ML20140G769
| ML20140G769 | |
| Person / Time | |
|---|---|
| Site: | 05000231 |
| Issue date: | 03/04/1969 |
| From: | Morris P US ATOMIC ENERGY COMMISSION (AEC) |
| To: | |
| Shared Package | |
| ML20140G249 | List:
|
| References | |
| FOIA-97-34 DR-15, NUDOCS 9705120135 | |
| Download: ML20140G769 (74) | |
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UNITED STATES
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ATOMIC ENERGY COMMISSION WASHINGTON. D.C. 20545
%.:e GENERAL ELEC7EIC COMPANY
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s , AND l SOUTHWEST ATOMIC ENERGY ASSOCIATES DOCKET NO. 50-231 1
PROVISIONAL OPERATING LICENSE License No. DR-15 i
The Atomic Energy Connaission having found that;
- a. The application for provisional operating license, as amended (Amendment Nos. 5, 6, 7, 8, 9, 10, 11, 12, 13, 14, 15, 16, 17, 18, 19, 20, 21, 22, and 23 to the license application, dated 1 July 25, 1967, December 5,1967, December T(,1%7, January 18, !
1968, February 14, 1968, February 29, 1968, February 29, 1968, l March 2, 1968, March 11, 1968, April 25, 1968, May 17, 1968, 1 May 24,1%8, July 18,1968, August 6,1%8, September 25, 1968, September 27, 1968, September 30, 1968, October 1, 1968, and October 9,1968, respectively) complies with the requirements of the Atomic Energy Act of 1954 as amended (the "Act") and the Commission's regulations set forth in Title 10, Chapter 1, CFR;
- b. The facility has been constructed in accordance with the application, as amended, and the provisions of Provisional Construction Permit No. CPPR-17;
- c. There are involved features, character.tp. tics, and components !
as to which it in desirable to obtain actual operating experience I before the issuance of an operating license for the full term requested in the application;
- d. There is reasonable assurance (i) that the facility can be operated at power levels up to a maximum of 1 megawatt thermal l in accordance with this license without endangering the health '
and safety of the public, and (ii) that such activities vill be I conducted in compliance with the rules and regulations of the Commission;
- e. The applicants are technically and financially qualified to engage in the activities authorized by this license, in accordance with the !
rules and regulations of the Commission; 9705120135 970505 PDR FOIA VARADY97-34 PDR
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The applicants have furnished proof of financial protection to satisTy the requirements of 10 CFR, Part 140; g.
The issuance of this license vill not be inimical to the common defense and security or to the health and safety of the puBlic;
. Provisional Operating License No. DR-15 is hereby issued to the General Electric Company (General Electric) and Southwest Atomic Energy Associates (SAEA) as follows:
- 1. This license applies to the plutonia-urania-fueled, fast-spectrum, sodium-cooled experimental reactor (the " facility").
The facility, known as the Southwest Experimental Fast Oxide Beactor (SEFOR), is located in Cove Creek Tbwnship, Washington County, Arkansas, approximately 19 miles southwest of Fayetteville, Arkansas, and is described in license application Amendment No. 5
" Facility Description and Safety Analysis Report," Volumes I and II, as supplemented by Amendment Nos. 6, 7, 8, 9, lo,11,12, 13, 14, 15, 16, 17, 18, 19, 20, 21, 22, and 23. Said 'Tacility Description and Safety Analysis Report" in Amendment No. 5, as supplemented and amended, is hereinafter referred to as the Safety Analysis Report.
- 2. Pursuant to Section 104b of the Act and 10 CFR Part 50,
" Licensing of Production and Utilization Facilities", and Subject to the conditions and requirements incorporated herein, the Atomic Energy Commission (the " Commission") hereby licenses:
A. General Electric Company and Southwest Atomic Energy Associates to acquire and possess legal title to the facility as a utilization facility; and B. General Electric Company and Southwest Atomic Energy Associates, 91th General Electric acting for itself and for SAEA:
- 1. To possess, use, and operate the reactor as a utilization facility at power levels up to one megawatt thermal;
- 2. To receive, possess, and use at any one time up to 550 kilograms of plutonium as contained in SEFOR fuel elemer.ts, foils and sources, and 0 5 kilogram of uranium 235 as foils, check sources or in instruments in connection with operation of the facility pursuant to the Act and 10 CFR Part 70, "Special Nuclear Material".
l 3,. Yo receive, possess, and use s curies of cobalt a0 l as sealed calibration sources and 2500 kilograms of
., natural and/or depleted uranium as contained in SEFOR fuel slements, foils and instrument check sources in l connection with operation of the facility pursuant to '
10 CFR Part 30, " Rules of General Applicability to Licensing of By-Product Material" and Part 40, " Licensing of Source Material". '
I 4.
l To possess, but not to separate, such by-product and I special nuclear material as may be produced by operation of the facility, pursuant to the Act and 10 CFR Parts 30 and 70.
- 3. This license shall be deemed to contain and is subject to t he conditions specified in the following Commission regulations in 10 CFR Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Sections 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; and is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified below:
A. Maximum Power Level General Electric is authorized to operate the facility at steady state power levels up to a maximum of 1 megawatt thermal.
B. Technical Specifications The Technical Specifications contained in Appendix A
' attached hereto are hereby incorporated in this license.
General Electric shall operate the facility in accordance ,
with the Technical Specifications and may make changes therein only when authorized by the Commission in accordance with the provisions of Section 50.59 of 10 CFR Part 50.
C. Reports In addition to reports otherwise required under this license and applicable regulations:
(1) General Electric shall inform the Commission of any incident or condition relating to the operation of the facility which prevented or could have prevented a nuclear system from performing its safety functions
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as described in the Technical Specifications. For
- each such occurrence, General Electric shall promptly
- - notify by telephone or telegraph the Director of the appropriate Atomic Energy Commission Regional Compiiance Office listed in Appendix D of 10 CFR Part 20 and shall submit within ten (10) days a ,
j report in' writing to the Director, Division'of Reactor Licensing (" Director, DRL") with a copy to the Division of Compliance. l (2)
General Electric shall report to the Director, DRL, in writing within thirty (30) days of its observed occurrence any substantial variance disclosed by operation of the facility from performance specifications contained in the Safety Analysis Report or the Technical Specifications.
(3) General Electric shall report to the Director, DRL in writing within thirty (30) days of its occurrence any significant changes in transient or accident analysis as described in the Safety Analysis Report.
(4) As soon as possible after the completion of six months of operation of the facility (calculated from the date of initial criticality), General Electric shall begin submitting reports in writing in accordance with the requirements of the Technical Specifications.
D. Records General Electric shall keep facility operating records in accordance with the requirements of the Technical Specifications.
4 This license is effective as of the date of issuance and shall expire eighteen (18) months from said date, unless extended for good cause shown, or upon the earlier issuance of a superseding operating license.
FOR THE ATOMIC ENERGY COMMISSION N N N y b Yy Peter A. Morris Peter A. Morris, Director Division of Reactor Licensing
Attachment:
Appendix A, Technical Specifications Date of Issuance: MAR 4 1989
, joiietud w/1.tr CM 'N )
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, [Fgulatory file APPENDIX A
.T_9_
PROVISIONAL OPERATING LICENSE DR-15 TECHNICAL SPECIFICATIONS FOR THE j SOUTHWEST EXPERIMENTAL FAST OXIDE REACTOR GENERAL ELECTRIC COMPANY m
SOUTHWEST ATOMIC ENERGY ASSOCIATES DOCKET NO. 50-231 i
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RETURN TO RE0ll.A"0RY CE ROOM 016
! RESUL\ TORY DOCKET FMCDPY l
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TECHNICAL SPECIFICATIONS FOR THE SOUTHWEST EXPERIMENTAL FAST OXIDE REACTOR Table of Contents s .
INTRODUCTION Section 1 - DEFINITIONS 1-1 Section 2 - LIMITS 2.1 Safety Limits 2.1 2.2 Limiting Safety System Settings 2.2-1 Section 3 - LIMITING CONDITIONS FOR OPERATION 3.1 Reactor Safety System 3.1-1 3.2 Reactor Control Systea 3.2-1 3.3 Reactor Core 3.3-1 3.4 Sodium Coolant System 3.4-1 3.5 Containment System 3.5-1 3.6 Electrical Systems 3.6-1 3.7 Radioactive Waste Control System 3.7-1 3.8 Irradiated Fuel Storage Tank 3.8-1 3.9 Operations Conducted-with Reactor Vessel 3.9-1 Head Removed 3.10 Approach to Power 3.10-1 >
3.11 Oscillator Tests 3.11-1 3.12 Excursion Tests 3.12-1 Section 4 - SURVEILLANCE REQUIREMENTS 4.1 Reactor Safety System 4.1-1 4.2 Reactor Control System 4.2-1 4.3 Reactor Fuel Rods 4.3-1 4.4 Reactor Coolant System 4.4-1 4.5 Containment System 4.5-1 4.6 Emergency Electrical Power System 4.6-1 4.7 Piping System Snubbers 4.7-1 4.8 Environment 4.8-1 4.9 Unexplained Reactor Behavior 4.9-1 Section 5 - DESIGN FEATURES 5.1-1
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l Table of Centents (bontinued)
- Page
'section 6 - ADMINISTRATIVE CONTROLS s .
6.1 Organization, Review and Audit 6.1-1 6.2 Plant Operating Procedures 6.2-1 6.3 Action to Be Taken in the Event of an 6*3-1 i
Abnormal Occurrence l
6.4 Action to Be Taken if a Safety Limit 6.4-1 l is Exceeded.
l 6.5 Plant Operating Records 6.5-1 6.6 Plant Operating Reports 6.6-1 l
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IN'IRODUCTION l
These technical specifications, which have been prepared in accordance with 4 ~ the requirements of 10 CFR 50.36, incorporate the significant safety limits, functional performance requirements, operating limits, administrative require- l j ments and surveillance schedules applicable to the Southwest Experimental l
Fast Oxide Reactor referred to as SEFOR.
The reactor is designed to operate at approximately 20 megawatts thermal i
(MWt) but initial operation will be limited to a power level of 1 MWt in
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order to permit initial fuel loading and preliminary testing.
These s'pecifications will be modified and supplemented at such time as the operating license is amended to authorize operation at a power level of 20 MWt.
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l Section 1 l :
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' DEFINITIONS
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1.1, Operable ,
A system or component shall be considered operable when it is capable of performing its intended function in its required manner.
1.2 Operatina A system or component shall be considered to be operating when it is performing its intended function in its required manner.
, 1.3 Reactor Conditions
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The reactor shall be considered to be Secured or Shutdown as indicated below. The reactor shall be considered operating under all conditions not covered in 1.3(A) and 1.3 (B) below. i A. Secured Secured shall mean that either or both of the following requirements ;
are satisfied:
(1) The reactor contains less than the minimum amount of fuel required to achieve criticality with sodium in the reactor.(1)
(2) At least nine reflector segments are fully lowered, the reactor coolant temperature is 300*F or higher, and the reactor operate mode switch is locked.in the " SECURED" position.
B. Shutdown Shutdown shall mean that either or both of the following requirements are satisfied.
(1) At least nine reflector segments are fully lowered, the. reactor l coolant temperature is 300*F or higher, and a senior licensed operator is in charge of operations.
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(2) At least eight reflector segments are fully lowered, the
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reactor coolant temperature is 300 F or higher, the reactor s-
. operate mode switch is in the " REFUELING" position, and a senior l
licensed operator is in charge of operations.
1.4 Reactor Shutdown Actions l
A. Rundown r
- A hydraulically driven, automatic lowering of the reflector segments one at a time.(2)
B. Scram An automatically or manually initiated lowering of all operating
. reflector segments at the maximum rate. (3)
I 1
15 Reactor Power The rate at which heat is generated within the reactor vessel.
i 1.6 Bated Flux' !
The neutron flun that corresponds.to a steady state reactor power level of 20 MWt.
17 Containment Integrity The SEFOR containment system consists of an inner and outer barrier.
Containment integrity means that all of the following conditions are satisfied: i A. The automatic containment isolation valves on the ventilation lines' through the outer barrier are operable or are closed.
B. All doors in the inner containment barrier are closed and operating, the valves in the purge lines bypassing the batch tanks are closed and
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s l locked, and all batch tanks arc either operable or have all their )
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batching valves closed.
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. C. At.least one door in each air lock through the outer containment I
- barrier is closed and both are operable.
D. h 12'3" diameter equipment door in the outer containment barrier is
! closed aad ope. rating.
? E. h atmoap' cers within the inner containment barrier contains lees I than 5 V/o oxygen.b)
F. Isakage rates through the inner and outer containment barriers.are i
equal to or less than the maximum allowable rates specified in i 1 l Section 3.
I 1.8 Abnormal Occurrence Abnormal occurrence shall be defined as any one of the following:
- (1) Any operating condition that exceeds a limiting safety system setting specified in Section 2.
(2) Any operating condition that violates a limiting condition for
- operation specified in Section 3.
l (3) Any unplanned reactor scram.
i (4) Any uncontrolled or unplanned release of radioactive material.
i 19 Written Order b ,
4 A written instruction approved and signed by the SEFOR Facility Manager or his designated representative.
1.10 Degree of Redundancy t
) Degree of redundancy is defined as "R" in the formula, R=N-M, where N is the
- total number of channels which are able, singly or in coincidence, to l-3
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, , perform a required safety action, and M is the minimum number of such I '
channels which.When tripped, v111 always perform the required safety action. For purpose of calculating R, a tripped channel shall be considered inoperable.
1.11 Protection Instrument Channel 1be arrangement of components and sensors required to generate a single l trip signal related to a plant condition requiring protective action.
J l.12 PROTECTION IDGIC SUB-CHANNEL 'Bae arrangement' of relay contacts from the gespective protection instrument channels, including the coils of the devices operated by such contacts.
1.13 PROTECTION IDGIC CHANNEL - The arrangement of contacts of the devices operated by the respective Protection Iagic Sub-Channel, and including the feeders and test devices for the trip actuators.
1.14 MODE SWI'ICH - A control switch which provides electrical contacts to cortrol the bypass relays. These contacts, together with permissive fun t; ion contacts, allow the bypass relays to be energized and thereby provide a bypass of the safety function contacts.
1.15 Channel Check - A visual inspection of the output analog signal indication V
during channel operation to determine that the channel is functioning properly.
If a surveillance function should not be performed because of an extended shutdown that function shall be performed before reactor operation is 1-4
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,, resumed, except that the channel check may be made after the applicable l channel -is operating.
I 1.16 Channel Test - Initiation of an input signal that will cause the instrument l channel to respond through the preset trip points and thus determine that 1
the trips respond correctly at the prescribed levels.
1.17 Channel Calibration - Adjustment of the channel instrumentation such that it responds in accordance with design requirements. Calibration shall be deemed to include the Channel Test as described above. ,
1.16 Gage Pressure, psig Differential pressure measured with respect to the ambient pressure 1
surrounding the named vessel.
l 1.19 FRED
'The Fast Reactivity Excursion Device used to perform excursion tests.
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1.20 Excursion Test A planned test in which reactor power is rapidly increased as a result of use of a specially built device (FRED) installed in the center drywell of the reactor.
1.21 bacillatorTest A planned test in which reactivity, main primary coolant flow rate, and main secondary coolant flow rate, are varied in a sinusoidal manner, either independently or in conjunction with each other.
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,8eferences:
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(1) SEFOR FDSAR, Volwae II, paragraph 12.2.1, item 1, page 12-3; and hble XII-2., O Power, page 12-2.
(2) SEFOR FDSAR, Volmne I, paragraph 10.2.2.2.3, page 10-5 (3) SEFOR FDSAR, Volume I, paragraph 4 5 2 7, page 4-50.
(4) SEFOR FDSAR, Volume I, paragraph 7 2.2, page 7-1.
(5) SEFOR FDSAR, Volume I, paragraph 10.3.1.2.4, page 10-23; and hble X-5, page 10-20.
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i 2.1 Safety Limits
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, _ Applicability
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.l Applies to process variables which affect the integrity of the primary l
! system.
Objective To assure the protection of the ptimary process system barriers against uncontrolled release of radioactivity.
Specification A. The maximum permissible reactor core flux shall be 110% of rated flux, except that the limit does not apply during an excursion test initiated by FRED.
B. The maximum permissible reactor core flux during an excursion test after the FRED is fired shall be 6.25 x 10 times rated flux.
C. The maximum permissible integrated energy deposited in the core during an excursion test shall not exceed the limit shown in Figure 2.1-1. The limit is exceeded when reactor conditions result 1 in a point above the limit line.
D. The reactor vessel outlet coolant temperature shall not exceed 1050*F.
E. The maximum permissible reactor vessel cover gas pressure shall be 45.5 psig.
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l-2.2 Limiting Safety System settinas l Applicability .
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These limits apply to trips which scram the reactor l [
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' Objective '
l To prevent reactor process parameters from reaching safety limits.
Specification I The Limiting Safety System Settings shall be as given in Table 2.2-1 l
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2.2-1
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SCRAM FUNCTION l
FUNCTION SAFETY SYS M SFFFINGS High Flux, I Wide 1%nge Monitor 105% of Rated Flux Lower Invel, I Beactor Sodium 3 inches below lip of operating level overflow pipe High Temperature, I 9000F Core Outlet-Upper Begion Iov Flow, I Main Primary 20% below the operating flow set point High Temperature, I 350 F for thermocouples on the Reflector Region reflector guide structure inner diameter and radial web.
I 275oF for thermocouples on the reflector guide structure, outer diameter.
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. 1 Section 3 l .
LIMITING CONDITIO8ES FOR OPERATION ;
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These requirements specify the minimum performance capability of each system l l
l required for safe reactor operation.
3.1 Reactor Safety System
( Applicability Applies to the reactor safety system.
Objective l
ib assure that process parameters will not exceed safety limits. I l
i Specification l The limiting conditions for operation shall be as specified in Tables 3.1-1 through 3.1-6.
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4 TABLE 3.1-1
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INSTRUMENTATION THAT INITIATES SCRAM ACTION i
l I II III Minimum Number Minimum Degree
}- of Operable of Conditions Permitting Function.,_ Channels Hedundancy .-- RypaD High Flux, i
Wide Range Monitor 3 2 None i
loss of Power, 2 1 1) MASER mode switch Bus 2A in NOINAL, and !
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- 2) OPERA E mode switch in ,
' SECURfD, and
- 3) all ten reflector 3 segments full down.
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i Ioss of Power, 2 1 None :
Main 2.4 KV Bus '
Iow Invel, Reactor 3 2 1) MASTER mode switch Sodium Level in N0fMAL, and '
- 2) OPERATE mode switch in muruw, and i
- 3) all ten reflector segments fulldown, or i
- 1) MASER mode switch in NORMAL, and
- 2) DPERKTE mode switch in REPUELING, and !
- 3) eight reflector segment l full down, and
- 4) reactor sodium below 40009, and
- 5) primary loop pressure below 1 psi.
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3.1-2
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hble,3.1-1 (Continued)
., I II III Minimum Rmber Minhnn Degree of Operable of Conditions Permitting
_ Function Channels Redunciancy Sypass
- 1) MASTER mode switch in N0lWAL, and 1
- 2) UPEITPE mode switch in ZERO MiD LOW POWER, and
- 3) High flux scram trip set to 500 kw or less, i
and
! 4) Se reactor vessel head is removed.
l High Temperature, 3 2 None i Core Outlet-Lower Region-l' High 'thaperature, 3 2 None
. Core Outlet-Upper Region Iow Flow, 2 1 1) MASTER mode switch in Main Primary Loop NORMAL, and
- 2) DPEITik mode switch in 1 N , and ;
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- 3) all ten reflector 1 segments full down, !
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- 1) MAS'IER mode switch in NOIMAL, and
- 2) OPERATE mode switch in ZERO AND IDW POWER ---
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1 or
- 1) MASTER mode switch in
. M , and
- 2) UPETfe mode switch in M , and !
- 3) eight reflector segment 1
- full down, and l
- 4) reactor sodium below !
400 F, and
- 5) primary loop pressure i-below 1 psi.
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MblS 3.1-1 (Continued)
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I II
_III Minimum Amber Minimum Degree of Operable of
" Function Channels Conditions Permitting ap4undancy Bypass s .
Sodium Imak, 6* 6*
and None Auxiliary 6 5 Primary Icop None Iow Invel, 2 1 Main Secaddary None Expansion 2nk Iov Imvel, 2 1 None Auxiliary Secondary Expansion Snk
, . High '3emperature, 2 1 Main Secondary None Cold Iag Iov Flow, 2 1 Main Secor.dary Same as for "Iov Flow, Io0P Main Primary Icop" t
Hidh 'fenPerature, 9 8 Reflector Region None Iov or High 9 6 None Tvesekres Reflector drive Accumulators Very High Radiation, 2 1 None Contaiment Ventilm-tion Exhaust Nitrogen Ihdiation 1 1 Monitor ** None
- monitor.
Applies only to circuits which act in coincidence with the nitrogen rediation
- Used only itt voincidence with sodium leak detectors in the auxiliary primary
! loop.
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I II III Minimum Number Minianan Degree
$ - of Operable of Conditions Permittins Function Channels Redundancy Bypass Protection Logic 3 2 None i Sub-Channel 1 (Busses.A,B,&C)
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' Protection Iogic 2 1 None Sub-Channel (Busses D, E, & F)
Protection Ingic 2 1 None Channel (Solenoid Busses i
' Dest Sections 1 & 2) I l
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, TABLE 3.1-2
., TRIP SETTINGS FOR SCRAM ACTION F L* .; ."T. ! 0:1 TRIP SETTING
'Ioss of Power, 5 380 Volts Essential Bus 2A Ioss of Power, 5 1,8 Kilovolts Main 2.4 KV Bus High Temperature, 900 F Core Outlet-Inver Region Imak, Auxiliary Primary Electrical Short Loop Low Invel, Main Secondary k 17 inchen below sodium level Expansion Tank at design power
-Iow Invel, Auxiliary Secondary 7 inches below sodium level Expansion 'Ihnk at design power High Temperature, 800 F Main Secondary Cold Iag Low Flow, Mair. Secondary g 20% less than the operating flow set point High or Low Pressure, Reflector Drive Accumlators f> 225 psi (high) 150 psi (low)
Very High Radiation, k 10 X Radiation Imvel at design Containment Ventilation power level Exhaust l
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3.1-6
' TABIE 3.1-3 t
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- ' .INSTfDGlNTATION THAT INITIATES A "BIDCK RAISING OF 'IEE REFIECTOR SEGMENTS" t
s .
I II III !
Mininnan Nanber Miniumma Degree of Operable ~ of Conditions Permitting I Funccion Channels' ReAnnA m y gpagg Iow Flux Source, 2 1 May be bypassed when either i l Bange Monitor intermsdiate range monitor l is upscale by at least :
- one decade, or when the i ;
reactor vessel head is '
removed and special startup instrumentation is installed in the center i drywell l
i Iow Flux, 3 2 May be bypassed when all l . Wide Range Monitor wide range monitor range l
switches are on or below (more sensitive) a position ~
that results in an -
indication between 50% and 100% of full scale at 1 0.2 watts. '
l
" Operate" mode bypass 2* 1* None switch in the " SECURED" position Protection Logic 2 1 None Sub-Channel Protection Ingic 2 1 None Channel
- Redundant contacts on the switch.
i l
3.1-7 l
l - _ - - - .
]
l l
TABLE 3.1-4
., TRIP SETTINGS TO BIACK REFLECTOR RAISE ACTION 1
s . 3 FUNCTION TRIP SETTLES
'Iow Flux, 5 10% of Full Scale Linear Source Range Monitor Iow Flux, >
10% Full Scale I Wide Range Monitor l
I i
l e
i 5.1-8
l J
' TABLE 3.1-5
' INSTRUMENTATION THAT INITIATES ColEAIIDENT ISOIATION ACTION I a i s .
I II III Minianan lhamber Minianan Degree of Operable of Conditions Permitting l Functton Channels Bedundancy Bypass High Radiation, 2 1 None Contairunent Vent Exhaust Very High Radiation, 2 1 None Contairnment Vent Rrhaust Protection Ingic 2 1 None Sub-Channels (Busses J, K, & L)
Protection Iogic 2 1 None Channels (Solenoid Busses A & B) l l
l l
l 3.1-9
f TABLE 3,1-6
., EIP SETTINGS FOR CONMIISENT ISOIATION ACTION s .
FO'!CTION gg gggg High Radiation, Containment I 2 X radiation level at Ventilation Exhaust normal design power level i i
Very High Radiation Invel, 2 10 X radiation level at Contairunent Ventilation normal design power level Exhaust 4
i i
j l
1 l
3.1-10
l l
. 3.2. Reactor C~ontrol System
.o Applicability -
Applies to the reactor control system.
Objective To assure proper operation of the reflector segments and the availability of adequate shutdown margin.
Specification A. At least nine reflector segments and their associated drives shall be operable.
B. 'Ibe insertion rate of any reflector segment aball be less than 1.2 inches per second.
i C. Upon initiation of a scram signal, the time required, including safety l system response time, for each operable reflector segment to move through 90% of its stroke from the fully raised position shall not exceed one second when the Master Mode Switch is in either the Nomal or Oscillator Test position and 1.1+ seconds when it is in the l Excursion 'Ibst position.
l D. Whenever outer containment integrity is intentionally breached, the the electrical power circuit to the reflector control hydraulic power l
supply shall be de-energized.
! l l
3.2-1
. ~. _ . - - . - - - . . . . - . _ . - . . . . - - . . - - . -. -.
.. .c i-l' H. . Guinea pig' fuel rods of 25.0% plutonium enrichment shall only be ,
i a
located below the six refueling ports. No guinea pig rods shall be '
~
s located under the three innermost refueling ports during steady state reactor operations-above 17.5 MWt.
I. No fuel rods shall be placed in the center drywell. '
l J. Fission chambers, experimental foils or oxide fuel samples having a i
total reactivity worth of less than 60c and containing a total of not ,
more than 0.5 Kg fissile material may be placed in the center drywell for irradiation at power levels equal to or less than 100 KWt.
Experimental foils containing less chan 10 mg of fissile material may be irradiated at reactor power levels above 100 KWt.
K. Fuel rods which have defects as defined below shall not be reinserted in the core: '
- 1. Visual observation of cladding rupture, cladding perforation or other visual defects may cast reasonable doubt on the integrity of the rods.
- 2. l Local swelling of the cladding is'in excess of 10 mils or bowing j of the rod is sufficient to prevent re-insertion of the rod into the core.
- 3. The column height of either fuel segment has increased by more l than 1/2 inch.
{
1 I
3.3-2
_m , . . _ . . _ . - . _ . . _ . , . - - -
l
- 3. 3 Reactor Core Applicability Applies to ;4 actor core loading configurations.
Objective 2 assure that core physics parameters remain within the expected range 1
a.
and that fuel rod cladding integrity is maintained. I Specification A. The reactor shutdown nargin at 350 F shall be equal to or greater than 1 $ and extrapolation of data obtained at or above 350 F shall demonstrate that the reactor would be suberitical at 300 F with one operable reflector segment raised to its most reactive position.
B. We excess reactivity available at rated power (20 MWt) shall be equal to or less than 0 5 $ when the core inlet temperature is at 7000F.
C. The reactor power coefficient of reactivity at constant inlet temperature and constant coolant flow rate shall be negative.
D. The isothermal temperature coefficient of reactivity at "zero" power shall be negative.
E. Following initial operations at a power level of 10 MWt, the reactor shall not be operated unless operating data from SEFOR demonstrate that the net non-fuel coefficient is negative and that the Doppler coefficient is negative with a magnitude equal to or greater than .
0.005.
F. We reactor shall have a phase margin of at least 30 degrees at the point where the Nyquist plot crosses the unit circle.
G. Se reactor shall have at least 600 fuel rods in the core at power levels above 10 MW. 3.3-1 I
~
3.4 Sodium coolant System
, Applicab,111ty Applien-to the main and auxiliary sodium coolant systems and the.
. irradiated fuel storage tank.
Obiective To assure reliable and adequate cooling of the core and to limit potential ;
radiological effects of the primary sodium.
Specification A. Each primary and secondary sodium coolant loop shall be operable and the sodium temperature shall be 300*F or greater.
B. The pump-around loop shall be operable.
C. The ortoc cover gas system and the argon vent vacuum pump shall be operable.
D. The fission product monitor shall be operating at power levels above 1 MWt.
E. The plugging temperature in each coolant loop and the irradiated fuel storage tank shall not exceed 425'F, and the plugging temperature be at least 25'F below the sodium. coolant temperature.
F. The cover gas pressure in the main and auxiliary secondary sodium expansion tanks shall be equal to or greater than the cover gas pressure in the reactor vessel.
G. The sodit:m leakage rate in the main IHX at normal operating pressures shall be less than 3 gal /hr.
H. The sodium leakage rate in the auxiliary IHX at normal operating pressures shall be less than 3 gal /hr.
4 I. The cover gas pressure in the reactor vessel shall not exceed 25 psig.
1
, 3.4-1
- - . - -- . - - . - - . - . - - - .. .. . . - . . - . ~ . .-
l i l
1 1
3.5 contalnment System , ,
l' i
Applicability
- l. I
\
,, Applies'to the operating status of the inner and outer containment barriers.
' Objective l To minimize and limit the inadvertent release of radioactive materials to the environs.
i Specification i
A. Containment integrity shall be maintained when either of the following -l 1
conditions exist: ,
j
- 1. The reactor in operating.
- 2. When the reactor head or the irradiated fuel storage tank cover
~
H is not in place.
1 B. The following exceptions to specification A.2, above, shall be permitted.
- 1. The equipment door in the' outer. containment may be left open while the fuel transfer cask is being used to transfer material to or from the refueling cell provided the reactor head is in place.
- 2. The marine hatch in the man access panel in the refueling cell may be used to transfer material into or out of the refueling cell during the core loading process prior to initial reactor operation. The hatch shall be closed whenever reflector segments are inserted.
- 3. If the reactor head or the irradiated fuel tank cover is not in i place and equipment failure prevents replacement, inner con- l tainment may be temporarily breached to effect repairs.
l C. The horizontal transfer port shall not be used.
D. The specified leakage rates for the outer and inner containments
! shall be as follows:
J 3.5-1 0
-n- - . . ,, - - -,
. ..-. . _ - . . -= . . . . . ._ . . _ . .. .
Outer Contautmant : ;
- 1. Lg = 1.2% of V, in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at 10 psig.
I
- 2. L = 1.4% of V in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at 10 psig.
., Inner Containment:
l
l
- 3. L' = 14.8% of V in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at 10 psig.
4 L = 16.5% of Vg in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at 10 psig.
The values to be used for containment volumes are:
1 V = 70,000 cu. ft, for the outer containment l i
i = 73,000 cu. ft. for the inner containment E. The reactor shall not be made critical if the leakage rate of the inner containment exceeds L t r if the leakage rate of the outer containment g
exceeds L .
o F. If the measured leakage rate of the outer containment exceeds L '
t o
but does not exceed L , individual penetrations shall be repaired as necessary, and the integrated leak test shall be repeated until the measured leakage is less than L .
o G. If themeasured leakage rate of the outer containment exceeds L,, the procedure described in F shall be followed. In addition, the surveillance period for those penetrations which require repairs in order to reduce the containment leakage below L g shall be shortened to two months for a period o
of one year.
H. If the measured leakage rate of the inner containment exceeds L , but t
g does not exceed fL , individual penetrations shall,be repaired as necessary, and the integrated leak test shall be repeated until the measured leakage is less than L g .
I. If the measured leakage rate of the inner containment exceeds L1 , the pro-cedure described in H shall be followed. In addition, the surveillance period for testable penetrations of the inner containment which 3.5-2
require rep-r in order to reduce the leaknge rate below L i
l shall be shortened to two months for a period of one year.
i
, J. The leakage rate for a single electrical or piping penetration i
( ,
through the outer containment shall not exceed 1/87 times L t*
o The total leakage rate for these penetrations shall not exceed 0.4 times .
K. Se total leakage rate through the group of 15 pipe tunnel penetrations which penetrate both containment barriers shall not exceed 15 times L g g.
L. Se leakage rate through each of the following five groups of components shall not exceed 0.1 times L .
- 1. Se doors of the personnel lock, e
- 2. The doors of the emergency escape lock.
- 3. Se equipment door.
- 4. Se vacuum breaker valves.
5 The reactor building ventilation valves.
M. The leakage rate through each pressure door in the inner containment shall not exceed 0.001 times L .
tg N. The leakage rate through each testable assembly in the man access panel shall not exceed 0.001 times L t*
- 0. The oxygen content of the inner containment atmosphere shall be less than 5% by volume.
P. Se freon content of the inner containment atmosphere shall be less than 1000 ppa.
Q. The water content of the inner containment Argen atmosphere shall be less than 125 ppe.
R. The dewpoint of the inner containment nitrogen atmosphere shall be less than 60"F.
S. he rate of change in temperature of the inner containment or outer containment atmospheres shall be less than 5 F per hour if the temperature change exceeds 500F.
3 5-3
l l
l l
l l 3.6 Electrical Systems I
Applicability
Applies to electrical power supply system including emergency power supplies.
Objective To assure a reliable source of power for operation of vital equipment during j reactor operation.
Specification I
A. ne 69 ky, 2.4 ky, and 480 y systems shall be energized.
B. Se 125 V de, + 26.5 V de, and I 26.5 V de systems shall be operable l with batteries for each system fully charged, and the battery chargers for each system shall be in service.
C. he main emergency diesel generator shall be capable of delivering rated power to the 480 V system within 1 minute after receipt of a start signal.
D. Se total amount of diesel fuel in the underground diesel fuel storage tank and the main emergency diesel day tank shall be at least 910 gallons.
E. The auxiliary diesel shall be operable whenever the reactor is operated above 5 MWt.
I l 3.6-1 1
i
+~ . - - , . - - - . - -..-.+wm.w -. ... . .- _ . .
3.7 Radioactive Waste Control System Applicability Applies to those components which control the collection, storage, and
' release of radioactive waste materials.
Objective To assure the capability for safe control of radioactive waste materiale and to define the limiting conditions for release of effluents from the reactor system.
i Specification ,
1 A. At least one of the three waste gas compressors shall be operable.
B. For reactor startup, at least two waste gas compressors shall be l operable.
C. The rate of discharge of radioactive effluents from the plant stack l i
shall not exceed:
- 1. Annual average release rate, except halogens and particulates with a half-life greater than 8 days:
4.0 x 10 10 C )pci/sec.
- 2. For periods less than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> in any seven consecutive days, hourly average release rate, except halogens and particulates w th half-lives greater than 8 days:
11 4.7 x 10 QC)pC1/sec.
- 3. Annual average release rate of radioactive halogens and particulates i
7 with half-lives greater than 8 days: 5.6 x 10 QC)pci/sec.
- 4. For periods less than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> in any seven consecutive days, l
l hourly average release rate of radioactive halogens and particulates with half-lives greater than 8 days: 5.6 x 10 %c)pci/sec.
C C is the concentration of radioisotope x in and must satisfy f1; where the (MPC)x are equal to the values in 10 CFR 20.
- 3.7-1
. . . _ . . . ~ . - - - . - - - - . - - -.-. -
I l
D. -
The gross instantaneous activity of the liquid effluent at the point 1
of release to the tile field shall'be equal to or less than the MPC,
, values specified in 10 CFR 20, Appendix B, Table II. If the isotopic content is unidentified, the gross activity shall be equal to or less -
than the unidentified mixture MPC,of 1 x 10~ p Ci/ml given in 10 CFR 20, Appendix B.
E.
The waste gas discharge radiation monitors shall be operating during periods of waste gas release.
F.
The stack dilution fan shall be operating during periods of waste gas release.
G.
The liquid waste radiation monitor shall be operating during periods of liquid waste release.
~
l 3.7-2
_ _. .. . . ~ - . . -. .. . -.
L 3.8 Irradiated Fuel Storage hnk Applicability
,, Applies'to parameters associated with the irradiated fuel storage tank
- whenever one or more fuel rods are stored in the tank.
'bjective O
To maintain safe conditions in the irradiated fuel storage tank.
l Specification 1 i
A. Se sodium temperature in the irradiated fuel storage tank shall be maintained between 300 F and 500 F.
B. De sodium level in the tank shall be maintained at or above the opening of the discharge duct attached to nozzle N-6.
C. Items A and B do not apply when the tank contains only new fuel rodsand/orfuelrodsthathavebeenirradiatedatreactorpowerlevels of 100 KWt or less.
D.. Re criticality factor within the tank shall be less than 0.95.
E. Se cover gas pressure in the tank shall be less than 5 psig.
F. The lower surface & the tank cover shield blocks shall not be elevated more than eight feet above the refueling cell floor.
References 1 SFFOR FDSAR, Volume 1, 9.4.4, page 9-8 and Table IX-3 2 SEFOR FDSAR, Supplement 17, Answer to Question M-6, page M-9.
l_ ,
,- ~
\
N 3.8-1 i
-. _ ... ~
, 3.9 Operations Conducted with Reactor Vessel Head Removed l Applicability '
' Applies to handling operations conducted while the reactor vessel head is remov.ed.
l Objective l
L l
To maintain safe conditions while the reactor vessel head is removed. l Specification A.
Either the main or the auxiliary coolant system shall be operable, unless the conditions specified in 1.3A(1) are satisfied and the core decay heat level is below 1 KWt.
B. The reactor Operate Mode Switch shall not be placed in the "HIGH POWER" position.
It may be placed in the "ZERO AND LOW POWER" position when the Wide Range Monitor contains a physical stop which limits the high flux scran set point to 500 KWt or less. The " SECURED" and " REFUELING" positions may be used according to Standard Operating Procedures.
C. No handling operations shall be conducted in the refueling cell when the reactor Operate Mode Switch is in the "ZERO AND LOW POWER" position.
D. Before raising the reflectors or moving fuel in the core, the in-core start-up instrumentation shall register more than six neutron counts /sec with a fuel loading equal to or greater than the first 36 fuel elements loaded into the innermost fuel assemblies (six fuel 1
elements per fuel assembly.)
9 3.9-1 l
__._m __ _ . _ . _ . _ . . __- - --
i 1
i i
E. Criticality checks are to be made between loading increments during l
. the initial loading to critical by raising all reflectors and noting o
the change in count rate on the in-core neutron detector and the Source Range Monitor (when the SRM count rate is in the useable range).
After the first two loading increments (108 fuel rods in each increment) no more than one rod in excess of one-half the number of additional rods predicted for criticality will be loaded in one increment. The number of fuel rods for criticality will be predicted by plotting curves of ""'"f"*,",*"vs.Pumassforcountratevaluesobtained from at least two detectors with all reflectors fully raised as well as fully lowered., The most conservative plot shall be used to establish loading increments.
F. The reactor vessel sodium temperature shall be less than 450*F.
3.9-2
3.10 Approach to Power
- Applicability
.g Applies to reactor power limits during the initial approach to full power fo'r Core I and Core II.
l Objective To provide a method of assuring a safe and orderly approach to full
- power.
Specification Reactor power shall be limited to 1 MWt.
.1
)
I i
1 I
i '
3.10-1 r
i
~
. - ....m.> .. _ _ . . . . . _ _ . _ . - . _ _ . .- . _ _ _ -. . . . . - . . _ . . ._. _.
! )
I 3.11 oscillator Tests '
. \
Applicability
'These limits apply to tests in which the rod oscillator mechanism is used to. vary one or more reactor parameters on a periodic basis.
!- Obiective {
l i To specify additional limits which are applicable only during oscil-l.
l l lator tests. i i-i
- Specification i A. The amplitude of reactor power. oscillation shall be equal to or ;
i less than 120% of the indicated power level.
i B. The amplitude of reactivity oscillator shall be equal to or less f
than 110 cents.
C. The amplitude of oscillation in the main primary and main secondary i
j coolant flow rate shall be equal to or less than 11000 GPM. ~
L..
l- D. The amplitude of the coolant temperature oscillation at'the vessel inlet or at the vessel outlet shall be equal to or less than -
160*F.
1-4 i 3.11-1
,, .. _ _ - ._ -- ~ ~ - -- -- --- - ~ " * ~ -
- . . , , . . . . v-.. , ,,e.-, u , -. ..---
-r. .. n. -. --- -
1 i
i 3.12 Excursion Tests .
Applicability
'These limits apply when excursion tests are conducted with the Fast s Beactivity Excursion Device (FRED).
Objective To specify additional limits which are applicable only during excursion tests.
Specification The FRED device may not be used during operations up to one MWt. Completion of the FRED is not required for such operation.
l l
1 3.12-1
[ .
I
Section 4 SURVEIIJAICE REqJIREMENTS s .
General his section specifies the minimum surveillance needed to assure that the limiting conditions for operation are met. De time intervals specified in this section shall be valid only during periods of normal reactor operation and do not apply in the event of reactor shutdown for periodo longer than the specified interval between tests. If a surveillance function should not have been performed because of an extended shutdown, that surveillance function shall be performed before reactor operation is resumed, except that channel checks may be made after the applicable system is operating. Specified intervals of three months and four months may be adjusted plus or minus two weeks, and specified intervalo of six months or longer may be adjusted plus or minus one month to accommodate normal test schedules. Following repairs or maintenance which could alter or impair the performance of a system, tests shall be performed to verify that the system is operable.
4.0-1
i I
~
l 4.1 Reactor Safety System Applicability
,, These tests-apply to instrumentation and sensors used to monitor parameters associated with reactor or radiological safety.
Objective To maintain proper calibration of instruments and sensors used in the safety system so that plant parameters do not exceed safety l limits, and to assure operability of instruments used to monitor radiological safety.
Specification A. Channels shall be tested, calibrated, and checked as indicated in Table 4.1-1.
B B. Radiation monitors shall be tested and calibrated as indicated in Table 4.1-2.
l l
4.1-1
- -.---~-
.- .- , . ~ . - . .- . . . - . . -... . . ~ . . ._ . . ~ ._-. . - - .- . . . . - . - -. -
- l. <- .
r l
l ABBREVIATIONS USED IN TABLES 4.1-1 AND' 4.1-2
. l
, 1
., ea strt up_ - Each start up of the reactor from shutdown.
j 1/d' - Once per day l 1/wk - Once per week 1/mo - Once per month
'l I
1/3 mo - Once every three (3) months 1/6 mo - Once every six (6) months l
)
l_ N/A. - Not Applicable 4
I l
i l l
.\
t l
l l i l
l I
o
- 4.1-2 l
I
- . - - .. . ~ . , -
T - T w
I I
L i
(- TABLE 4.1-1
, - MINIMUM FREQUENCIES FOR *ti: STING OF SAFETY INSTRUMENTATION s .
Channel Channel Channel Channel Check Test ~ Calibration n.--vka Source Range Monitor sa strt up 1/wk. 1/6 no Wide Range Monitor 1/d 1/wk 1/6 mo l
Undervoltage Relays N/A 1/mo 1/6 mo a) 2.4 KV Main Bus ,
b) 480 V Bus li Sodium Level Probes 1/d(1} 1/mo( ) 1/6 mo (1) Check of DC current to a) Reactor Level the probe.
b) Aux. Expansion Tank (2) Change of process level c) Main Expansion Tank to effect a Temperature Monitors 1/d 1/wk 1/6 mo I l a). Reactor Core Outlet Upper Region l
b) Reactor Core Outlet {
Lower Region '
c) Reactor Cavity.
l d) Main Secondary Cold Leg i Flow Monitors 1/d 1/wk 1/6 mo a) Main Primary b) Main Secondary 1
Pressure Switches N/A 1/3 mo 1/6 mo a) Reflector Accumulator Switch (Hi)
! b) Reflector Accumulator Switch (Lo)
- Reflector Accumulator N/A 1/3 mo 1/6 mo l Leak Detector
- Ventilation Radiation Monitor 1/d 1/wk 1/6 mo 4.1-3
. -- .- . . . . . - . . __ ~... - . - - .- .
TABLE 4.1-1 (Cont. )
e s .
Channel Channel Channel Channel Check Tent Calibration Rem. Aa Godium Icak' Detector N/A 1/ma 1./6mc
) (1.) A ply a to::t chart out. side contalnennt.
(2) Apply e test short at detector connector,-
and perform continuity test
- hnual Scram N/A 1/mo !!/A c) Right Side Button b) left Gide Button Manual Containment N/A 1/mo N/A Isolation Manual InitiaLion of N/A 1/mo N/A Block Raise Action with Operate Mode Switch in " SECURED" position Scram Protection Logic N/A 1/mo ti/A Sub-Channel Scram .Totection Logic Channel N/A 1/ma 1:/A Containment Isolution Sub-Channel
!!/A 1/mo :/A
!' Containment Isolation
(. Channel N/A 1/mo I:/A l Bloch Raine Action Dub-Charmel !!/A 1/mo !!/A I
Elech Ibice Action N/A 1/mo :'/A Channel 4*l~4 l
l.
l- . r l
l .
I TABLE 4.1.2 ,
MINIMUM FREQUENCIES FOR TESTING 0F RADIATION MONITORING INSTRUMENTATION Channel Channel Channel '
Channel Check Test Calibration Remarks !
Nitrogen Radiation Monitor 1/d 1/wk 1/6 mo Liquid Waste Radiation Monitor 1/d 1/wk 1/6 mo Test prior to ;
each release of radio- l active material.- i Waste Gas Discharge Radiation Monitors 1/d 1/wk 1/6 mo Test prior'to each release of radio-active material.
Spent Fuel Storage Monitor N/A N/A 1/6 mo Test prior to each movement of fuel to or from the irradiated fuel storage tank.
Area Radiation Monitor 1/d i/2 wk 1/6 mo Calibrate with Co-60 source Refueling Cell Radiation Monitor 1/d 1/wk 1/6 mo Calibrate with Co-60 source Cutie Pie Survey Meter Each time N//. 1/mo Calibrate with radio-used active source ,
Beta-Gamma Survey Meter Each time N/A 1/mo Calibrate with radio-used active source Alpha and Neutron Eact time N/A 1/mo Calibrate with radio-active source i
r 4.1-5
1 %
i 1
4.2' Reactor Control System I
s Applicability l
Applies to the core configuration and to the reflector control and drive system.
Obiective To assure safe control of the reactor under all operating conditions.
Specification
{
A. The core loading limits specified in paragraphs 3.3A and 3.3B shall l
j be demonstrated at least once every four months. After each core 1 i'
rearrangement, compliance with these limits shall be checked by use 1
of calibrated control segments and extrapolation to the specified temperature conditions using the best available data. In addition, the daily checks indicated in Section 4.9A shall be reviewed to
- assure that no reactivity changes have occured which would result in exceeding the limits specified in paragraphs 3.3A, 3.3B and 3.3C.
B. The requirements of Section 3.2 for the reflector segments shall be demonstrated at least quarterly.
C. Before each scheduled reactor startup, all operable reflector segments shall be rs. sed one at a time to a minimum height of ten inches, be driven down and be scrammed from a minimum height of six inches. One reflector segment shall be scrammed from the fully raised position.
A different reflector segment shall be chosen each time for the scram from full height.
D. During continuous reactor operation for periods longer than one week, each operable reflector segment shall be moved through at least 25%
of its stroke at time intervals not to exceed one week.
4.2-1
, . ... ._ . ._. . _ . . . _ - . . . - . _ . . _ _ _ . . _ . _ . . . ~ - . _ _ _ ..._ - . _ _ __. _ __.. . . _ _ - . _
L . ,
1
\
- i. 4.3: Reactor Fuel Rods l . i l
I '
'f Apolicability . l
- .s .
l l Applies to fuel rod examination made in the refueling ~ cell. 'l t
Obiective i
To assure maintenance of fuel rod cladding integrity during reactor-operation.
Specification A. Two or more guinea' pig fuel rods.which have operated at power densities.
higher than the power density of standard fuel rods nearest the center of the core shall be removed from the reactor after operation at reactor i
power levels of 15 a.nd 17.5 MWt, and shall be examined in the refueling i
cell by visual observation, dimensional checks, and gamma scans. 'The I l
maximum interval between these examinations, af ter reaching a power !
level of 15 MWt, shall be six months.
B. Before the start of the sub-prompt critical excursion tests and before the start of the prompt critical excursion tests, a minimum of.one guinea pig fuel rod and one standard fuel rod shall be examined by the methods described in "A" above.
C. After each prompt criti:al excursion test, at least one guinea pig rod and one standard rod shall be examined by the methods described in "A" above.
D. If the examination of a guinea pig rod should indicate damage as l
described in Section 3.3K, additional guinea pig rods shall be examined l
l-i to determine the extent of additional damage, if any.
i l
4.3-1 l
l
', .. , - , , , - . .2 . . . . - ... . . - . . - - . :.
I 9
4.4 Beactor Coolant System ps,licability
.o s This series of tests applies to the sodium coolant systems.
Objective l
To provide for. surveillance of sodium system conponents to assure that I
the limiting conditions for operation are met.
- Specifica
- .4 g A. The reactor safety vessel shall be leak tested at intervals not to exceed six months.
i i B. The irradiated fuel storage tank safety vessel shall be la.k tested l
l st intervals not to exceed six months while in service.
l l
C. Se auxiliary inlet check valve shall be operationally tested at least quarterly.
D. The check valve in the reactor overflow line shall be operationally l -tested at least once quarterly. i l
l E. The Marmon clamp on the auxiliary primary reactor vessel outlet dip j tube shall be leak tested at least quarterly.
li- F. The vacuum breaker valve for the reactor cover gas shall be functionally ,
tested at least quarterly. I G. Bar-type tensile specimens shall be retwved from the reactor vessel p
following the 3rd, 4th, 5th, 8th, and 10th year of reactor operation and subjected to specified tests. (1) (2)
H. The plugging temperature of the primary sodium system shall be measured daily when the plugging temperature exceeds 400 F and at intervals not
! to exceed one week when the plugging temperature is below 4000F.
1 I .
4.4-1
-- . ._ . - ~ _ . .. . . - . .- - -- - . .-
2 . .
i I. The plugging temperatures of the main secondary and auxiliary secondary sodium systems.shall be measured daily when the plugging temperaturc exceeds 400 F, weekly when the plugging temperature is between 300 F f and 400 F, and monthly when the plugging temperature is below 300 F.
1 J. She plugging temperature of the sodium in the irradiated fuel storage e
tank shall be measured weekly when the plugging temperature exceeds 400 F, and monthly when the plugging temperature is below 400 F.
J d
-1 i
i i
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i e
i 1
i 1
a E
e i
j References ,
(1) SEFOR FDSAR, Supplement 16,Section VII, pp 7-1 ff.
(2) SEFOR FDSAR, Supplement 19, Answer to Question 9, pp 67 ff.
4.4-2
l' -
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l 4.5 Containment System l
Applicability- -
i s .. ;
Applies to inner and outer containment barriers, including penetrations,-
i isolation valves and high velocity check valves. i l \
Objective l'
i L
To determine that the containment system continues to meet specifications with regard to allowable leakage and valve operation.
. Specification A. Inner and Outer Containment Leak Tests A containment leak test shall be performed annually for each barrier i
at a pressure differential of 10 psig across each containment barrier. l l l B. Outer Containment Penetration Leak Teste 1.
-)
The following penetrations shall be tested quarterly for leakage ;
at the indicated pressures:
- a. Inner and outer doors of the personnel lock and emergency
-escape lock: 10 psig and 1 psig,
- b. Equipment door: 30 psig and 1 psig, quarterly and following each closure.
c.. Vacuum breaker valves and reactor building ventilation valves:
- 30 psig and 1 psig,
- d. Piping and electrical penetrations-through the outer con-tainment: 30 psig.
- 2. Piping and electrical penetrations may be tested individually or may be manifolded together in groups and tested simultaneously.
l l
l-l 4.5-1 i
_ - - - .- -_.~ -. - ~_ - - ..- .. _ .._. _ . - - - - . - . .-
l l
(
l
, 3. Penetration leakage rates shall be calculated from the pressure i .
f decay rate, based on the free volume contained in the penetration (s)
- s .
l and piping used for the test. Test periods of one hour or longer shall be used to determine leakage rates, f
l 4. If the leakage rate-for a group of penetrations shows that leakage l through one or more penetrations may exceed the limits specified in section 3.5, the penetrations in that group shall be tested individually.
- 5. Individual penetrations shall be leak tested as specified above, whenever they are modified or repaired.
C. Inner Containment Penetration Leak Tests
- 1. Pressure doors in the inner containment shall be itak tested
- quarterly and prior to reactor startup following final door j k !
l closing by pressurizing the cavity between double seals.
l
- 2. Man-access panel glove port covers, the marine hatch,'and the neck ring assemblies shall be leak tested quarterly and prior to reactor startup following use of this equipment.
- 3. The man-access panel mounting gasket, windows, helmet, base rings, and electrical penetrations in this panel shall be tested using soap-bubble technique or its equivalent with norma' refuel-ing cell to air zone differential pressure each time they are replaced. Leaks shall be repaired without undue delay.
- 4. All other penetrations of the inner containment shall be checked as part of the annual-inner containment leak test.
(
I 4.5-2
- _ ~ _ . . . - ~ . - - . . - . . _ . . - - . ~ . . . ~ _.. - - - - - . . . . . _ . . .. .
l
. D. Isolation Valves
, 's i l- !
, All isolation vdives which are actuated by the reactor safety circuitry
! shall be checked quarterly for closing in response to manual scram or simulation of two out of three signals from any one of the following 4
sources:
i
i High radiation containment vent exhaust Very high radiation containment vent exhaust l Malfunction vent radiation acnitor i
l High pressurt containment building.
- E. High-Velocity Check Valves l
?
l High velocity check valves in the argon and nitrogen systems shall be checked annually for closing in response to a differential pres-t sure at the orifice taps corresponding to a flow rate of 2000 scfm or on manual initiation from the control panel.
F. Nitronen Cooling Refrigerant Isolation Valves The supply line solenoid valves and return line back pressure valves shall be tested semi-annually for closing in response to loss of pressure in the liquid supply lines.
G. Weste Gas Discharge Filter Pressure drop across the waste gas discharge filter shall be monitored each time gas is released from the decay tanks, and the trend shall be plotted. A decrease in pressure drop may be indicative of filter damage and shall be cause for filter inspection (and if damaged, replacement) prior to further usage.
4.5-3
4 t .
)
i
.H. Oxyaen. Water, and Freon Content of Inner Containment
- 2. The water content of the argon region of the inner containment
, shall be monitored daily.
! 3. The water content of the nitrogen region of the inner containment
- shall be monitored at least weekly.
4 -
e 0
1 4.5-4
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d
)
i -
4.6 .Baeraency Electrical Power System -
'f Applicability l . t
! s .
l These tests apply to the emergency power systems.
I objective j To assure availability of the emergency power system at all times.
I Specification 1 1
, A. Main Diesel Generator 480 V AC Supply j' 1. The diesel generator shall be started and loaded to 540 kw st monthly intervals.
I
~
- 2. Automatic emergency starting of the main diesel generator shall -
I l
l-1 be demonstrated once each month.
1
- 3. Operability of the emergency system tie breakers shall be demon-j strated semi-annually.
l l B. Auxiliary Diesel Generator i
i .The auxiliary diesel' generator shall be started and loaded to 30 kw l st monthly intervals whenever its operability is required by j Section 3.6. One hour availability shall be demonstrated.
C. Batteries i
l Periodic checks of battery performance shall be made on all three i
4
, battery stations (+125 V DC, i 26.5 V DC and +26. 5 V DC) and the a
diesel starting batteries as follows:
- 1. Measure and record daily the battery floating bus voltage, the pilot cell specific gravity reading and adjacent cell temperature.
The designated pilot cell shall be changed sach month.
- 2. Measure and record monthly the floating charge, amount of water added, and specific gravity of each cell. Measure and record the temperature of every sixth cell.
4.6-1
. a l
- 3. Inspect all electrical connections for tightness every six months.
v At the same time, subject each battery station to a heavy discharge l .
l condition. Monitor bus voltage and current as a function of time to establish that each battery station performs as expected, and check amperehour rating against draw-down. Except for the diesel starting batteries, calibrate panel voltmeters against a known standard.
l l
i 4.6-2
l l -
4.7. Pipin2 System Snubbers e
, Applicability -
Applies to all pipe snubbers in the reactor building.
Obiective To assure proper operation of the snubbers.
l Specification A. All pipe snubbers shall be checked for oil level and leakage at ,
l six-month intervals. 1 B. Samples of oil shall be placed near snubbers operating in represen-tative radiation fields. The viscosity _of oil samples shall be measured every s'ex moaths to assure that it is in the range specified l
l by the snubber nanufacturer.
C. Representative pipe snubbers in accessible regions will be exercised at a yearly interval.
D. Pipe snubbers shall be replaced r repaired if such action is required l to assure proper operation of tne scibbers. 1 I
1 i
l.
4 l
4.7-1 l
l
l
,4.8 Environment I
Applicability .
s .
Applies to the environmental survei? lance program in the vicinity of the site.
j Obj ective l
l To assure that release of radioactive material from the plant is not l
l
!- significantly affecting the level of radiation of the off-site environment.
l Specification !
The environmental surveillance program specified in References 1 and 2 i shall b2 Laplemented.
l l
References
- 1. SEFOR FDSAR, Volume 1, 9.4.4, page 9-8 and Table IX-3.
- 2. SEFOR FDSAR, Supplement 17, Answer to Question M-6, page M-9.
i i 4.8-1
. . . , . . . . - ~ . . - . - - - . _ , . . . . - ~ . - _ . - - . . . . _ . - - . . . . . . - .
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\ -
,4.91 Unexplained Reactor Behavior
~#
Applicability -
Applies to unanticipated changes in reactor process and nuclear variables during operation. ;
Obiective To assure that reactor characteristics are properly interpreted and sufficiently understood for safe operation and safe conduct of the experimental program.
l I Specification A. Lona-term Unexplained Trends In addition to the more frequently taken records, records shall be kept of observations made at least once daily of sodium flow for given pump conditions and the concurrent reactor sodium inlet and outlet temperatures, reflector segment positions and the reactor operating power. These records shall be examined daily during periods of reactor operation for short-term changes and analyzed in detail at least monthly to determine if there are any unexpected trends of significance which might indicate a change in reactor or component performance. Any unexpected trends which are observed shall be j reviewed by the Site Safety Committee. Reactor operation at a higher power level shall not take place unless identified long-ter.a trends are satisfactori)y explained, or the Site Safety Committee concludes that the long-ter= trends observed do not indicate a l deterioration of performance which could affect plant safety during the next planned period of operation. The conclusions of the Site j Safety Committee shall be documented and transmitted to the SEFOR Safety Review Committee and the General Manager, APO, immediately 4.9-1
!, l 1
, after their conclusions are reached. In the event that there is no
'# 1 satisfactory explanation of long-term trends, independent evaluation l
- by members of the Safety Review Committee shall be obtained within one raonth af ter identification of the trend. The General Manager, APC, upon advice of his technical staff, the SEFOR Site Safety Cammittee, and the Safety Review Committee shall make a determination of the future mode of operation of the reactor. These determinations j and the supporting documentation shall be transmitted within one week to the AEC.
B. Jhort-term Unexplained Reactor Behavior ~
Reactor operation shall be continuously monitoled for short-term changes..
For purposes of.this specification, short-term changes shall include but not be limited to, unexplained changes'in reactivity, primary coolant flow rate, or upper reactor vessel outlet temperature. Upon 1
observation of a short-term change that is not readily explainable and which in the operator's judgement has possible safety significance, reactor power shall be reduced at least to a level of 50% of that a which the change was observed. The SEFOR site facility manager shall be notified immediately. Reactor power shall not be increased unless the cause of the change has been determined. If the cause is not immediately apparent, the SEFOR site manager shall determine whether operation at reduced power may continue or whether the reactor should be shut down. As soon as Practical, he shall call a meeting i.
of the Site Safety Committee which shall investigate the unexplained j_ occurrence and recommend further action. If the cause of the i
occurrence is not identified or if it is determined that there is a 4.9-2
- . - - .- - - ~ . . . - _ ._ .- .
l s .
I ,
- potentini safety problen, resumption of operation at the initial power level where the dhange was observed shall not be permitted until a report has been made to the General Manager of APO and an investigation has been conducted by members of the APO technical staff. The General Manager, APO, may authorize higher power operation of SEFOR upon evaluation of the reports of his technical staff and the Site Safety Committee. Upon such authorization, a report of his decision and supporting documentation shall be forwarded to the AEC. If operation is resumed, the conclusions of the Site Safety Committee and the APO technical staff shall be documented and circulated to the SEFOR Safety Review Committee and their independent evaluation shall be obtained within one week of the resumption of operation.
l l
l 4.9-3 l
l J
l l
i l
l
'., Section 5 s
l DESIGN FEATURES l Applicability l
Applies to those features of the plant that are not covered elsewhere in these l
specifications and are applicable to physical barriers.
Objective l Tb control changes and maintain safety margins in the design and location of equipment.
Specifications The reactor facility shall be located within a restricted area. The exclusion l distance shall be at least 0.4 mile. . l l
l l
1 5.1-1
. . . - . . - . _ . - . - ~ . . - . . . . _ ~ - _ . . . . - . - . - . - - . .. _ .- . . - --
e
. Section 6
~# -
ADMINISTRATIVE CONTROLS s .
6.1 Ornanization. Review. and Audit This specification applies to the organization for management and for review and audit of facility operations. Its objective is to delineate l
responsibility for management.of the facility, to assure maintenance of a high level of staff competence, and to specify an independent pro-gram for review and audit of facility operation.
A. -Ornanization
- 1. The organization for management, operation, and audit of SEFOR facility operations shall be as given in Figure 6-1. The over-all full-time responsibility for operation of the SEFOR facility j .and compliance with these Technical Specifications shall reside I l
in the SEFOR Facility Manager, who in turn is responsible directly l to the APO General Manager. The APO General Manager reports to i
the Nuclear Energy Division General Manager. As indicated in Figure 6-1, technical engineering assistance will be provided as l
l necessary through the APO Manager of Plant Engineering and Projects.
l
- 2. The minimum functional organization required for operation of the facility shall be as follows:
l l a. An operating shift shall consist of a shift supervisor and at least three additional operators. Refueling operations may be carried out under tue supervision of a senior operator with a ,
minimum crew of two persons trained in fuel handling procedures.
- b. When the reactor is secured, a reactor operator or a senior supervisor and two additional persons trained in carrying out emergency procedures shall be at the site.
6.1-1
l
? .
l .
? )
., c. A licensed senior operator shall be in charge of startup, l approach to power, normal operation, recovery from unplanned !
or unscheduled reductions in power, shutdown, and refueling operations.
L d. Personnel requiring Part 55 licenses shall be as indicated in Figure 6-1.
l 3. Qaalifications with regard to education and operating experience for key supervisory personnel shall be as follows:
l.
l a. SE1POR Facility Manager l' B.S. in Engineering or Science or equivalent in experience.
l l_
Ten years experience in the design, construction, installation, l
! operation, development, and maintenance of nuclesr facilities.
Demonstrated detailed and comprehensive knowledge in related technical fields, including reactor physics, radiological l
hazards control, nuclear engineering and instrument engineering.
l Five years experience in the supervision and management of l the construction and operation of reactor facilities.
Demonstrated ability to plan, organize, and direct reactor plant operations for several reactor types.
- b. Manager, Plant Engineering B.S. in Engineering or Science or equivalent in experience.
j Ten years experience or equivalent in the operation and maintenance of power-generation facilities, including a minimum 3
l of three years in responsible supervisory positions in the operation or maintenance of such facilities.
6.1-2
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. Ability to plan, program, and direct activities of engineering and craf,t personnel, s .
Demonstrated ability in the design and application of equipment and devices, and have a thorough understanding of process equipment such as pumps, fans, beat exchangers and generators, beaters, etc., as applicable to nuclear facilities.
- c. Manager, Operations B.S. in Engineering or Science or equivalent in experience.
Five years experience in the operation and maintenance of reactor or nuclear power facilities, including minimum of one year in supervisory positions in the operation and maintenance of such facilities.
Demonstrated ability to organize and coordinate plant operations.
Comprehensive knowledge of problems associated with startup and initial operation of reactor facilities, including knowledge
.of radiological hazards, technical aspects of reactor operation
! of control systems, radiation shielding, contamination control, etc.. Demonstrated good judgment necessary to umke correct decisions j under rapidly changing conditions.
- d. Manager, Programs and Analysis B.S. in Engineering or Science or equivalent in experience.
Five years experience in the design, operation, analysis and progranuning of a variety of reactor types or nuclear power facilities, including at least one year in responsible supervisory l
L position in such organizations.
4 i
6.1 5
L %
i
. Comprehensive knowledge of reactor physics, reactor design,
! reactor operation, radiation shielding, fluid flow, thermo- l dynamics, instrumentation, and related technologies. l i
Demonstrated capability for directing the efforts of physicists j
and engineers.
Ability to develop techniques and test procedures to carry out a reactor experimental program.
Demonstrated knowledge of the practical aspects of the operation of reactors, including the characteristics, limitations, and safe operating requirements. !
- e. Specialist. Radiation and Industrial Safety B.S. in Chemistry or Chemical Engineering or equivalent in experience.
Three years experience in analytical chemistry and health physics, and one year in radio-chemistry.
Deomonstrated ability in evaluation of radiation hazards, design and development of radiation monitoring equipment, and in conducting health-physics studies.
Thorough understanding of radiation dosimetry and a working knowledge of design of radiation facilities, shielding calcula-l tions and design of ventilation control, radioactive waste processing, calibration of radiation measuring instrumentation, maximum permissible radiation exposure levels, and good radio-logical safety and health protection practices.
6.1.5
- , - - . . . - - - _ . _. . - - - . - . - .. - ~.- _._ ~. . . - . -. . - .
I i
Must be cognizant of local and state industrial safety requirements.
Demonstrated ability in teaching, lecturing, and implementing s
l .
l safe practices and procedures.
- f. Supervisor, Mechanical Maintenance
! High school education and apprenticeship training in metal-l .
l working fields, coupled with at least ten years practical
! experience in mechanical shop or industrial processes, including l maintenance and fabrication. j l 1 Five years of supervisory responsibility in a mechanical shop or industrial establishment.
t Comprehensive knowledge of all phases of mechanical unintenance, l
including such mechanical crafts functions as machining, pipe !
fitting, velding, carpentry, and rigging on reactor process equipment, including those which have been exposed to radiation.
Cognizance of radiation and safety procedures and regulations, l as applicable to nuclear facilities.
- g. Instrumentation Engineer B.S. in Electrical Engineering or equivalent in experience.
! Three years experience in design, installation, calibr's tion and maintenance of process and nuclear instrumentation.
Cognizance of significance of control and instrumentation systems with respect to reactor operation and safety. Demonstrated
[ ability to analyze systems for adequacy to meet systems require-1 l ments and to conceive, assemble, and install necessary modifications a
! to meet systems requirements.
t 6.1-6 w ., , , e m -, +
.. . - . . - . . . - . . - - . - - - - = . . _ . - . . . . - - - - . - - . = - - - . - - - . - . - . -
i.
9 ,
s l l .
j-
- b. Ghift Supervisor B.S. in Engineering or Physics or equivalent in experience.
s .
Three years experience in the operation of reactor or nuclear facilities.
Knowledge of reactor startup methods and procedures, including radiological hazards and their control, plant maintenance, modern physics and other technical aspects of reactor facilities. i Ability to plan, coordinate, and direct the efforts of operations i personnel as indicated by previous supervisory experience or ,
l satisfactory progression in job positions.
l l Licensed as a Senior Beactor Operator.
l B. Review and Audit l
l Organizational units for the review and audit of plant operations shall be constituted and have the responsibilities and authorities outlined below:
- 1. SEFOR Site Safety Commaittee
- a. Membership l
I Chairman: Facility Manager or designated alternate, Manager, Operations or designated alternate.
Manager, Plant Engineering or designated alternate.
Manager, Program and Analysis or designated alternate .
Specialist, Radiation and Industrial Safety or designated alternate.
- b. Meeting Frequency At least every two weeks and more often as deemed necessary by the Chairman.
6.17 l
I
.g. .-
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t
- c. Quorum Chairman or his designated alternate, plus two other members.
'd. Responsibilities I
(1) Review all proposed normal, abnormal, emergency 1
3 procedures, and procedures for maintenance which are significant to reactor safety and proposed changes to j these procedures.
l (2) Review all proposed tests for the planned experimental 3
program, and plant tests which may have significance to reactor safety.
(3) Review proposed changes to Technical Specifications.
I j (4) Review proposed changes and modifications to plant i
systems or equipment which would require a change in, or would be covered by procedures in d(1) above.
i (5) Review plant operation to detect potential safety hazards.
(6) Review reported violations of Technical Specifications.
(7) Perform special reviews and investigations and make recommendations thereon as requested by the SEFOR Facility Manager.
(8) Report to the APO General Manager and to the Chairman of the SEFOR Safety Review Committee on all reviews and investigation conducted under items d(6) and d(7) above.
6.1.8
o, ,, i s .
l
.(9) Make tentative determinations regarding whether proposals
'f considered by the committee iny'lve o unreviewed safety questions in accordance with 10 CFR 50 59 and recommend the necessary actions or safety analyses to the Facility Manager.
- e. Authority (1) 'Ibe SPOR Site Safety Committee shall be advisory to the Facility Manager.
(2) The SWOR Site Safety Committee shall review and recoussend I
to the Facility Manager approval or disapproval of proposals under items d(l) through d(5) above.
- a. In the event of disagreement between the reconsendations of the SPOR Site Safety Committee and actions comtemplated by the Facility Manager on safety matters, the decision and action to be taken shall be the responsibility of the Facility Manager.
- f. Records )
I Minutes shall be kept for all meetings of the S U OR Site Safety Committee. Copies of the minutes shall be forwarded to the APO General Manager, the Chainman of the SDOR Safety Review Cousaittee, and to the Manager, Safeguards and Analysis.
- g. Procedures Committee rules and regulations shall be prepared and maintained describing the function of the committee, its meeting schedule, methods for review and approval.of evaluations and recommendations, ,
or designation of meetings and such other matters as may be appropriate.
6.1-9
-. __ . _ . __ _ _ _ .