clean, SH flooded K - 1.1232 Full power, ept-11brium'Xe and'Sa',
435 exit voids K = 1.0184 Temperature defect, delta k/k (including doppler) - .0245 Superheater draining, delta k/k + .0051 Voids (43% exit voids), delta k/k - .0329
. .e Xenon, delta k/k -i.0265 Samarium, delta'k/k - .0102 (f) The maximum reactivity addition rate shall be 5 cents /sec.
(g) Worth of Lieutd Poison -The reestivityWorth of the liquid poison system shall bet Reactee'-wessel open to shield pool. - 0.07-delta k,gg Normal water level in reactor, -
0.30 delta k,gg Time to inject 1000 gal. min. 18 Max. time to inject enough solution to obtain 4% negative reactivi.ty, minutes 5 I
(h) Fuel Burnuo Average Mwd / tonne of contained U for the.first >
core 7800 i Maximum Mwd / tonne of contained U 22,000 5.7 Principal Core Operatina Limitations 5.7.1 Reactor Power Level (a) Refuelina Partial core tests may be run, but the reactor power shall be limited to 1.0 Mwt, exclusive of core decay
' heat; %en using 'the" plutonium-beryllium source, the '
power shall be limited as described in 5.5.1 I (b) Reactor Operation The reactor shall be operated within the following limits:
5-8 l i
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l Minitsum Core Burnout Ratio,* steady state 1.9 Transient Minimum Burnout Ratio ** for maximum heat flux conditions 1.5 Maximum Steady State Heat (Flux BTU /hr ft2 447,000 Maximum Steady State Fuel Clad Stress, 55 Percent of Stress Yield Superheater Power Maximum Steady State '
Power Fraction *** 0.17 Boiler Power Maximum Steady State Power
. Fraction of 190 Mwt 0.86 MaximumSteadyStatePowerLevel,Mw(thy 190 Maximum Steady State Vatee 6f.C6re Power Density, Core Volume,Total,CorgPowerDividedbyTotal kw/ft 1280 Minimum Reactor Pressure at Rated Power, Psig 500 Maximum Steam Temperature, F ,
750 Maximum Reactivity (delta k/k) in St'eam Volds .035 Maximum Reactor Pressure During Power Operation, Psig 660 Minimum Recirculation Flow Rate,**** !
8Pm/MW(th) above 1 MW(th) ,
297 Maximum MWD / tonne of Contained Uranium '
for an Individual Fuel Rod- 22,000
- Burnout ratio is refined as the ratio of burnout heat flux to actual heat flux at a point in the core and shall relate to the burnout correlation stated in Sec. 5.6.1.
- Evaltiated at 115% of rated power avid compensated forl total measurement errar.
- The superheater power fraction shall' increase to a maximum
.at a core life of approxipately two months and then decrease.
The maximum'shall not be greater than .17.
- $ And'as spioified in Section 4.7.
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The control rod withdrawal rate during power operation shall be such that the average rate of reactor power increase is ,
less than 50 W(th)/ minute when power is less than 120 W(th) and less than 20 W(th)/ minute when power is between i 120 and 190 W(th).
5.8 Control Rod Drives !
6 Twenty control rod drives shall be mounted on the vessel head. The h drives shall be of the rack and pinion type, driven by an induction i sear-motor. The reactor shall be allowed to continue to operate if f' not more than one control rod drive is inoperative because of failure, the cause of f ailure ascertained, and the cause of f ailure is not apt j to progress to other drives, and the shutdown margin stated in 5.9.2 l l can be met with the remaining drives. l l
There shall be sixteen crucifers control rods in the boiler region. l There shall be four rod-type control assemblies in the superheater i region. ! l Each control rod shall be driven by the gear-motor through a magnetic clutch and a can clutch. The boiler control rod drive magnetic clutches shall release and allow the boiler rods to fall into the core on a reactor scram signal. A scram signal shall also initiate "run back" on all control rod drive motors. i The control rod position shall normally be transmitted to the control room by a Selsyn-type electrical servo transmiter. The position indication system shall have a change of position accuracy of t 1/2 inch.
The boiler rods shall be accelerated during a scram by a compressed spring when the drive rack is in the " full out" position. The drive rack and attached control rod shall be decelerated during the last six inches of the downward stroke by a hydraulic dashpot. The magnetic clutches for tho6superheater control rod drives shall be powered by
- .is >.
j Ce6 trol rod drive festures shall be as follows: .
M -
Rack and Pinion Epraal Stroke Length, Inches 73.0 .
Maximum 91thdrawal Velocity In/Sec.
Outer Boiler and Superheater 1.2 laner Boiler .4 Minimus Insertion Velocity In/Sec.
Duter Boiler and Superheater 1.0 Inner Boiler 0.3 l
5-10 l
l
- Maximum Scram Insertion Time, Sec., 90% of Stroke 2.0 5.9 Control Rod System 5.9.1 Control Rod Performance ,
The following tests shall be perfornwd at each major refuelhag
)
.thutdown. .
(a) ~ Continuous withdrawal and insertion of each drive over its stroke to verify velocities stated.;in section 5.8.
(b) Withdrawal of each rod to check the functioning of the position indication system.
(c) Scram of each drive from the fully withdrawn position.
Maximum scram time from rod release to 90 percent of insertion shall not exceed 2.0 seconds.
Each drive shall be moved to determine by motor current that !
drive Se6eteen is normal' at each major refueltag but not less- '
frequently than quarterly during the period of the provisional operating license.
5 9 2 , Core Shutdown Marain verification The reactivity of the core loading shall be such that it is always possible to maintain k~egg at less than 0.997 with the most valuable
' ~
r'eactiviYy-worth control rod c5mp1'etely'wittidfavn~frsin the coFe'in any operating condition.
The core shutdown margin shall be verifled by a demonstration that the reactor is suberitical with the most valuable reactivity-worth rod fully withdrawn, the superheater drained, and-a rod or a rod group withdrawn to a position known to contribute 0.003 k,gg or more to the off active multiplication.
This verification shall be performed prior.to startup after any shutdown in which the siystem has cooled sufficiently to be opened to atmospheric: pressure and any of the following events have occurred e& ace the previous verifications l (a) Fuel has been added or repositioned in a way which is not definitely known to reduce reactivity; or 2'
(b) A control rod has been changed and presence of poison has not been verifled; or
- The time interval after scram signal causes the boiler control rod clutches to release until the rod has traveled the specified distance i of its full stroke length. . I l
5-11 ,
1
, l l e *
(c) Poison shtms have been relocated or removed from the !
coral or [
~
(d) 31,000 WD(th) have been generated by the plant since the previous margin demonstration. i During power operation, if core reactivity increases for no explainable reason, the reactor shall be brought to the cold 3
shutdown condition. -
5.9.3 Control Rod Drive Temocrature ,
The rod drive rack housings shall be filled with water and i submerged in the reactor shield pool. The housing and '
condensate temperature shall be approximately the temperature j of the shield pool water, however the shield pool need not be '
flooded during reactor operation if the reactor coolant is below 200*F. !
5.9.4 Control Rod Latchina Checks I I
The control rod latch shall be designed to prevent the removal l of the latching tool unless the rod is completely latched or i untetched to the drive.
l The operation of the latch shall be verified by observing the !
drive motor current dif#erence while moving the drive before i
, and after the rod is attached.
l The operation of the latch shall also be verified by observa- '
, . - - ~ . . ~
gion og the nuclear-instrumentation response-to Tod withdrawal. -
i 5.9.5 Control Rod Exercisina Durina Sustained Power Ooeration l i
- Each control rod, which is either partially 'or completely with-drawn, shall be exercised at least once per two week' period.
5.9.6 Abnorms1 Behavior of the Control Rod System
- An lamediate and thorough investigation shall be made of. the occurrence of any abnormal behavior of the control system to
, determine the cause and safety significance of the occurrence.
The reactor shall be shut down unlesst l
(a) It is determined by the investigation that any as1 function l which has occurred neither impairs the ability to control l the reactor nor indicates the imminent impairment of the l performance of additional components of the control rod system.
5-12 9
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(b) The core shutdown margin requirement (described in 5.9.2 above) can be met with the remaining operable control rods.
Evaluation of this requirement shall be based on previous experimental measurements.- '
5.9.7 Minimum Accuracy of Rod Position Indicatina System
, The position indicating system is of a. selsyn type and indicates the drive position over full range of the drive with a change of position accuracy of + 1/2 inch. If the position indicator for a particular control rod malfunctions, the rod shall not be moved except for scram or complete insertion by runback until the indication has been restored.
5.9.8 Operating Reautrements The maximum-burnup boiler and superheater control rod shall be removed from the core at one-half design burnup, 0.5 and 0.25 a/o burnup respectively and examined for any detrimental effects.
The first inspection shall be made not more than one year"after' reach-ing full power. Subsequent inspections shall be made of at least one rod at each major refueling.
The surface of the control rods shall be examined. Particular care shall be taken in the area of maximum burnup and in weld areas. Any discontinuity shall be examined at higher magnification.
The superheater rod shall be checked for excessive swelling.
_. . . _ - . ~ - - -
Any-Eod'~ showing indications of-' surf ace cracks or other defects -- -
which are jt.dged to adversely aff ect reactor operation shall be replaced in the core by,a new rod. Further, all rods of.the type showing the defect shall be removed and examined in the prescribed manner.
5.10 Boron Iniection System Desian Material Na2.B 08 13 16 H2 ,
Available Quantity of Solution, Gal. 1000 l Total Weight of Boron, Pounds 230 .
Maximum Total Injection Time, Minutes 20 System Actuatiov. Remote Manual A 1000' gallon tank containing a boron solution shall be located in the reactor building and shall be operable during reactor operation. 'The ~<
solution.shall provide a minimum of 7 percent negative reactivity when 5-13 '
i
~
KIN'.y%.'. t Y l' F-the reactor is open to the shield pool. The seal pumps shall be used j to inject the solution into the reactor. The boron injection shall be manually controlled from the main control room. The system shall !
serve as back-up for the normal control system. When the reactor is in operation, room temperature shall malatain the boron solution above 58'F.
5.11 Boron Iniection System .
The boron injection system shall be available for operation at all times during Refueling Operation and Power Operation. The reactor shall be shut down in any situation where the poison solution tank level drops below an equivalent 1000 gallons of .23 lbs. boron / gal.
solution. The minimum temperature of the solution shall be 580F.
The minimum worth of the liquid system (based on normal reactor water level) shall be 30% delta k,gg. The liquid poison system shall be used at any time when subcriticality cannot be assured by the normal 4hutd6sCmechanism. Injection shall be continued until a minimum shutdown margin of 0.01 delta k,gg/k,gg is assured in the most reactive core. The maximum time to inject enough boron to obtain 4% negative reactivity shall be five minutes. The system components shall be ;
checked for operability at least once every two months of Power ,
~
Operation. The reactor shall not be operated after poison has been i injected until the boron concentration in the reactor water has been reduced to 100 ppm or less.
5.12 Reactivity Coefficients The reactivity coefficients shall meet the following requirements:
1 5.12.1 The effect upon reactivity of increasing voids at l constant pressure shall always be negative.
l l 5.12.2 The moderator temperature coefficient-(inferred from '
l l critical control rod position) shall always be negative. l l
5.12.3 The overall effect of increasing reactor power at .
constant pressure shall be a loss of reactivity whenever l l the reactor is operating so as to produce a net steam flow.
5.13 Reactivity Additions During Core Alterations The limits and requirements which apply to reactivity additions l are as follows: t 5.13.1 Any refueling operation which may increase reactivity shall utilize the procedure outlined in 5.9.2 before and after the alteration to verify the core shutdown margin of 0.3% delta k,gg/k,gg with the superheater . voided and the 5-14 g_ ,-..,_,-,..-.y..--._ _- ,_ . . . - , .-,. , . , . _ , - . ,
most valuable rod completely withdrawn. All rods shall be fully inserted during fuel additions. Checks shall be l made at frequent intervals during core alterations to assure 1 that the core shutdown margin requirement is being met:
I 5.13.2 Core alterations which increase reactivity shall be limited between suberiticality checks to fuel loading incremnants !
(
which do not exceed-one-half the reactivity addition for j criticality or the placement of one boiler fuel bundle. l At no time vill the shutdown margin be knowingly allowed to be less than 0.3% delta k /keff with the control rod of highest reactivity worth Ibly withdrawn from the core and ,
the superheater voided. - -
i f
5.14 Reactivity Additions During Power Operation Roetine control rod withdrawal sequences shall be established for use during normal Power Operation. These shall be in a sequence of steps involving withdrawal of only one control rod at a time.
The maximum reactivity insertion rate when the keff of the core is ,
greater than .997 shall be 5 cents /see and this insertion rate shall continue for no more than 10 see in any 20 second interval.
A e
5-15 -
- t. m._ ___ ___m
6.0 PLANT SAFgTY AND MONI.,,EING SYSTEMS (
This section specifies the general arrangement,' design features, and operating requirements of the plant safety and monitoring systems. These systems shall include the nuclear instrumentation, process instrumentas-tion which initiates safety system actions, reactor safety system, control rod withdrawal permissive system, and the plant radiation monitoring i systems. The term " rated power" as used in this section shall mean a reactor power level not exceeding the highest level of which all the operating limitations of Section 5.7 are met. , .:.
2._
No bypass of the automatic safety functions, including interlocks, shall be provided except as specified in this section. Except as explicitly provided in this section, no such function shall be bypassed in the course of any operation for which the service of that function is required by these Technical Specifications.
i 6.1 Reactor Safety System 6.1.1 General Features The reactor safety system shall consist of sensing devices to monitor reactor-associated parameters of plant operation and
' related circuitry designed to initiate appropriate safety, action when operating parameters exceed applicable setpoint-boundary conditions. The related circuitry shall be designed l
to initiate runback, scram, main steam isolation scram, closure of containment penetration valves, operation of the emergency condenser, or interlock action to prevent control rod withdrawal as appropriate safety actions.
Control s, witches shall'he' located in the control room to
- - permit manual initiation of scram, main steam isolation - - -
scram, closure of containment penetration isolation valves, and operation of the emergency condenser. The design shall also include control ' room provisions to manually control i cperation of' the reactor building spray system, emergency condensate supply valves, and the boron injection system.
. ... - - ~ . . , . . . .
6.1.2 Scram control System This section specifies the general arrangement and design f features of the scram control system. Principal components of this system shall be two input logic circuits and two silicon-controlled-rectifier type scram-clutch power supplies.
The input! logic circuits which control the excitation to each supply shall receive input signals from the flux-level = trip circuits of nuclear channels 5, 6,. 7 and 8; from 2/3 logic circuit which receives short period trip signals from channels 1, 2 and 3; and from' process monitoring instrueentation.
This system shall be designed to cause reactor scram upon receipt of high flux level trip signsls from any two of channels, 5, 6, 7 and 8; short period trip signals from any 6-1 :
l l
a . .
,.e.... . , two of c' unto 1, 2, and 3; er a trip ci-ci from any enn prac o monitoring iactrument. The a, Aes shall b3 so l
- designed that when reactor power is below a preset low level, >
power channel 7 shall insert one of two coincident signals required to cause scram. .
. . The scram control' system shall also have: a backup system which enables operation of either input logic circuit to cause reactor scram by interrupting a 120 volt ac power to both clutch power supplies. The backup system shall include four ,
backup circuits. Each backup circuit shall consist of a relay l driver which controls a relay capable of interrupting the 120 !
l volt ac power to both clutch power supplies. Each input logic ,
circuit shall control two of the described backup circuits. !
> The backup system design shall include provision for testing
~
i without causing reactor scram. Such test circuitry shall j permit _ momentary bypass of no pore than two backup relays at i one time and shall include design features of spring action l to prevent sustained bypass by operator error. .
t 6.1.2.1 System Response Times I The maximum response time from generation of signal by '
detectors of channels 5, 6, 7 or 8 until rod motion is detected by operation of the limit switch shall be 300 milliseconds. The maximum response time of the scram !
control system shall be 30 milliseconds. l 6.1.3 Nuclear Instrumentation The nuclear instrumentation consists of eight channels. The I instrumentation shall monitor reactor power from source level ,
to 150% rated power. The channels and associated safety
~
circuits shall be as specified in Table .la.
6.1.3.1 Channels 1. 2. and 3
, l f Channels 1, 2, and 3 shall provide logarithmic neutron flux level and period information for the reactor safety system and indication from source level to a level covered by operating channels 5 and
- 6. The principal components of each channel shall be a neutron detector, high voltage supply, pulse amplifiers, log count rate meter and period meter. -
The detectors shall be BF3 gas-filled proportional counters with minimum rated sensitivity of 12 ,
cps /nv thermal. Each channel shall have two trip i e circuits for short period and one for low count .
l rate output signals.
6.1.3.2 Channel 4 I
Channel 4 shall provide logarithmic neutron flux ;
level and period information for the reactor safety ,
system and indication from approximately 10-31 to l 150% rated power. The principi components of the ,
i 6-2 !
l
._____J
1
,AS FOLLOWSt Tcblo 1 AUTRONIC INSTRUMENTATION SHALL Range Safety Circuits l Chann01 Type Number of Channels l Start-up BF3 .3 Source to 105 Rod withdraw interlocks -
cps Permit rod withdraw at 362 cps, 2/3 Scram and Alarm -
Minimum period not less than 4 sec, 2/3 coincident logic No coincident logic reg'd for
- alarm Int rmediate CIC 1 .001% to Rod withdraw interlock -
Log - N 150% rated Permit rod withdraw at *.001% of power rated power (key switch bypass available )* ,
Rueback and Alarm -
Minimum period not less than 5 sec (key switch bypass available)*
Intermediate CIC 2 .001% to Runback and Alarm -
LINEAR FLUX 150% rated -Maximum indicated level not power greater than 110% meter indication on any selected range Runback on 2/4 logic with ch 5.(-?>S
- Rod Withdraw Interlock -
If either channel is " Low Out
of Range," set point not less e .than $1 of selected range, except on selected ranges at or less than 10-31 rated power range. l Scram and Alarm -
Enz indicated level not greater than 115% meter indication on any selected range. Scram on 2/4 logic with ch 5-6-7-8 2 1% - 150% Scram logic switch -
Pow:r IC rated swer Logic change from 1/3 'to -2/4 coincidence (ch 5-6'-7-8) when ch 7 )s 10% rated Tower Runback and Alarm -
Level not more than 110% rated power Runback logic 2/4 coincidence ch 5-6-7-8 -
Scram and Alarm -
Level not more than 11S1 rated i power.' Scram logic 2/4 coincidence ch 5-6-7-8 o S:e sec. 6.2.1 for operating requirements on bypass key switches i
! channel shall be a neutron detector, detector voltage i power supply, log-N moter, period meter and log-N recorder. The detector shall have a gamma-com ion chamber with sensitivity rated at 4 x 10-l'pensated '
I j amperes per av. .This channel shall have two trip
- circuits for short period and one for low level out-put signals.
6.1.3.3 Channels 5 and 6 i
Channels 5 and 6 shall provide linear flux level and selected range information for the reactor safety system and indication from approximately 10-31 to 3
1507. rated power. The principal components of each channel shall be a neutron detector, detector voltage power supply and picosameter with console-mounted range sutteb. The detectors shall be gamma-compensated ion chambers with sensitivity rated at 4 x 10- w amperes per av. A recorder shall record range switch ~
position and linear power indication of the selected channel. Both channels shall sound an alarm when the indication is low out-of-range. The low out-of-range alarm may be bypassed during startup by switching the channel to the lowest useful range. Each channel shall
~
have one trip circuit for low level and three for high level output signals.
6.1.3.4 channels 7 and 8 Channels 7 and 8 s' hall provide linear neutron flux level information for the reactor safegy system and indication from approximately 1% to 150% rated power.
.--The principal components tf sach channel-shall be- a--- - -
neutroa detector, . detector power supply. .and linear power meter. The detectors shall be uncompensated ion chambers..with sensitivity rated at.not less than '
c 2.6 x 10-14 amperes per av. A comparator shall i sound an alarm when the difference between channel 7 and 8 power level indication exceeds the comparator I
< set point of not less than 10% of full scale. Each
' channel shall have two trip circuits for high level output signals and channel 7 shall have an additional trip circuit for low level output signal.
6.1.3.5 General Service Recorder and Scalar .
A general service recorder with selector switch shall.
be provided to record log count rate or period from channels 1, 2, or 3; period from channel 4; or: power level ,from. channels 7 or 8. I
~
A scaler with selector switch shall be provided to 4
measuut. count-rate from discriminator pulse-output '
l of channels 1 2, or 3. ,
6-4 .
6
.~- -._y_.., - -, - . -- ,-o,.__..._ ,..m. __. .._.c- _
,- - - - . , . _ , . . . - . -.-.3-. . .w - , . . , . .
- 1; 6.1.4 Main Steam Isolation Scram Main steam isolation scram conditions shall initiate scram through operation of the scram control system. By use of a related circuitry these s,ame conditions shall also initiate the following actions: j I
(1) Close the main steam line isolation valves ,
(2) Sound reactor building evacuation alarm (3) Operate emergency condenser (4) Close reactor safety valve discharge isolation valve ,
(With a delay of up to one minute) . I
- i s
A main steam isolation scram may be initiated manually from -
the control room and shall be initiated automatically if any of the following conditions exist when the main steam isola- t tion valve is open. ,
I' Condition Set Point Loss of circulating water pumps Breaker trip or loss of irus power l
Turbine building ventilation not more than 10 tissa exhaust radiation high normal background or i 10 ar/hr, whichever is larger ,
i Air ejector exhaust not more than 10 times radiation high normal full power back -l 1 ground Main steam line radiation high not more than 10 times - .
normal full power back ground Low reactor water level not less than -3 feet
- Reactor building pressure 5 i 1 psig 4 l
6.1.5 Reactor Building Isolation Reactor building pressure of 511. psig or higher shall isolate f the reactor building by closing the following isolation valvee-l i i
- 1. Purification System Isolation Valves !
- 2. Shield Pool Cooling System Isolation Valve l l !
- 3. Recirculation PumpSeal Water Return Isolation Valve -
l
- 4. Sump Isolation Valve ,
- 5. Heating System Condensate Return Isolation Valve s
- 6. Ventilation Cooling coil Water Return Isolation Valve ;
I o Note: Reference water-level sero is 4 feet above fuel 6-5 i
,y. ,
. . m N
- 7. Ventilation Inlet and outlet Isolation Valves
, 8. Safety Valve Isolation Valve
- 9. Main Steam Isolation Valves. . .
1 1
These last three valves shall also clos,e on radioactivity levels as discussed in Table 6 of Section 6.4 and in Section 6.1.4. ,
a 6.1.6 Reactor Scram Mode Selection The reactor safety system shall include a " flood-operate" -
mode switch.which permits the selection of two modes of f operating conditions of the syates. The following !
conditions shall automatically initiate reactor scram and ,
shall not be bypassed by the " flood-operate" mode switch.
i Condition Set Point j Reactor Pressure Low not less .than 500 psig (Manual startup bypass cleared automatically when pressure rises to normal.)
Reactor Pressure High not more than 700 psig Reactor water level low not less than - 2 feet
- Reactor water level low (backup) not less than - 3 feet *
. Reactor building pressure high not more than 5 i i psig
~~' - ~
Reactor safety valves open i Reactor building ventilation exhaust not more than 10 times radiation high normal full power back-ground or 10 ar/hr, whichever is larger I Nuclear instrument scrans (see Table 1)
Control power key switch Off 40 lb instrument air header not less than 25 psig ,
pressure low
! o - Note: Reference water level zero is 4 feet above fuel.
6-6 l
- , , , , , -,,,,,,I
?
When the mode switch is in the " flood" position, the reactor shall scram when nuclear chaamel 5 or 6 is switched to a ,
range for which the full scale indication is greater than 47. '
rated power. The following conditions shall satamatically '
scram the reactor when the mode sensak is in the " operate" position and asy be bypassed by the mode switch in the " flood" l position.
- - Condition Set Point .
I ;
Turbine stop valves tripped i
Main steam isolation valve closing ;
(bypassed for exercising)
Turbine inlet valves and dump valve both closed l Improper power to flow ratio normalized ratio not !
greater than 1.15 Water in main steam line float switch in drain line j
. I Reactor water level high .
not more thea + 5 ft* l
\
6.1.7 Rumback l The following conditions shall automatically initiate reback with the mode switch in " flood" or " operate" position.
Condition Set Point loss of voltage on both 480 v bus sections (loss of power to bus) j Ioss of all three recirculation breaker operation :
f; Pumps I
Reactor water level low not less than - 15 inches
- i Fuel transfer valve open .
Nuclear instruesses See Table 1 i
! i 0 Note: Reference water level sero is 4 feet above fuel.
6-7 ;
i
l Reactor water level-less than 11 feet shall cause rumback when i the mode switch is in the " flood" position. The following l l conditions shall initiate runback when the mode switch is in !
the " operate" position and may be bypassed by the mode switch in I the " flood" position.
Condition Set Point Low feedwater temperature not less than 2750F Low feedwater pressure not less than 600 psig a,fter a 15 sec
) delay-Reactor water level high 15 inches or less above normal set poin) j Dunp valve hydraulic oil not less than 1800 pressure low psig 6.1.8 Control Withdrawal Permissive System Interlocks (a) Interlocks shall prevent control rod withdrawal when two or three startup channels (1, 2, and 3) read less than 2 i counts per second. This interlock may be automatically bypassed when Log-N channel (4) reads greater than 0.001%
rated power or manually bypassed by key-locked switch.*
,(b) InterlocksshallprEicntcontrolrodwithdrawalwhen
, either operating channel (5 or 6) indicates less than 5%
on any range, and may be bypassed during startup y-switching both channels down to approximately 10" rated y power or lower ranges. ,
i, (c) Interlocks shall also prevent withdrawal of inner boiler -
rods when outer boiler rods are not fully withdrawn.
This interlock may be automatically bypassed when reactor ;
water temperature is at least 3000F and may be =amially !
- bypassed by key-locked switch
- for procritical rod opera-
! tion checks, and specially supervised tests.
i I
l 6.1'. 9 Control Room 1 .
i 6.1.9.1 Control of the reactor and most of the other plant
- systems and equipes
- cohall be centralised in the f
' control room-located in the administration building. l i
l
- S:e' Section 6.2.1 for operating requirements on bypass key switches 8 6-8 l I
I
, 1 l
4
- _ . . . __ -. -. _ . . - ~ . . - _ _ _ . .- ._ - - - - , . - , . _ . . -.
o , ,
t 6.1.9.2 The control room shall be designed and shielded to permit continuous occupation up to two hours fellowing ,
an accident releasing 1001 of noble gases, 501 of the halogens, and 1.0% of the solids in the end-of-core-life ==i== fission prooucts. l 6.1.9.3 Sufficient protective clothing, supplied air masks, ,
and portable radiation monitoring equipment shall be ;
serred in the control room and available for use in
. emergency. li i
6.1.9.4 smorgency lighting and communication equipment shall
> be provided for the control room. Two-way communica-tion between the control room and in-plant communica-tions posts and between the control room and off-site locations shall be possible. ,
6.1.9.5 In addition to those specifically described in other sections of this specification, iastruments shall be located in the control room to indicate reactor operating parameters of main stema flow, temperature and pressure; Stadrater. flow, temperature and. pressure; purification. flow; reactor water level; reactor vessel saturated steam pressure; instrument supply air pressure; turbine main-condenser vacuum and condensate l condactivityc. dump valve and turbine inlet valve j positions; and main generator electrical load.
6.1.9.6 An annunciator system shall be located in the control room to indicate off-normal plant conditions.'
l Electrical system, fire alarms, pumps, turbine system, l ventilation system, and nuclear gepa supply system I
. annunciators shall be provided. The nuclear system J- ; "
annunciator points shall include those listed in Table 2.
6.2 Reactor Safety System Operating Requirements 6.2.1 Bypasses The specifications relating to the design of reactor safety system. bypasses and the conditions under which system functions may be bypassed are contained in this section.'
6-9 4
e i
I -
,,.. - . . ., ,- ,.- . . . ,- , . . . - _ . - . ,+,,----a , , , ,, . - - - , , , -, . . . , , - ~ . - - - - . - - -
. . o l
.=
(
j IABLE 2 !
, NUCLEAR SYSTEM ANNUNCIATOR POINTS
, Superheater outlet temperature-- High Reactor feedwater temperature - Low
'Superheater outlet pressure - High ,
Reactor pressure .High Turbine 102% overspeed - Tripped Turbine trip-stop valves - Tripped
- Reactor Control - Runback -
Reactor Control - Scram Nuclear instrumentation reactor period - Short Main steam isolation valve by-pass flow - Low Main steam inclation valva - Tripped-Closed Main steam isolation valve - Loss of Powgr Superheater outlet temperature - Low Superheater outlet pressure - Low Main steam safety valves - Open Reactor recirculating pump motors bearing temperature - High' Reactor recirculating pump motors - Overload Reactor recirculating pump motor temperatures - High.
Reactor recirculating discharge valves - Loss of Power Reactor water level - High Reactor water level - Low .
Reactor feedwater temperature - High !
Reactor recirculating water temperature - Low Reactor feedwater temperature control set point - Low '
Reactor building shield pool seals - Leaking
- Reactor building air lock doors - Open Reactor building pressure .High
"' ~
-Reactor vent- temperature - High Radiation - Hi' g h,,, Isolation - Trip Reactor control rod drive motors - Loss of Power Reactor control god drive noters - Overload Reactor control. rod drive motors - Reverse Phase Reactor control rod drive seals leakage flow - High Reactor control rod clutch pcwor - Trouble -
l Esin steam isolation valve interlock switch - Out-of-Position Nuclear _ instrumentation power range flux channels 7 & 8 - High differential Nuclear instrumentation - Trouble Reactor pressure control pressure error - High - Low Nuclear instrumentation short period - Runback Trip Nuclear instrumentation channels 5 & 6 - Runback Trip
.Ecclear instrumentation channels 7,and 8 . Runback Tr'ip Reactcr control loss of feedwater - Runback Trip Reactor control ejector exhaust high activity - Isolate - Scram Reactor control loss of circulating water - Scram Trip Reactor control main steam dump failure , Scram Trip Nuclear instrumentation channels 5 and 6 - Scram Trip Nuclear instrumentation channels 7 and 8 --Scram Trip i
6-10
. . l
.. . . -; - l l l
\
Key lock switches shall be used for all bypasses listed in this ! i Section 6.2.1. !
I
~
~1. Source Indication Broass.
i This key switch may be used to bypes the rod withdraw permissive interlocks associated with nuclear channels,1, 2, 3, and 4 to permit maintenance of channel 4 when reactor ,
power is above the startup channel range, provided the miniane operability requirements for nuclear channels are otherwise satisfied. l t
I
- 2. Short Period Runback Bypass I'
This key switch may be used to bypass channel 4 period rumback in the boiling range of reactor operation,. and permit maintenance of channel 4 provided the minimum operability requirements for nuclear channels are other- j wise satisfied. l
- 3. Inner Boiler Rods Permissive Interlock Bypass l This key switch may be used only .co permit procritical rod operation checks and to permit the conduct of specially supervised tests.
- ,. c . .. .,
4r. Main Steam Isolation Valve Exercise Bypass , ,, . , , ,,_ . ,
This key switch may be used only to bypass the valve-closing scram interlock to permit exercising the main steam isolation valve. Use of this bypass shall require l direct. supervision by the Shift Supervisor if reactor is operating above 20% rated power.
- 5. Bypass Kev Control j The keys to key locked bypass switches shall normally be l kept in a locked key cabinet under the direct super- '
vision of the Shift Supervisor or higher plant manage-ment. The keys shall be captive in the switch lock when the bypass is in effect.
6-11 I
t y- y - . - . . . . - - - - _ , . , . - _ - - - .m ,,..,-- - . . , - -, .--,-,- - .r- - - , . . .w---, ----.---w- , _ _ . - .
l
.. . . - ..w ..- .7 n z ~. ,
6.2.2 geram control System Testian j i :
(a) The scram control system, exclusive of backup relay I l
.. . system, shall be tested with the installed test provisions ; ,
at least once per shift on each shift in which reactor j
. operation occurs and prior to each reactor startup. !
3 Such test shall utilise installed test switches to. verify . j design operation of each input logic circuit and associated i clutch power supply by simulation of each input signal and I combination of input signal that should cause scram action. l (b) . The scram control system backup relay system shall be functionally tested prior to each startup except that more t than one such test in-any. twenty-four hour period shall
~
i not be required. Such test shall utilise installed test ;
.' circuitry to simulate scram signals at the input to the relay drivers and verify that the relay scram contacts
, operate in accordance with design function. ! ,
(c) Redundant components of the scram control system, exclusive of backup relay systen, shall be inspected prior to each reactor startup if such inspection has not ;
been performed within the previous 1000 hours0.0116 days <br />0.278 hours <br />0.00165 weeks <br />3.805e-4 months <br /> of circuit
- , . operation. This requirement shall apply to redundant l diodes of the two input logic circuits, redundant transistors of the logic circuit emitter followers, and redundant transistors of the rectifier-control excitation circuits. Such inspection shall requiire tests as in 6.2.2 (a), or equivalent to verify design function of each component with the redundancy of the associated component temporarily disabled. i
, 6;2.3 Nuclear Instrumentatien Testina and Operatian Reauireme.?ts_,
(a) At least one startup. channel shall monitor neutron flux
.. during shutdown if the reacto'r'cantains- fueW-~* A m + ~u~... .E (b) At least two startup channels shall monitor neutron flux during reactor startup.
(c) When operating in thh startup channel range, coincident withdrawal of more than one startup chamael detector drive shall act~be permitted and control rod withdrawal shall not be permitted while a detector, is being withdrawn.
6-12 c.;;*.a1
, ,--- -, , , . - - u , . - - - - , ,, . - - - - - - - - - - - , . - -
1
- 4
. . - : 4 (d) A minimum of both operating channels (5 and 6) or one operating channel and channel 4 shall be required to monitor neutron flux in the intermediate range. During any period in which channel 4 is used to satisfy this minimum requirement, the short-period trip normally used for runback action shall be reconnected to have scram capability.
(e) A minimum of three of channels 5, 6, 7 and 8 shall be available to monitor neutron flux in the power range.
When one of these channels is removed from service for i maintenance or repairs a scram signal from that channel shall be inserted at scram control input logic circuits to maintain an effective 1 of 3 noncoincidence scram logic
? under this condition, excep that coincidence logic may be used only to permit testing required by paragraphs i and j.
f e
(f) Consistent with the requirements of 6.2.3(a) thru. 6.2.3(a) nuclear channels may be removed from se vice for mainten ..
ance, testing or repairs.
(3) When channel repairs are made which could significantly affect circuit time constants, response time checks shall be p'erformed.
(h) Detailed calibration procedures shall be performed on each channel control room chassis circuit prior to reactor startup, if such procedures have not been performed within the previous 1000 hours0.0116 days <br />0.278 hours <br />0.00165 weeks <br />3.805e-4 months <br /> of channel operation.
(i) Channel calibration checks shall be perfomed on each channel with installed function switch circuitry prior
_... g,. reactor startup. During reactor operation such checks - --
shall be repeated at least once every shift on those channels actively menitoring flux, except that if such checks be required during the conduct of particular reacto: testing under the provisions of the power operation test progratg which extend beyond the period of one shif t, the ebeeks may be postpened until the completion of the particular test.
~
(j) The set point and operation of each nuclear channel trip l circuit shall be verified previous to each reactor startup.
I During reactor operation such checks shall be repeated once
- every shift on those channels actively monitoring flux, except that if such checks be required during the conduct of particular reactor testing under the provisions of the j power operation test program, which extend beyond the I period of one shift, the checks may be postponed until the completion of the particular test.
6-13 l . l l 1 4
, . l
. - i
. e ,
I (k). Coincidence logic shall not be used in the intermediate {
range. 1 (1) In the event modules are replaced or repaired, calibra- g ,
i tion tests shall be conducted to verify channel operation k before the channel is returned to service. }
i 6.2.4 Process Instrumentation This section specifies the operating requirements for instrument ,
' channels that meter process parameters and operate the safety sysgen trip signal devices whose set points are specified in !
Sections 6.1.4, 6.1.6, 6.1.7 and 6.1.8. The provisions of this j.
section shall exclude the radiation monitoring instrumentation ;
,~
for which operating requirements are specified in Section 6.3.4.
l 6.2.4.1 All process instrumentation components which develop input signals to the reactor scram safety system shall i be tested at least each thne the system is depressurized ,
in accordance with the established preventive-meinten- l ance methods if such tests have not been . conducted within i the preceding 90 days. . Such tests, to the extent l practicable, shall' include application of calibration j input signals at the transducer of each integral signal ;
channel to verify operability, accuracy and set point I of each component-unit in the channel from point of (
input to final trip device.
6.2.4.2 All process channels applicable to safety system scram !
l l i control surveillance of plant operation at rated ,
j conditions above twenty percent power shall be tested at least once every three months. Such tests shall
'~ --
consist of insertion ofLabaulated input signals to verify operation and set point of the final trip device.
l 6.2.4.3 All process instrument channels applicable only to l safety system scram control surveillance of plant I startup operation to rated conditions below twenty
! percent power shall be tested as specified in 6.2.4.2
! prior to reactor startup, or before placing'the safety system mode switch in the position where such protec-tion is designed to operate, if such tests have not been conducted within the previous thirty days.
6-14 i
1 4 .a . s,
, . . - . . - . . ~ < . . , , .. .,-
. . ~c s* .
~'. * *
. +- *. -
l t 1 6.2.4.4 Bypassing individual process channels from the scran control system during reactor operation aka11 be permissible whca such chamral is backed up by another i chamael providtag the same safety function provided '
that: !
(a) Procedures for testing and bypassing have been approved by the Operationa Committee. ,
I (b) All bypasses for testing and maintenace shall -
be. approved and recorded in the reactor log by the Shift Supervisor..
. (c) The Shif t Supervisor witnesses the installation !
] and xmmoval of the bypass.
6.2.4.5 Bypassing individual process channels from the scr y ?
. control system during rea: tor operation shall be . - .- e .
permissible in the event such enannel is not backed up by another A -==1 performing the sama safety function provided that: .
l (a) The provisions of 6.2.4.4 (a), (b), and (c) t shall apply. l (b) During any period such a channel is bypassed, a .
licensed reactor operator is assigned primary [
responsibility to perform the function of the bypassed function and to take any required ;
safety action in the event of off-normal operating t conditions. !
~
~ ~(c) No imori than on~~e'such ' channel may be' bypassed at ,
any time. j (d) If the maintenance or testing for which the bypass l is installed cannot h accompliebed in less than eight hours, the reactor shall be brought to a
, condition War which such protection is not i
! required.
i 6.2.4.6 Bypasses required for special reactor testa may be ;
! : installed to permit such tests provided that such ;
bypasses and test procedures shall be approved by the ,
Operations Committee.
- 6-15 t
i i
t
.- .- - - . =. - - - _ _ - . . --
}
,L- Q .: & .W S.*O f k T ~
e-'W.Q ?i,$i? 5 % .h.Ill Nc $ $ @ G$.k
(? *^Y^ , ,
l I
6.2.5. In-ogre Monitorina Systes
' ^ '
ITI5~.~2.5.1 'Descriotion and Purpose -
An in-core flux monitoring system shall be installed
- in the superheater region and shall be used to deter-mine superheater flux, distribution, the effects of;
~
,specifled rod programs on flux distribution and [
dynamic coupling between superheater and boiler cores. l l
J he monitoring system shall consist' of three' vertical
~
strings of three~1on chambers each with a flux wire' ;
thimble included in the assembly. The 9 ton. chambers shall monitor neutron. flux from about 10% to 120% !
I r ratedop*ampwithratedchambersensitivityof 5m@ s per ny and rated range of 1012 to- l l
1014nv. There shall be three picosameters~ arranged to indicate relative flux level at any three chambers in any one vertical string, any one vertical plane, ;
or any one diagonal plane. The picosameters shall 1 be readable from the control room and located in ear. -
or adjacent to the control room. Seitching provisions I for chamber selection shall be located at the h e~
meters. ;
, t The detector assemblies shall be' designed to operate at rated reactor conditions. The detector and wire ,
thimble assemblies shall.be located in dumusy super-
- heater assemblies, one'near the center of the super-heater and the other two approximately diametrically
. . . . . . opposite near..the superheater peripherymThe chambers -
---e of' each string shall:be spaced about eighteen inches apart and shall be located such that the center 1 chamber is at about. core midplane. ~
6.2.5.2 Ooerstina Reautrements At least five of the nine ion chambers shall be operable when reactor power is above 50% rated power.
At least seven of the chambers shall.be operable ; ,
during testing to verify design calculations. .
i Chambers shall be calibrated by use of flux wires as required for design verification i;ests and there-after shall be calibrated in accordance with ,
established schedules to limit ind6esteendrift to less than 10% of indicated flux levels. l L 6-16 i
l
o 5
. . I f
6.3 Plant Monitorina System The plant monitoring systems include the process gaseous, liquid, and ventilation monitoring systems, the area monitoring system, and i environmental monitors.
6.3.1 Process Radiation Monitorina Systems The process monitoring systems consist of the stack monitors, air ejector nonitor, off-gas monitor, main steam line
> wonitor, 7-11guid monitors, and_4-ventilation monitors.
Monitors may temporarily be taken out of service for main-tenance calibration and repairs in accordance with the wek requirements of Tables 6 and 7 Spare parts shall be on hand to allow necessary repairs to be made promptly.
(a) Stock Monitor The stack effluent gases shall be monitored by a gaseous and a particulate monitor. Each monitor shall measure the concentration of activity in tho-stack gas by monitoring a~ representative side-stream sample of the gas. The stack gaseous monitor shall be set to isolate the gaseous vaste disposal system from the atmosphere and to isolate the reactor purification system if the instantaneous stack release limits are exceeded.
The particulate monitor shall measure stack particulate activity with a moving filter paper mechanism and there
. . .,...shall be a carbon. filter..in the particulate monitor.._.._ _ .
. sample return line to the stack. The filter shall be analyzed for halogen activity at least weekly to verify -
that the release rate of halogens is below the release rate limit. If the filter indicates an activity of above 50 percent of the average release rate for halogens, the particulate filter shall also be analyzed for halogen activity.
(b) Air Eiector Monitor-The air ejector monitor shall detect activity in the air ejector exhaust. .It shall take approximately three ofnutes for radioactivity from the reactor to appear at the air ejector monitor during full power operation.
. (c) Off-nas Monitor
. The off-gas monitor shall detect activity in the off gas discharge to the stack downstream of the holdup tanks and filter. The off-gas monitor shall initiate isolation of i the gaseous waste disposal system frcm the atmosphere 6 -l'. l
o
, 4 j
.. l l
i l
, \
l l
at an activity level no higher than that which would result in the stack instantaneous release rates being exceed ed. The set point shall be no-higher than 90%
of full range of the instrument.
(d) Main Steam Line Monitor' The main steam line monitor shall detect activity in the ,
main steam line entering the turbine building basement.
This monitor shall initiate a main steam isolation scram on abnormally high radiation levels. In no event shall the monitor be set to initiate action at greater than ten times'the normal operating background level.
(e) Liauld Monitors !
There shall be liquid monitors at the following locations:
- 1. Turbine Building Cold Sump
~
- 2. Liquid Waste Discharge Line
- 3. Two at the Liquid Waste Holdup Tanks
- 4. Outlet of the Waste Treatment Demineralizer i
- 5. Purification System'
- 6. Feedwater F,ilters
. Each monitor shall detect activity of liquid either' flowing by or in storage. The liquid waste discharge, radioactivity
--- concentration and flow rate to the environment- shall-be ---
l recorded in the control room. The liquid waste discharge i monitor shall initiate closure of the liquid waste discharge and laundry waste discharge valves if, limits are' approached.
N (f) Ventilation Monitors I (1) The exhaust monitors shall detect activity in the reactor and fuel building exhaust ducts.
l-(2) The turbine building ventilation. exhaust monitor shall have two detectors, one which shall monitor activity in the turbine building exhaust duct and one which shall monitor activity in the flash tank vault exhaust duct. .
6.3.2 Area Monitors l
l The location and number'of area monitors shall be as f611ows:
h
- 6-18 i
l
+ .
. .. . . y , : _. . O;:: ." . l.. . .
. . .. : a.. i ; wag: ,g, q2.p7:, L.3,..
. .f,.
...c...... . = . . . . .
m
. -e-w .
Locat19e _ h Reactor Bellding 3-
~
]
Fuel Handling Building 4 {
Control Room 1 )
s-Turbine Building _ 2 f I
Hot Chemical Laboratory 1 l t
Each unit shall contain an aTare bell, an alazu light, an indicating meter, and an integral detector. A remote panel i :
for reactor building monitors shall be loca'ted~outside the !
reactor building near the personnel access airlock. j At least one monitor shall be in service ne'ar the fuel storage ,
area during ~yc1 handling operations. A remote panel for the the fuel handling building monitors shall be located on the fuel handling building operating floor.- ,
1 The area monitoring systiem shall normally be in operation; }'
however, individual monitors may be taken out of service for' maintenance and repairs. Spare parts shall be on hand to. allow necessary repairs.to be'1nadenromptly.44Geon'e+afedpsiredoarea ettor is not in operation, portable radiation detection instru- I mentation shall be provided and used by personnel in the area ordinarily monitored by the inoperative instrument.
6.3.3 Environmental and Health Physics Monitors
~'
At"least two environmental film monitoring istations~ahall"be;' ri. '.
~ ~
l ' ' '
~ ~
provided for determinths the integrated'gamme dose rate to the site environs during operation at stack. release Tates of up to 0.01 curies per second. Films with a minime sensitivity of 20 mr shall be provided at each site monitoring station.- These '
i stations shall' be located at the site boundary in the areas where the river enters and leaves the exclusion area.
l Operation at stack release rates above 0.01 omo6e~per second of j noble gas shall not exceed eight hours without at least ten film ;
monitoring stations'in service which will be used to provide 4 assurance that downwind doses received by persons living near l the site do not exceed allowable limits. The film at each station i 'shall be replaced and analyzed at least once per month.
All persons leaving radiation areas likely to became contaminated l shall be mohilored for radioactivity using a radiation detection ;
probe. r a . i 4
6-19 ,
. . . . . - . _ . ,- _, 4. _, - ., .____., . , , , . , - _ . . . . . , , . , - ,
s .
{
. , . . . .. ,.,,... ,. ,. ..,, u, .
[
- . . t The health physics monitors shall meet the requirements listed in Tables 3 and 4.
TABLE 3 - AIR SAMPLING EOUIPMENT h .. .-. . -. . -
' ,Mg,' gam, J q .
Number Radiation Sensitivity Calibration .
.Typ3 of Instrument Available Detected Range Frequency.
Moving Filter -
10 uc/mi will Air Monitor 1 S ,. .y give .2 cpm build- 35 days uo>at I cfa.
Fixtd OC,S,y in See sensitivity Filter 3 counting of counting 35 days
- instruments. instruments.
TABLE 4 - PORTABLE MONITORING INSTRUMENTS
. Minimma Number Radiation Sensitivi ty - Calibrati.on Type of Instrument Available Detected Range' Frequency Low Range Cutie-Pie 5 S, y 0-2500'mr/hr 35 days Low Range Integrating
- Cutie Pie _ _. 1. S, y 0-2500 mr 35 days i
i High Range Cutie Pie 2 y 0-250 R/hr. 35 days.
' ' ~ ' -
I Survey Heter 2 OC 2000 per
- ' ~2'eg/
em min. 35 days Neutron Probe for Survey Meter 1 n 0-300 arem/hr 35~ days Portable Geiger Survey 3 S, Y 0-20 mr/hr 35 days
.GM Frisker 5 S,y 50-50,000 cpm 35 days I
i 6-20 ,'
I
_..A_ s.
i 6 . .
(
. .: . - ,i ..
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-6.4 Radioactive Usate Dissenal Svstems 6.4.1 Airborne Radioactiva Vastes !
Cases or131asting in the reactor and passio6 with the steam ;
to the asin condenser shall be removed from the condenser by l sir ejectors. These gases shall pass through approatastely l 75 feet of 6 inch bold-up pipe 'which shall provide approxi- ,
mately 3 min of holdup. hespesshallthenpassthrough baffled holdup tanks of 270 f t volume to allow approximately 12 minutes additional decay time, and subsequently pass through t a high efftelency f11ter (99.9% efficieat, .3 microm particles).
H e gases shall be compressed in holdup tanks if additional l decay time is desired. Two holdup tanks shall be provided seek i of which Whall be sized to allow storage of the reactor off-gas for approximately 11 hours1.273148e-4 days <br />0.00306 hours <br />1.818783e-5 weeks <br />4.1855e-6 months <br /> of plant operation. Off-gas any be passed directly to the stack from the holdup pipe for a short :
period of time to allow maintenance of equipment. The air !
ejector and the stack moalters aball all be la service wesen
. the holdup tanks are bypassed, and the gases shall pass through a ;
high efficiency filter (99.97. efficient, .3 micron particles) and r
- be diluted with ventilation air before being discharged.
~
i j
Ventilation air from around the reactor shall pass directly to i the inlet plenum of the induced draf t fans for the stack.
Each induced draf fan shall be designed to deliver 18,000 cis.
The stack shall be divided into two flow chaenels. The ,
minisam stack flow during operation shall be 25,000 cfs. If i the induced draft fans are lost from service for five sinutes, the l plant shall be shut down. The top of the stack shall be 107 feet above' grade lesel. In' addition, the radioactive gas l disposal systes shall have the following characteristics: L (a) Noncondensable gases shall'be removed from the turbine '
steam seals by the gland seal ==hanster. The gases shall be discharged free the exhauster to the stack plenum.
(b) All ventilation air free the reactor containment vessel and the tv:bine building shall be discharged through the i stack.
(c) All other potential sources of gaaeous radioactive westes, i
except vent 11stien air from the Hot Chemical Laboratory, shall be discharged to the stack. The concentration of i radioisotopes at the point of discharge from the Hot Chemical Laboratory to the atmosphere shall not exceed the lir.its in Column 1. Table II, Appendix 3 of 10 CFR Fart 20 when averaged over a year.
t I 6-21 9
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6.4.2 Solid Waste Solid active waste shall be collected, packaged in suitable containers,-and. shipped offsite for disposal in accordance with 10 CFR 20.
6.4.3 Liquid Waste Liquid waste shall be processed by any or all of the following:
filtration, domineralization, sedimentation, storage for decay, and dilution. Liquid waste shall be diluted with river and cooling tower blowdown water. During those periods when water 3
is discharged from the cooling tower to the waste discharge ditch the activity of the cooling tower water shall be checked weekly to verify =that it does not contain significant acitivity ,
of plant origin.
Tanks in the liquid waste system shall have the following capacities:
Tanks Approximate Gallons 4 - Low Solids Holdup .3000 em 2 - Waste Surge 500 ea 2 - Reclaimed Water 1500 ea Neutralizing Holdup 300 Neutralizing Tank 1500 High Solids Holdup 2000 Concentrated Waste Storage 2000 2 - Spent Resin Storage 6000 ea 4 - Sumps 7200 6.4.4 Operating Requirements ,
l (a) "'s annual average stack release rate of radioactive isotopes, other than particulate matter and: halogens with half-lives longer than eight days, shall not exceed S x 101 1 cm3/see times the MPC for individual isotopes and mixtures presented in Column 1, Table II, Appendix B of 10 CFR Part 20. The maximum annual average stack release rate for particulate matter and halogens with half-lives longer than eight days shall be 7x108 cm3/sec times the MPC for indivicual isotopes and mixtures presented in Column 1, Table II, Appendix B of 10 CFR Part 20. The . instantaneous release rate limit for all radioactive isotopes shall be a factor of ten times the annual average release rate.
6 22
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(b) The 11guld radioactive vastes say be released if the gross activity of plant origin in the effluent from the discharge ditch can be regulated so that it does not exceed, on the anrmal average, the values stated in Celsmus 2. Table II, Appendix 3 of 10 CTE Part 20.
A current inventory of 11guld and solid radioactive
, vastes stored en site shall be kept.
TABLI 5 - AAEA NCEITMS Area Monitors Frequency of -
Activity Location of Alars Calibration Location Detected Alars Setroint Check Esnae 3 dNes Control Room y Suilding None 35 days entrance
! and local i l
!vice back- .
All cther y Control Roce 'E * 'CY' 35 days 3 decades required and Local CIE#lW*-
acniters ubichever is greater I
t Q
6-23
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TABLE 6 - VENTILATION AND MISCELLANEOUS MONITORS
- _ __ _ _ _ _ Ventilation Monitorp Frequency of '
Activity Location of Instrument Automatic Action Calibration
_ Location Detected __ _ Indicator Recorder Setpoint Action Check In Service Reautrement Feel Fuel If-irradiated fuel is stored Building S,y Building Control in storage pool, this or Entrance Room None None 35 days storage pool area monitor or portable replacement is j required. .
Reactor S,y Fuel Control _10 times Reactor May bypass for repair or ,!
Building Building Room normal full Scram and 35 days , replacement for up to si: 1 Entrance power level or Bldg. Vent hours.
10 mr/hr which- isolation l ever is larger !
8 ^
Turbine Fuel Control -
Main Steam While out of service, Building S,y Bldg. Room or 10%'R%ch Isolation 35 days operator shall actuate main Entrance ,.arsh laddry'- .- Scram , steam isolation scram on j 1p_i_t,1gtionof t stas holdup. i Mi_ sed 1'fancous' Monitors -
Frequency:of- '
Activity Location of Instrument Calibration
__ Location Detected Indi,ggor_ Rec _ order Location _of Alarm _ _
Check,,,_ In_ Service Requirement Laundry B, y Local None Laundry Room ,
35 days If out of service, determine background outside fan room _ daily if occup h _ _ !'
l Stack S, y Local Control Control Room 35 days ntinu81 88mPl es shall be koom and fuel taken if gaseous release handling -
rate indicates that the particulate and I-131 release' rate is .1 or more of the permissible release rate when I t,3n!p,95grily _out of _ service,
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< -l TABLE 7 - PROCESS MONITORS l
Process Monitors - Gaseous ,
Frequency of [i Location of Instruments Location of Calibration
'I Recorder Alarm Check Range In Servicedeoutrement Location Detected' Indicator Air ejector y Mezz Fuel Control Control Room 35 days 3 decades. Either air' ejector or i.
discharge Building Room and at inst.- main steam monitor must be in service. I i
Main Steam y Control Control Control Room 3 months 3 decades Either main steam or . l line Room Room ejectormonitormustbeinl service.
Off-gas Fuel B611d- Mezz Floor Control Room 35 days 5 decades ' Not required if holding Discharge. Y ing Entrance Building & gases or stack gas monitor Control Rs. In service.
! Stack Gas B,y ' Fuel Build- Control Control Room 35 days 5 decades Not required if holding *
! .ing Entrance Room gases or off-gas discharge i*
i& Mezz Floor monitor in service.
Process Monitors - Liquid ,
Frequency of l Activity. Location of Instruments location of Automatic Action- Calibration i l Location Detected Indicator Recorder ~ATannn _ Setootnt and Action Check Ranne ,
3 ~
I
- Plant Dis- y Waste.Panet' -Control Waste Panel and Close discharge valve ~
-charge .. Room Control Room at one decade above 35' days 5 decar
' analysis value.
y Weste Panel Weste Panel Waste Panel Node 3 months 5 decades All other process by selection : ,
i, monitors }
j -
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, 70 OPERATDG PROCEIIRES -
This section describes those plant operating procedures and procedural safeguards which have a potential effect on safety. Operating principles ami procedures are presented for initial start-up of the plant, for normal and esorgency operation of the plant, and for the initial phase of testing. These procedures shall not modify the express requirements
.of Parts 50 and 55 of the Commission Regulations.
71 GE!GSAL OPERATIE} PRINCIPLES
- a. Operation and control of the reactor and most of the process equipment shall be centralized in the control room.
- b. There. shall be at least two operations personnel in the control j room for startup and shutdova of the plant. There shall be at least one ABC licensed operator in the control room at all times, except when the reactor is in the cold shutdown condition. The minimum shift complement during reactor operations other than cold shutdown shall consist of a Shift Supervisor and two operations personnel. An AEC licensed senior operator shall be at the plant during power and refueling operations and other operations involv-ing reactor criticality, operations which may adversely affect core '
reactivity, operation involving modifications of core components while fuel assemblies are in the reactor, and operati6ns involving 1
movement of fuel assemblies. Normally for startup and ' shutdown operations a senior operator shall be in the control room.
- c. Operators may perform certain operating functions that may effect the reactor outside of the control room, but only at the dizwetion of or-with prior knowledge of the licensed operator in the control room.
- d. Radiation monitoring by fixed or portable' instrumentation shall be performed to establish radiation levels before initial entry into i
5 radiation zones.
l e. All personnel leaving radiation zones, and all equipment being removed from such zones, shall be surveyed to an extent adequate
, for control of contamination.
i
- f. Procedures for operation of the plant equipment, including loading, starting up, maintenance, testing, shutting down, unloading and other operations involving changes in reactivity or operation with the reactor. critical, shall be in ac.cordance with detailed written ,
instructions. Written instructions pertaining to emergencies I shall be available to all personnel in convenient places. Plant I personnel shall be trained in and familiar with standard and e a - 1 i
emergency procedures which he is required to perform. )
- g. -In the event of any situation which may compromise the safety of t
continued operation, it shall be required procedure to shut the
, plant down and to take other planned emergency action to protect persons and property. i l
. l 7-1 l
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l -h. Incidents and acts having a potential detrimental er'fect on nuclear
- safety shall be investigated to prevent recurrances. Intial review shall be by the Operations Committee. Further review shall be by .
. the Safety Committee.
- i. The Plant Superintendent shall have the overall on-site responsibility "for the plant. Technical support within the plant organization shall include personnel with training and experience in the areas of reactor engineering and operation; instrumentation, chemistry and radiation protection.
7.3 PROCEDURAL SAFEGUARDS i
The following procedural safeguards have been established for the operat- i ing safety of the plant. .
7.2.1 Security Access to the restricted area shall be controlled to prevent entrance of unauthorized personnel. Visiting persons request-ing admittance shall be required to identify themselves before entering the restricted area and shall be admitted only under criteria established by plant management.
7.2.2 Detailed Operating and Emergency Instructions
- a. Written instructions for normal and emersonoy' k operation -
(The Plant Operations Manual) shall be prepared, approved by the Operations Committee, and issued prior to 'startup of the plant. The Safety Committee shall review portions
- of these instructions which involve nuclear safety and' portions which are appropriate to the Safety ConrAttee',s authorization or pl
- :nt operations.
4
. The above instructions shall be reviewed and approved by '
responsible persons on the Plant Operating staff and by appropriate respresentatives of the Company's General Office I in Minneapolis. These instructions shall conform to the j Technical Specifications. Copies of the Site Emergency Plan j shall be kept in the Control Room, Information Center and -
j the Company's General Office.
. b. The Plant Operations Manual shall include Radiation Control Procedures to cover aspects of the plant's radiation protection program. ;
I 7.2.3 Administrative Procedural Controls The following control's shall be : employed to promote safety for the plant.
- a. Training of the operating staff so that each employee is acquainted with his specifed duties and responsibilities and the action to be taken in the event of off-standard i conditions.
l 7-2 k
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' b. Training of nav personnel to a level consistent with their specific duties and responsibilities.
- c. Periodic management review for strict adherence to the normal and emergency procedures, the radiation control procedures, the operating limits and requirements for the plant, control of access to the plant, and the procedure for investigating and reporting - - =1 or unexpected
~ incidents. Bis review shall be on an interval not to exceed 6 months and is the responsibility of the Safety Committee.
) 7 2.4 operational Review Procedures Day-to-day maintenance and operating experience and plant incidents shall be reviewed periodically by the Safety Committee.
Recommendations with respect to both nuclear safety and equip-ment protection shall be made to Northern States Power Company management.
73 PREOPERATIONAL TESTIlO A program of preoperational testing shall be conducted prior to the initial operation of the plant. It shall be the purpose of this program to demonstrate that the plant has been built in accordance with specifi- i cations and is ready for initial fuel loading and startup. Bis program shall include the following tests relating to the main steam system and reactor contain= ant.
- a. The control rod system shall be tested to demonstrate that it functions
" " " properly - ~ 8uch testing shall include determination of the scram
- ~~
times, normal withdrawal and insertion times. ,
- b. The boron injection system shall be tested with water to d==anatrate that all valving and instrumentation function properly..
l c. Se reactor protection system and all associates controls and instru-1 mentation shall be tested to the ==W== extent practical at this I time. All such instrumentation shall be calibrated, when practical and it shall be established that the reactor nuclear instrumentation is responsive to source neutrons.
i l d. Se primary system shall be pressure tested and heated by means of j an auxiliary boiler to demonstrate proper thermal expansion of l components. ,
c- - e; The emergency condenser system shall be tested to demonstrate its proper operation by circulating' heated reactor water through the emergency condenser.
- f. Pressure and leak rate tests of the reactor building shall be made.
- g. The containment spray system shall be tested to demonstrate its proper operation.
7-3 .
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- h. The' reactor clesn-up domineralizer systemt shall be tested to check l its ability to maintain the specified water quality.
- 1. Se radioactive vaste disposal system shan be tested insofar as j l practical to establish that intended disposal of contamination can .
be acc eplished. l l t
l !
74 IIGTIAL CCEE lOADIEG AND CRITICAL TESTS t Detailed procedures and a description of tests to be conducted during the ;
startup program shan be prepared and submitted to the Operations Comittee for review. The procedures shall also be submitted to the Safety Casittee
> for review. Fon owing committee approvals, the Northern States Power ,
C-==y personnel shall be responsible for the safe operation of the plant during the startup program and subsequent operation. l 7 4.1 Basic Test Conditions i The l W inn and critical testing program shall begin after the !
special initial loading instrumentation and the necessary reactor equipment have been checked and found to be in a safe and +
l operable condition. ,
At the start of la=A45, the reactor vessel water level shall be at least one foot below the fuel. ,
A 6 curie Pu-Be neutron source shan be provided to yield meaningful readings on neutron sensitive chambers. The neutron population shan be monitored during the le=A4ng. ;
i
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~
rol rod scram circuit shall' he operated by at least four --- -t neutron sensitive channels with single channel coincidence whose .
chambers shall be capable of seeing neutrons or4=4==+4ne in the l fuel. :
\
. t These shall consist of:
(a) Two compensated ion chambers connected to picommneters having l !
high flux scrsa trips in the regular safety circuit.
(b) Two BF3 proportional counters connected to log count rate !
meters and period meters. mese channels shan be connected ,
I into the regular safety circuit. i '
This instnamentation shan be utilized throughout the low-level testing program until the operating source is instaned at which time the normal out of vessel instrumentation shan be used.
Between fuel loading steps, sensors vin be moved as necessary l to- accommodate the increaees in core si;* . Such moees shan be }
limited to two sensors between loading steps. Only one sensor shan be moved at a time and the instrumentation channel shan be checked for proper operation prior to movement of the next i senor. , ,
7-4
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Each' control rod, before it is encompassed by fuel, vill be checked 4 for proper functioning in all modes. ' Presence of poison in the control rod vill be verified by the . time it is encearpassed by fuel. j i
7.4.1.1 Core Inading and Test Program - General Conditions {
(a) Daring each water fill after fuel has been added the control 4 room shall be staffed by at least one Northern States Power i ,
Supervisor, a Reactor Operator,-one Physicist or Nuclear
~
Engineer, and one Allis-Chalmers Operations Engineer.
(b) he Supervisor shall be in cc-mmication with the man in charge in the reactor building whenever fuel is being moved. ,
l (c) Should an extended interruption of the loading to the initial I
~
critical configuration occur, the fuel loaded to that time I
. shall be removed.
t i
When loading is resumed, the complete procedure shall be '
repeated.
(d) Should a.short interruption of loading to the initial critical
~
configuration occur, multiplication of the previous step l taken shall be verified before loading is continued. !
l i (e) During the low power test program foils may be irradiated i l
to determine the power distribution for various core and I
- control rod configurations. l
_7 4.2 Boiler Slab Core Ioading and Criticality ,j l The number and configuration of the boiler elements for this minimum ,
- mass criticality test are predicted by calculations and extrapola- l tion of Pathfinder critical facility slab experiments.
l
- i With the core drained, the boiler boxes and control rods shall be loaded into the vessel. Nuclear instrumentation for information l and protection shall be operable in accordance with 7 4.1. A
- 6-curie Pu-Be neutron source shall be located near the periphery :
of the superheater region. The boiler fuel elements shall be I loaded into the dry core to fona a slab array. The neutron i population shall be monitored during the loading.
l l S e reactor vessel head shall be installed and a control rod worth about4%Ak/kshallbecocked. Se neutron multiplication shall be monitored as water is added to the vessel. At specified levels, 4
vater addition shall be stopped and count rates taken. Se =mvimm ,
i rate of flow shall not exceed that of both seal water injection ,p. !
pumps (ratedat30gpaeach). t i ' The calculated' shutdown margin of this minimum critical mass assembly with all control rods full"in shall be about 13% Ak/h.
The slab core loading and water fill shall be perfonned as if l
7-5 ,
- criticality is expected at any point even though the fully nc,deratei assembly is expected to be at least 9%Ak/k suberitical with a rod cocked (4%Ak/k).
Criticality shall not be achieved any time during the fuel loading or water fill. If criticality is predicted at any time, the core region shall be drained and fuel removed to reduce reactivity.
Criticality shall be approached by the withdrawal of boiler control rods. If criticality is not ambieved, the core shall be drained andasingle2.2W/oboilerelementaddedtothbslab. Se core !
shall again be water filled and criticality approached by control rod withdrawal. Dese steps shall be repeated adding a single f' 2.2 v/o boiler element at each step until criticality is achieved.
After initial criticality is achieved, the core shall be drained, and the test repeated adding additional boiler fuel elements to determine the incremental increase in core reactivity associated with additional single boiler elements.
7 4.3 Boiler Fuel Core L W i = Criticality With the core drained, poison shims worth appr^t=ly 4%Ak/k shall be inserted. Boiler elements shall be added to complete the boiler core in a manner similar to that outlined in Section 7 4.2.~ Re neutron population shall be monitored during this process.
Se calculated shutdown margin of this assembly with all rods and 4% Ak/k in poison sh4== inserted. chall be approvinately 10% Ak/k.
Criticality shall not be achieved at any time during fuel addition ' ~ ~ ~
or water fill. If criticality is predicted at any time'during ~ 'i vater fill, the core region shall be drained and additional poison shims added to reduce reactivity.
Criticality shall only be achieved by the withdrawal of control rods.
After the reactor is critical at a low power, various reactivity date shall be taken and the shutdown margin determined. i 1
75 POWER OPER MION TEST PROGRAM The power operation test program, consisting of three phases, shall j commence only after.the initial loading and critical test pmgram has been completed and the results of this program found to be satisfactory and shall include at least the following tests:
I 751 Phase ik 200 KW (th) or Less Phase 1 experiments shall be performed (not necessarily in the order listed) at low power to, establish the reference core. Critical m
control rod configurations, core power distributions, and reactivity' l
coefficients shall be measured. Rose experimental results shall be compared with calculations. D ese measurements and comparisons shall provide verification of the shutdown margin and the validity of the i analytical model used for the design calculations of Pathfinder.
7-6 l
I .
- - _ _ . ~ _ _ _ . - _ ._
- 6 .
. . l t 7 51,1 Superheater Fuel LWing and Full Core Criticality The superliester fuel ahall be loaded and the neutron population monitored during this loading.
De reactor vessel head shall be installed and control rods vot'th ,
about4%4k/kcocked. The neutron multiplication shall be monitored as water is added to the vessel. At specified levels, water addition i shall be stopped and counts rates taken. The mart === rate of- water additiond. shall not exceed 60 spa. The superheater steam passages remain voided (most reactive configuration) during water addition. .
The calculated shutdown margin of this assembly with all rods and
- poisonshimsinsertedshallbeapproximately10%Ak/k. The super- ,
heater fuel loading and water fill shall be performed as though.
criticality were expected at any point, wven though the moderated assemblyisexpectedtobeatleast6%4k/k'ubcriticalwiththe s
cockedrodpattern(4%Ak/k).
Criticality shall be achieved by the withdrawal of boiler control rods.
7 5 1.2 Establishment of the Reference Core The reference core shall be defined as that, full corew'hich is sub-critical by at least .003 with the most reactive control rod with-drawn and the superheater voided.
Poison shims shall be removed incrementally until this criterion is.
met, thereby establishing the reference core. If the most-reactive-
~~
rod-withdrawn criter$on is met with substantial' shutdown margin,
~
56sitid shiin' boiler fuel ~ elements (3 2 w/o U-235- enrichment) shaII.v
-~
be available to increase the core excess reactivity for.c.,p 4,ional
'E'
- purposes.-
7 5 1 3 Reference Core Cold Flooding Coefficient i '
The superheater steam passage shall be flooded with the reactor shutdown. Se reactor shall be brought to criticality and the reactivity defect associated with the flooding the steam passages shall be evaluated by means of a calibrated control rod.
7 5 1.4 Fitz Maps At various poison concentrations, flux vires shall be loaded into- ;
the superheater and boiler. The boron concentrated shall be varied-e to provida control rod configurations typical of those to be' ,
i encountered during operation from reactor start-up to the control-
,.,g rods-ftall-out condition. The data shall be analyzed for superheater-boiler power' sharing, various power, shapes, and nuclear instrument-
" ' ~ ' " ' ' " ~ - - ~ - -
sti6n calibration. ,
Boron resioval shall be checked by returning to the boron-free- ,
condition and comparing the critical rod heightf with that measured -
during the establishment of the reference core. "
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7 515 Insertion of Reactor Source anii Refueling Test Preparatory to further testing, the antimony-beryllium reactor i source shall-be loaded in the superheater and the Pu-Be source ---
removed. Se reactor vessel shall be ccampletely filled, 4= 1'dia- ;
the superheater steaua passages, during this operation. Reference l core criticality shall be repeated to correlate the regular r l instruisentation indication to known conditions. -
l l
Tests shall be run to determine the worth of 2.2 and 3 2% boiler elements in various locations and configurations. Sabeequent refueling operations shall be consistent with these tests so that '
i refueling vill not be an unreviewed safety question.
7 51.6 Cold Core Pressurization he system shall be gr*=11y pressurized. Calibrated boiler control rods shall be used to determine the reactivity @ ====
resulting from the pressurization.
7 517 Tes:perature Coefficient Se temperature coefficient shall be measured'from ambient to about h20F. The reactor may be pressurized with nitrogen gas -
to prevent boiling during this test. Se reactor shall be taken critical and a slow heating rate established with the startup heater.
7 51.6 Hot Ccre Flooding Coefficient h e reactivity change associated with f1 w ing the superheater ,
steam passages with approximately 440F vater shall be determined
- by means of calibrated control rods.
7.5 2 Phase II, 5 W(th) or Iess he 6bjective of the Phase II testing shall be to raise the reactor power level in a safe manner from essentially zero power to a level at which the onset of boiling in the flooded superheater is expected, ani tc deterutne the superheater radiative cooling abnity. .
1 i
7 5 2.1. Power Calibration I he power level shall be increased to 5 W(th), and the n=-laae instrumentatica calibrated with the superheater flardad.
7 5 2.2 Superheater Radiactive Cooling Ability De purpose of this test shall be to determine the ability of the
- superheater fuel elements to dissipate reactor decay beat under the j q no-steam-flow condition by thermal radiation and conduction to the ;
j superheater moderator water.
l
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With the superheater drained, the steam line isolation valves closed, and the boiler water temperature near its operating point, reactor power shall be slowly increased in steps. At each step the superheater fuel temperature shall be measured by thermocouples on special instrumented assemblies. Se test shall be teminated at the power level at which superheater surface measurements are approaching a pre-determined limit of 1270 F.
7 5 2 3 Steam Flow to Superheater With the reactor shut down, steam flow to the condenser shall be established by opening the bypEssisolation valve. Superheater I
' fuel and bulk steam temperatures and boiler coolant temperatureu and pressure shall be monitored during initial steam flow. Reactor power ahall be increased to 5 MW(th) with the superheater power fraction' suppressed by means of the rod program.
7 5 2.4 Reactor System Tests The performance of the emergency condenser, other reactor protec-tion, and appropriate process systems shall be evaluated at this power level.
753 Phase III, Full Power or IAss De ob,jective of this phase shall be to reach full power in a safe manner. During Phase III, power shall be increased in about five steps; starting from some power near 5 MW(th) and going to full i power. Bere shall be approximately a 20%, power increment between J
._. _ . steps. Se following tests shall be perfomed at ,each step.
7531. Power Calibration At each power level heat balance calculations shall be performed ,
to. calibrate the nuclear instrumentation at steady-state conditions.
7532 Radiation Testing s
At three power levels radiation data shall be taken to verify i
shielding calculations.
! 7 5 343 superheater steam-operation ;
~
Infomation on the relation between steam flow and temperature
- and superheater fuel temperature shall be generated.. The super.
heater perfomance evaluated in Phase III at each power step shall
! be a continuation of Phase II testing. ;
7534 Fluid Dynamics Effects
- i The reactor response associated with changes in each of the {
! following variables shall be detemined: feedwater temperature, i '
feedwater flow, recirculation flow, and reactor pressure.
! 7-9 i
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l 7 5 3 5 Response to Runback ,
At each power level the response of the reactor to rod runback shall be determined. l 7 5 3 6 Transfer Function !
Transfer function measurements shall be made at various power levels l m h ansasure reactor stability margin. The oscillator rod shall be cidibrated at several angular positions. The =av1=mn peak-to-peak ok,gg/k,ggshallbelessthan10 cents. Measurements shall be made j of the zero power transfer function. The frequency range investigated ;
- shallextendfromabout0.01 cycles /seetoabout12 cycles /sec. i After extrapolation of the stability margin to full power, the -
oscillator rod shall be removed from the core. ,
l i
7 5 3 7 Xenon Reactivity t The reactivity associated with transient xenon shall be determined at several power levels by operation until near-equilibrun poisoning is reached--then decreasing the. reactor power substantially and following the resultant reactivity changes with rod movement.
~
7 5 3 8 water Inve' calibration .
Two-phase ? face level shall be determined at each power. level by means -vessel detectors and correlated to the indicated water level. A calibrating curve of two-phase interface versus power level shall be established.
7 5 3 9 Steam Dryer Efficiency At each power level samples of steam, before and after passing
- through the steam dryer, shall be taken and its moisture content detezimined. The efficiency of the steam dryer for various steam rates shall be determined. ,
7.6 NTMAL OPERATION l 7 6.1. General Detailed operating procedures for each nozinal mode of plant operation shall be prepared prior to operation. The following l is an outline of the principal normal operation procedures having
- a potential effect on the safe operation of the plant.
I 7 6.2 Cold Start-Up After Extended Shutdown ,
r ,
l A cold start-up shall occur each time the reactor is returned to ;
'}
service following an extended shutdown. The procedure for a !
nozinal cold start-up involving turbine operation shall be as follows:
7-10
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. s s .
- a. A start-up check list shan be followed prior to be=4n=4n- ,
the actual start-up so that applicable equipment and systems i ug}shan be in condition for start-up. Containment vessel j integrity provisions shall be in effect. ;
- b. Each control rod shan be exercised and scremmed as a check i of the control rod system and the reactor safety system. A coupling verification check shall be included? prior to'or j
- a. during start-up, if the control rods have been unlatched '
during the outage.
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- c. Se start-up check list shan be reviewed and approved by the Shift Supervisor prior to start-up. .
- d. Tests shan be performed to verify that the start-up channel ccoc. rate is at least two counts per second due to neutrons. Se Operations Committee shan review the start-up eha-1 count rate and determine when these are to be performed.'
- e. Se reactor shall be brought critical by control rod withdrawal fonoving a prescribed withdrawal pattern.
- f. Se power shan be adjusted once criticality is reached to maintain a reactor vessel tesqperature rise rate not to exceed 200 F per hour.
j g. The turbine shaft sealing system nFall be placed in service using steam from auxiliary supplies. ,
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' .: h. Se condenser shan be evacuat.ed and.the air ejector win be. . . _ . - . . . .
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) 1. Turbine heating shall be started during this operation j sequence. After turbine heating is cosipleted, and the reactor i j reaches rated pressure, the turbine shall be geh11y brought I up to speed.
J. The mode of turbine control shall be transferred to the reactor I fpressure'combrol system.
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!- k. .' he control rods shan be adjusted to provide the desired power i level.
7 6 5 Hot Startup '
i Whenever the plant has been shutdown for a period of time with the l reactor vessel and ==114 aries . -inina pressurized, a hot startup I i
i passedure shan be followed to return the plant to service., his
- procedure vi n be essentially independent of the cause of shutdown assusing that the cause is recognized and any non-standard conditions j ha've been corrected. The reactor instrumentation shan be reset and -
l downscaled and a hot startup check list shall be ecsqpleted prior to ;
- tlie withdrawal of control rods. Se start-up shan then proceed in l I accordance with Paragraphs (d) through (k) of 7 6.2 of the normal ! j cold start-up procedure outlined above. l i l l 7-n l l
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7 6.4 Normal Power Operation During nomal power operation, the pressure control system ahall maintain the reactor pressure at its rated value by cperating the turbine admission valves. Se turbine-generator load shan be established by the control rod positions. Be principal function of the operating personnel during this period shan be as follows:
- a. Se maintenance of a continuous watch in the control roca for prompt attention to any annunciated alams.
- b. Se adjustment of the control rod pattern to accesusodate changes in reactivity and to maintain the desired power distribution,
- c. Se evaluation of abnomal conditions and the initiation of l corrective action as required. l l
7 6.5 Extended Shutdown An extended shutdown vin be accomplished as fonovs: s
- a. Reactor power shan be reduced by manipulation of, the control rods, and the main generator load shall be decreased. Se turbine-generator win be separated from the system.
- b. Ail control rods shall be inserted. .
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- c. Se removal of reactor decay heat and the reduction.of reactor pressure shall be accomplished by controlling reactor steam flow.
-ihte rate of cooling,of the reactor iressel shan not be' anoved
'T)%td exceed 200 F per hour.
- d. Be reactor shutdown cooling system shall be placed in opera-tion after the superheater has been flooded. This system vill I
complete the cooling of the reactor water.
- e. A minimum of one start-up channel and one power range channal shall be left in operation. All instrumentation pertaining to ,
control of activity release shan be left in operation.
7.6.6 Short Duration Shutdown A shutdown of short duration may be accceplished with maintaining system pressure. The turbine-generator shall be unloaded and separated from the system. Reactor decay heat removed shan be ,
accommodated by system losses or bypassing steam to the amin Condenser.
7 7 REFUELIlOG OPERMION ,
' The refuling operation shan be conducted in accordance with the fonoving basic principles:
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- s. Wr'itten procedures shall be available prior to refueling.
- b. The insertion and removal of fuel assemblies shan be done through the top of the reactor. vessel after opening reactor vessel head closures as appropriate. Water akselding shall be provided by flood-ing the reactor vessel and the' shield pool. Fuel bundles ahall be handled by means of a handling tool, transfer carriage, and crane.
Fuel movement shall fonow the fonowing sequence for each fuel assembly replaced:
- 1. Removal of selected assemblies from core and transfer to spent fuel storage.
- 2. Reshuffling of remaining assemblies in core as desired.
3 Insertion of new assemblies in the vacant positions.
- 4. Partial core crit.icality tests may be perfomed to establish the core reactivity condition.
Shutdown margin checks shan be as described in 5 9 2. Assembly replacement shan proceed as described above until the desirste1 number of fuel assemblies have.\been changed.
- c. Attleast two start-up nuclear channels shall be in service and measuring neutron flux during an refueling operations.
d.~ Written procedures shall be used for core alterations which are known to increase reactivity. Communications between the control room and the loading are.a shall exist during an core alteratim s.
- e. The liquid poison system shall be available and ready for use.
- f. Containment integrity provisions shan be in effect during actual refueling operations except for the fuel transfer valve.
- g. Unirradiated fuel shan normany be stored in air in a new fuel storage area.
- h. Irradiated fuel shall be stored as described in 4 9 9 (c).
- 1. The minimum refueling crew during refueling operations shall be four men. There shall be a licensed operator in the control room at all .'.
times, and the Qualified Supervisor shan be in charge. ~
7 8 MAIlffENANCE The following basic principles shall guide the maintenance program at the plant-1 7 8.l* Damaged or defective equipment shall be repaired or replaced.
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7.8.2 Maintenance check lists shall be used wherever practical to assure.
that equipment is included in the systematic preventive maintenance program and to guard against error damage in carrying out the maintenance effort.
7 8 3 A system of equipment history records shall be kept in which will l be recorded the extent of'end type of repair, the regular preventive l maintenance actions, as well as any non-routine maintenance which is l l required.
i 7 8.4 The preventive maintenance program shall include a' schedule _for-exercising of normally idle components. a l.
7 8.5 Instrumentation and control syst.eas, especially the neutron power !
level instrumentation and the reactor safety system, can be tested !
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periodically with the plant in operation, and certain portions of the systems can be replaced'with spare units while the plant is in operation should it be necessary.
7 8.6 Radiological protection practices shall be hbserved in maintenance activities.
7.8 7 Control rods shall be inspected periodic' ally to detezzine their applicability for continued operation. P = @ fuel.shall be removed so that the reactor is more suberitical with that rod out than it i was prior to its removal for inspection. I 7 8.8 It shall be permissible to remove a control ro4 drive from the core
- when the reactor is in the cold xenon-free condition. The core rshutdown margin of 0 3%6k ff/k,ff with the strongest rod out of the core and the superheater voided, shall be met. The. rod power switch shall be locked in the off position and all associated equipment properly tagged. A spare control rod drive mechanism shall be used to replace the removed drive .iamediately upon removal of the , .
- defective drive.
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8.0 RESEAIlCH AND DEVELOPMENT PROGRAM i
During the initial start-up and during subsequent operation various R and D pro 6 rams will be in progress. The. Technical Specification limitations will apply with the following exceptions:
Four special fuel assemblies and the oscillator rod will replace one of the inner boiler control rods, quad box, four regular fuel assemblies and the normal control rod drive.
During the experiments with the slab array the maximum possible reactivity addition rateshallbeashighas40 cents /sec. During experiments with the boiler only the maximum possible reactivity addition rate shall be as high as 10 cents /sec.
While building the slab array and during part of " boiler only" experiments, two of the quad boxes may be removed to accommodate an in-vessel chamber which is part of the core monitoring system.
Up to 5 superheater fuel assemblies instrumented with thermocouples shall be in the core to determine relative metal temperatures.
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