ML19326A694
| ML19326A694 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 07/18/1975 |
| From: | BABCOCK & WILCOX CO. |
| To: | |
| Shared Package | |
| ML19326A693 | List: |
| References | |
| NUDOCS 8002270612 | |
| Download: ML19326A694 (53) | |
Text
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,o NY Y 3/4.9 REFUELING OPERATIONS -
- BORON CONCENTRATION Icl4 LIMITING CONDITION FOR OPERATION
. ?- (i[I 3.9.1 With the reactor vessel head unbolted or removed, the boron concen-tration of all filled portions of the Reactor Coolant System and the re-
,j fueling canal shall be maintained uniform and sufficient to ensure a Keff of no greater than 0.95 which includes a conservative allowance of (A M % Ak/k for uncertainties.
e-i .a APPLICABILITY: MODE 6*. ACTION: -g With the requirements of the above specification not satisfied, immediately M *~ suspend CORE ALTERATIONS and initiate and continue boration at > ( /c) gpm of (PM8) ppm boric acid solution on its equivalent until Keff is reduced to < 0.95. The provisions of Specification 3.0.3 are not applicable. V W vyg , SURVEILLANCE REQUIREMENTS h
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4.9.1 The boron concentration of the reactor vessel,._ctm genes containing nrim w and refueling canal shall be determined by r:N chemical analysis at least 3 times per 7 days with a maximum time interval JMd between samples of 72 hours. W g J
*The reactor shall be in MODE 6 when the reactor vessel head is unbolted or removed.
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%d 3/4 9-1 June 30, 1975 B&W-STS
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-dN %IW REFUELING OPERATIONS M INSTRUMENTATION i - '5.y'il LIMITING CONDITION FOR OPERATION z.. < R .4 j As a minimum, two source range neutron flux monitors shall be
%. , 3.9.2 operating, each with continuous visual indication '- t"c ce"* c' roca .. ,.J oud on: "4 th rf Sh adication i , the -nnuinet. Ah ~5'j APPLICA8ILITY: MODE 6. M"+g 'e^'
ha ACTION: 7 With the requirements of the above specification not satisfied, immediately suspend all operations involving CORE ALTERATIONS or positive reactivity y h; changes. a Y SURVEILLANCE REQUIREMENTS ?:94
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- r. . J 4.9.2 Each source range neutron flux monitor shall be demonstrated OPERABLE by performance of :
A CHANNEL FUNCTIONAL TEST at least once per 7 days, and l a. l
- b. A CHANNEL FUNCTIONAL TEST within 8 hours prior to the initial start of CORE ALTERATIONS, and
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- c. A CHANNEL, CHECK at least once per 12 hours during CORE ALTERATIONS.
o June 30, 1975 B&W-STS 3/4 9-2
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i REFUELING OPERATIONS a, v
'1 DECAY TIME l'
LIMITING CONDITION FOR OPERATION
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3.9.3 The reactor shall be subcritical for a minimum of (MC7 hours [, - prior to movement of irradiated fuel in the reactor pressure vessel. '~y:
- APPLICABILITY: MODE 6.
6 ACTION: 7 2. With the reactor subcritical for less than (.M67 hours, suspend all cj e operations involving movement of irradiated fuel in the reactor pressure vessel. -74,;Ji - D
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%% .s' SURVEILLANCE REQUIREMENTS 12-4.9.3' The reactor shall be determined to have been subcritical for at least (Wff hours by verification of the date and time of subcriticality prior to movement of irradiated fuel in the reactor pressure vessel.
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y:q 2 - l 1 _s l 3/4 9-3 December 30, 1974 B&W-STS
Nu k "hWNkN WE /g W REFUELING OPERATIONS CONTAINMENT PENETRATIONS
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. LIMITING CONDITION FOR OPERATION 4.] .
3.9.4 The containment penetrations shall be in the following status: 4
- a. The equipment door closed and held in place by a minimum of
$'] four bolts,
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^ ( r, .} b. A minimum of one door in each airlock closed, and d c. Each penetration providing direct access from the containment atmosphere to the outside atmosphere shall be either:
- 1. Closed by an isolation valve, blind flange, or manual valve, or
- 2. Be capable of being closed by an OPERABLE automatic containment purge and exhaust isolation valve.
APPLICABILITY: During core alterations or movement of irradiated fuel
~
within the containment. $5b4 sg ACTION: With the requirerr.ents of the above specification not satisfied, immediately suspend all operatians involving CORE ALTERATIONS or movement of irradiated fuel in the containmer.t. SURVEILLANCE REQUIREMENTS gL P Mib '* 4.9.4 Each of the above required containment penetrations shall be determined to be either in its isolated condition or capable of being closed by an OPERABLE automatic containment purge and exhaust valve within 100 hours prior to the start of and at least once per 31 days l during CORE ALTERATIONS or movement of irradiated fuel in the contain-ment by:
.. -1
- a. Verifying the penetrations are in their isolated cor:dition, or
~* b. Testing the containment purge and exhaust valves per the applicable portions of Specification (4.6.4.1.2).
3/4 9-4 June 30, 1975 B&W-STS
-- F N .Nf_ [[~kh3 hj-b REFUELING OPERATIONS
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"1 COMMUNICATIONS
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' 4'rq-i , LIMITING CONDITION FOR OPERATION
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; 3.9.5 Direct comunications shall be maintained between the control
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- room and personnel at the refueling station during fuel or control rod
;,.,1 movement within the reactor pressure vessel.
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- ' APPLICABILITY: MODE 6.
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_ ACTION: When direct communications between the control room and personnel at the 4W) refueling station cannot be maintained, suspend all operations involving fd" fuel or control rod movement within the reactor pressure vessel. D
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SURVEILLANCE REQUIREMENTS 4.9.5 Direct communications between the control room and personnel at the refueling station shall be demonstrated within one hou-. prior to the start of and at least once per 12 hours during fuel or control rod move-ed ment within the reactor pressure vessel. M a~._a
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March 10, 1975 B&W-STS 3/4 9-5
y -h @ h ' b ' V REFUELING OPERATIONS FUEL HANDLING BRIDGE OPERABILITY
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7 $..,$,; LIMITING CONDITION FOR OPERATION
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f.1 3.9.6 The fuel handling bridge,shall be used for movement of control
' 'C rods or fuel assemblies and shall be OPERABLE with:
$,.j a. A minimum capacity of 32r0pounds, including weight of grapple
't -
tube, and M 2rco
- b. An overload cutoff limit of pounds, including weight of d - grapple tube.
During movement of control rods or fuel assemblies.
. ,j APPLICABILITY:
N ACTION: With the requirements of the above specification not satisfied, suspend use of the inoperable bridge (s) for all operations involving movement of control rods or fuel assemblies. V
.nQ QJ SURVEILLANCE RE0VIREMENTS 4.9.6- The fuel handling bridges shall be demonstrated OPERABLE within 100 hours prior to the start of moving control rods or fuel assemblies by performing a load test of at least 32d pounds and demonstrating an automatic load cutoff when the bridge load exceeds 2Napounds. ?#-y -
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1 3/4 9-6 June 30, 1975 ; B&W-STS I l
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)~ .g , DECAY HEAT REMOVAL ft.. .sn
.k LIMITING CONDITION FOR OPERATION e%
.%fd p 3.9.8 A minimum of one decay heat removal loop shall be OPERABLE.
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,'- APPLICABILITY: MODE 6.
i~ d de.E ACTION:
.ff' '@ With the requirement of the above specification not satisfied, suspend all operations involving an increase in the reactor decay heat load or a l reduction in boron concentration of the Reactor Coolant System and close . .q the containment penetrations pursuant to Specification 3.9.4.
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SURVEILLANCE REQUIREMENTS D 2&oo 4.9.8 A decay heat removal loop shall be demonstrated OPERABLy and ! capable of circulating reactor coolant at a flow rate of > GGeCTT gpm _ l within 24 hours prior to the start of CORE ALTERATIONS and at least once i
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per 24 hours during CORE ALTERATIONS. 12's.q , I l l B&W-STS 3/4 9-8 June 30, 1975
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.. REFUELING OPERATIONS w4 FUEL HANCONS-
'- 9 CRANE TRAVEL - SDr"'
"'t"t "' ^ : = 25 BUILDING
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- a. c A LIMITING CONDITION FOR OPERATION d 56A */n M y ee. g* mM M
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,_.2 3.9.7 C_ I_ N. - ,- -- e .
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e'/c r rr;diated #"c! $cceMr i e i- th; stor;;; pcol.
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;y;;.'. APPLICABILITY: With fuel assemblies in the storage pool.
' .] ACTION:
N'1 JD - With the requirements of the above specification not satisfied, place load in a safe condition.
. .~s W.A SURVEILLANCE REQUIREMENTS Q .i.$ L M di d f A W $
2I 4.9.7 Crane interlocks and/or physical stops whibccv;r.t arme tr ;cl ith !ced? 4a ex:::: # ; cur.d; secr irradiated fuel assemblies , shall be demonstrated OPERABLE within 7 days prior to crane operation and at least crice per 7 days during crane operation. (~
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3/4 9-7 June 30, 1975 B&W-STS
5 $ f Y .Y.h '* V REFUELING OPERATIONS . tT'M CONTAINMENT PURGE AND EXHAUST ISOLATION SYSTEM 7
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(4'.c ; LIMITING CONDITION FOR OPERATION
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.\ Ob 3.9.9 The containment purge and exhaust valve isolation system shall be l
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~ .1 OPERABLE.
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- APPLICABILITY: MODE 6.
;m d '/ ACTION: '?
D . . . With the containment purge and exhaust valve isolation system inoperable, each of the penetrations providing direct access from the containment atmosphere to the outside atmosphere. f.< M SURVEILLANCE REQUIREMENTS l ib*h 4.9.9 The containment purge and exhaust valve isolation system shall 1-@ be demonstrated OPERABLE within 100 hours prior to the start of and at least once per 7 days during CORE ALTERATIONS by verifying that containment purge and exhaust isolation occurs on manual initiation l and on a high radiation signal from each of the containment radiation monitoring instrumentation channels. , i bh
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Ase.4 , 3/4 9-9 June 30, 1975 i B&W-STS l
he % [ k Y' '- s' ~ W REFUELING OPERATIONS e WATER LEVEL - REACTOR VESSEL s.y.
' E *y LIMITING CONDITION FOR OPERATION +-y .' N 3.9.10 As a minimum, 23@ feet of water shall be maintained over the 1 c , c ,3 -.? top of irradiated fuel assemblies seated within the reactor pressure m vessel. * **v 9 4-$ APPLICABILITY: During movement of fuel within the reactor pressure 6N vessel.
ACTION: With the requirements of the abcve specification not satisfied, suspend all operation involving movement of fuel within the reactor pressure h<ih. v vessel. V SURVEILLANCE REQUIREMENTS
.$N*
/t9 4.9hThewaterlevelshallbedeterminedtobewithinitsminimum required depth within 2 hours prior to the start of and at least once per 7 days during fuel movement within the reactor pressure vessel.
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B&W-STS 3/4 9-10
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.c s STORA' G E P00L WATER LEVEL g /;,, LIMITING CONDITION FOR OPERATION
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" 3.9.11 As a minimum, 23 feet of water shall be maintained over the top
..'- of irradiated fuel assemblies seated in the storage racks.
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,y APPLICABILITY: Whenever irradiated fuel assemblies are in the storage w pool.
ci-8 ACTION:
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- With the requirement of the specification not satisfied, suspend all movement of fuel and crane operations with loads in the fuel storage
-* area and restore the water level to within its limit within 4 hours. l v SURVEILLANCE REQUIREMENTS w.m .Q
- P_4 4.9.11 The water level in the storage pool shall be determined to be at least its minimum required depth at least once per 7 days when irradiated fuel assemblies are in the fuel storage pool.
y :.yg 4 , June 30, 1975 B&W-STS 3/4 9-11
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STORAGE POOL VENTILATION SYSTEM
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~D LIMITING CONDITION FOR OPERATION
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'. 3.9.12 Two separate and independent fuel storage pool ventilation 2 systems shall be OPERABLE.
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">;'. I APPLICABILITY: Whenever irradiated fuel is in the storage pool. . :Q i 1
ACTION: With one fuel storage pool ventilation system inoperable, fuel a.
.s movement within the storage pool or crane operation with loads -W over the storage pool may proceed provided the remaining fuel M storage pool ventilation system is in operation and discharging thru at least one train of HEPA filters and charcoal adsorbers.
- b. With no fuel storage pool ventilation system OPERABLE, suspend all operations involving movement of fuel within the storage pool or crane operation with loads over the storage pool until at least one
- U spent fuel storage pool ventilation system is restored to OPERABLE status. . r .g~ ,
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- c. The provisions of Specification 3.03 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS g%. 4.9.12 The above required spent fuel storage pool ventilation system l
. shall be demonstrated OPERABLE:
7" --
- a. At least opce per 31 days by:
- 1. Initiating flow through the HEPA filter and charcoal adsorber train and verifying that the train operates for at least 10 hours with the heaters on, and s.. 2. Verifying that the filter train is aligned to receive electrical power from an OPERABLE emergency bus.
^# b. At least once per 12 months or af ter every 720 hours of system operation and (1) af ter each complete or partial replacement of a HEPA filter or charcoal adsorber bank, or (2) after any structural maintenance on the HEPA filter or charcoal adsorber housings, or (3) following painting, fire or chemical
. release in any ventilation zone communicating with the system i by: 1 B&W-STS 3/4 9-12 June 30,197c l l l l
2 Y. $ $'- ff .h Yi W REFUELING OPERATIONS M h- SURVEILLANCE REQUIREMENTS (Continued)
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"$.9 1. Verifying that with the ventilation system operating at a flow rate of cfm i 10 percent and exhausting through Q '-
the HEPA filters and charcoal adsorbers, the total bypass
-%;.] flow of the ventilation system to the facility vent, including leakage through the ventilation system diverting valves, is 1 1 percent when the ventilation system is 7' *W.eTh tested by admitting cold 00P at the storage pool venti-p lation system intake.
- 2. Verifying that the charcoal adsorbers remove 1 99 percent
/: m of a halogenated hydrocarbon refrigerant test gas when they are tested in-place in accordance with USAEC Report N DP-1082 while operating the ventilation system at a flow T rate of cfm i 10 percent.
- 3. Verifying that the HEPA filter banks remove 1 99 percent of the 00P when they are tested in-place in accordance with ANSI N101.1-1972 while operating the ventilation O system at a flow rate of cfm i 10 percent.
%g' 4. Subjecting the carbon contained in at least one test y
canister or at least two carbon samples removed from one of the charcoal adsorbers to a laboratory carbon sample analysis and verifying a removal efficiency of 1 90 percent for radioactive methyl iodide concentration of 0.05 to 0.15 mg/m3 , 1 95 percent relative humidity, and
> 125 F; other test conditions shall be in accordance Jith USAEC RDT Standard M-16-lT, June 1972. The carbon ;:My samples not obtained from test canisters shall be prepared 5
by either: 5 7 a) Emptying one entire bed from a removed adsorber tray, mixing the adsorbent thoroughly, and obtaining samples at least two inches in diameter and with a length equal to the thickness of the bed, or b) Emptying a longitudinal sample from an adsorber tray, mixing the adsorbent thoroughly, and obtaining samples at least two inches in diameter and with a length equal to the thickness of the bed. fg
- 5. Verifying a system flow rate of cfm i 10 percent during system operation.
B&W-STS 3/4 9-13 June 30, 1975 l
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'GW REFUELING OPERATIONS
...q y,'j SURVEILLANCE REQUIREMENTS (Continued) $Nd I' c. At least once per 18 months by: W-"
- 1. Verifying that the pressure drop across the combined HEPA
.'7%@ filters and charcoal adsorber banks is < (6) inches Water Gauge while operating the ventilating system at a flow N'9 rate of cfm + 10 percent. ?$t) 1.l::::2 2. Verifying that the air flow distribution to each HEPA filter and charcoal adsorber is within + 20 percent of
~
the averaged flow per unit.
-5%
bQ3 _. 3. Verifying that on a high radiation signal, the system automatically starts (unless already operating) and directs its exhaust flow through the HEPA filters and charcoal adsorber banks.
- 4. Verifying that the ventilation system maintains the spent fuel storage pool area at a negative pressure of > (1/4)
~M inches Water Gauge relative to the outside atmosphere W 4 during system operation. g_,
- 5. Verifying that the filter cooling bypass valves can be manually opened.
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- =a June 30,1975 l B&W-STS 3/4 9-14
N$ Y bY h? bh$5Y $ n g ,< 3/4.10 SPECIAL TEST EXCEPTIONS
# GROUP HEIGHT AND WITHDRAWAL LIMITS 2")
- :1
. v LIMITING CONDITION FOR OPERATION l$ 3.10.1 The group height, withdrawal limits and overlap requirements
- may be of Specifications 3.1.3.1, 3.1.3.2, 3.1.3.5, 3.1.3.6 and & .
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suspended during the performance of PHYSICS TESTS provided: 1a 7 d /w.
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a.
. H QA The Nucler '% gfin
..cr trip setpoint is reduced to < (85)% of J 7.' RATED THERMAL POWER, and
' ' z .t,; The limits of Specifications {3.2.11, 23.2.2) and 43.2.3~) are l b. maintained and determined at the frequencies specified in , l
*4 Specification 4.10.2 below.
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APPLICABILITY: MODES 1 and 2. ACTION: I With any of the limits of Specifications 3.2.1, 3.2.2 or 3.2.3 being ' exceeded while the requirements of Specifications 3.1.3.1, 3.1.3.2, g - 3.1.3.5, 3.1.3.6 cr 3.1.3.7 are suspended within two hours either: l i to satisfy the requirements ] M;g, ,%
- a. Reduce THERMAL POWER sufficiently)E and $3.2.31, or
' -- of Specifications 13.2.11,13.2.2
- b. Be.in HOT STANDBY.
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SURVEILLANCE REQUIREMENTS l 4W 1 4.10.2.1 The Nuclear Overpower trip setpoint shall be determined to be M set at < (85)% of RATED THERMAL POWER within 12 hours prior to the initiation of and at least once per 12 hours during PHYSICS TESTS in which l the requirements of Specifications /3.1.3.11, [3.1.3.2J,43.1.3.51,
.f 3.1. 3.6 ), or (i . _ . are suspended. 2. s.2 4.10.2.2 The Surveillance Requirements of Specifications (4.2.i} , f 4.2.2).
^ and 14.2.33.shall be performed 3t the folicwing frequencies during PHYSICS TESTSinwhichtherequirementsofSpecificationsf3.1.3.l},43.1.3.23,
'h-
'M 33.1.3.5),}3.1.3.6) or W are suspended: f & /M ? l
- a. Specification (4.2.1) - At inst once per ~ hours. j l
- b. Specification (4.2.2) - At least once per __ hours.
hours. I
- c. Specification (4.2.3) - At least once per s
3/4 10-1 March 17, 19 l B&W-STS
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SPECIAL TEST EXCEPTIONS V PRESSURE / TEMPERATURE LIMITS - REACTOR CRITICALITY l I4
- h, j LIMITING CONDITION FOR OPERATION --
%j ' Y; ? 3.10.2 The minimum temperature and pressure conditions for reactor criticality of Specifications 3.1.1.4 and 3.4.9.1 may be suspended -4 during PHYSICS TESTS provided:
v tA The THERMAL POWER does not exceed 5% of RATED THERMAL POWER,
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C b. The4 reactor trip setpoints on the OPERABLE TL i n c M rpc.;cr
~~ Channels are set at < 25% of RATED THERMAL POWER, and
,M c. The Reactor Coolant System temperature and pressure relation-WJ ship is maintained within the region of acceptable operation d shownonFigure{3.4-2%
APPLICABILITY: MODE 2. ACTION: - - - D~
- a. With the THERMAL POWER > 5% of RATED THERMAL POWER, immediately open the reactor trip breakers.
{ With the Reactor Coolant System temperature and pressure b. relationship within the region of unacceptable operation on Figure f3.4-25, innediately open the reactor trip breakers and restore the temperature-pressure relationship to within its limit; perform the analysis required by Specification (3.4.9.1)' prior to the next reactor criticality.
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g . . SURVEILLANCE REQUIREMENTS 4.10.2.1 The Reactor Coolant System shall be verified to be within the regiqn _of_ac_ceptable operatio_n .of Figure 3.4-2 at least once per hour. 4.10.2.2 The THERMAL POWER shall be determined to be < 5% of RATED
*. THERMAL POWER by indication on the Intermediate Range and Power Range Channels at least once per hour. (
4.10.2.3 Each Intermediate Range and fLdar ?:rpcocr-Channel shall be subjectad to a CHANNEL FUNCTIONAL TEST within 8 hours prior to initiating PHYSICS TESTS. 3/4 10-2 March 17, 1975
. B&W-STS
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$. hh' ? k b h ? h S*
., Defar SPECIAL TEST EXCEPTIONS (No fe,r p
.)
. _./ _ BORON' CONCENTRATION
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%d LIMITING CONDITION FOR OPERATION
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.. ..o 'fh 3.10.3 The requirement of Specification 3.9.1 to maintain a uniform baron concentration in all filled portions of the Reactor Coolant System and the 7 refueling canal sufficient to ensure a Keff of no greater than 0.90 with ". d*J all full length rods fully inserted may be modified to permit Keff to be ~,Qej no greater than 0.95 with all full length rods fully inserted during initial .g fuel loading operations.
^"
APPLICABILITY: MODE 6. A Og ACTION:
..wa With Keff > 0.95, immediately suspend CORE ALTERATIONS and initiate and continue boration at > (70) gpm until Keff is reduced to 1 0.95.
The provisions of Specification 3.0.3 are not applicable. $b : SURVEILLANCE REQUIREMENTS 4.10.3 The boron concentration of the reactor pressure vessel, steam generators containing primary water, and the refueling canal shall pM be determined by chemical analysis at least once per 12 hours when Keff [g-is > 0.90. ~m u
' B&W-STS 3/4 10-3 March 17, 1975
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SPECIAL TEST EXCEPTIONS V PHYSICS TESTS LIMITING CONDITION FOR OPERATION 3.10.4 The limitations of Specifications 3.1.3.6, 3.2.1, 3.2.2, 3.2.3 and 3.2.4 may be suspended during the performance of PHYSICS TESTS provided: l
- a. The THERMAL POWER does not exceed 5% of RATED THERMAL POWER, and y,3 L Pto f%er la-
. b. TheAreactor trip setpoints on the OPERABLE "ecitar 0;;rp;wc,e r
channels are set at < 25% of RATED THERMAL POWER.
- c. The nuclear instrumentation Source Range and Intermediate Range high startup rate control rod withdrawal inhibit is OPERABLE.
APPLICABILITY: MODE 2. ACTION: With the THERMAL POWER > 5% of RATED THERMAL POWER, immediately open the reactor trip breakers. _p SURVEILLANCE REQUIREMENTS 4.10.4.1 The THERMAL POWER shall be d'etermined to be < 5% of RATED THERMAL POWER at least once per hour during PHYSICS TE5'TS by indication on the Intermediate Range and Power Range channels.
?%n Aw 4.10.4.2 Each Source and Intermediate Range and Nuuicai 0, m p;w;3re l channel shall be subjected to a CHANNEL FUNCTIONAL TEST within 8 hours prior to initiating PHYSICS TESTS.
l l l B&W-STS 3/4 10-4 May 19, 197S ~ e
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) 3 /4. 9.'l BORON t0NCENTRATION The limitations on minimum baron concentration ensure that: 1) the
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- . reactor will remain subcritical during CORE ALTERATIONS, and 2) a un'iform boron concentration is maintained for reactivity control in the water volumes having direct access to the reactor vessel. The limitation on Keff of no greater than.Q.Wis sufficient to prevent reactor criticality with all safety and regulatai g rods fully withdrawn.
e75
, .- 3 /4 . 9. 2 INSTRUMENTATION The OPERABILITY of the source range neutron flux monitors ensures that redundant monitoring capability is available to detect changes in the reactivity condition of the core.
' /- 3/4.9.3 DECAY TIME
~
l The minimum.reiuirement l for reactor subcriticality prior to movement
- of irradiated fuel in the reactor pressure vessel ensures that suf-
! ficient time has elapsed to allow the radioactive decay of the short
! lived fission products. This decay time is consistent with the assump-l tions used in the safety analyses.
3/4.9.4 CONTAINMENT PENETRATIONS
~
The requirements on containment penetration closure and OPERABILITY
- I ensure that a release of radioactive material within containment will be i restricted from leakage to the environment. The OPERABILITY and closure i requirements are sufficient to restrict radioactive material release l from a fuel element rupture based upon the lack of containment pressuriza-tion potential while in the REFUELING MODE.
s
,' 3/4.9.5 COMMUNICATIONS C .1 The' requirement for communications capability ensures that refueling j station personnel can be promtply informed of significant changes in the i
facility status or core reactivity condition during CORE ALTERATIONS. s g* .
; ! Ba'4-STS B 3/4 9-1 May 7, 1975 4
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YhkfYb? 5Y5kkh I 5 f ' ~5 W ? h b Y ?N 1 Q REFUEllflG OPERAT10lls k nr BASES' 7 ft/e/ HanaMQ 8Nayl es T. 3/4.9.6 E-ME OPERABILITY - EONTAIN".:7 EUILDP!G bN c(5 e s d .y The OPERABILITY requirements of the Or m ased for movement of it fuel assemblies ensures that: 1) each crane has sufficient load capacity to lif t a fuel element, and 2) the core internals and pressure vessel
~
-$t are protected fiom excessive lif ting force in the event they are inadver-f@ C tently engaged during lif ting operations. W .a dh 3/4.9.7 CRAtlE TRAVEL - SPEilT FUEL STORAGE BUILDIllG - 9 The restriction on movement of loads in excess of the nominal ! weight of a fuel element over irradiated fuel assemblies in the storage l
# pool ensures that no more than the contents of one fuel assembly will be ruptured in the event of a fuel handling accident. This assumption is 4"
consistent with the activity release assumed in the safety analyses. t 4 3/4.9.8 DECAY HEAT REMOVAL !' vl W* *Y The OPERABILITY of one rmm neat removal loop ensures that sufficient cooling capacity is available to remove decay heat. This Q.a
, requirement is sufficient to maintain the water in the reactor pressure vessel below 140 F as required during the REFUELIllG MODE.
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3/4.9.9 ara ? " . .10 CONTA:.JENT 7A :'T!:S '"'P:' "PC ., J-C0ftTAINMErlT l PURGE Afl0 EXHAUST ISOLATION SYSTEM The OPERABILITY of these systems ensure that the containment purge
-- and exhaust penetrations will be automatically isolated upon detection M;, -
of high radiation levels within the containment. The OPERABILITY of .g these systems are required to restrict the release of radioactive M material from the containment atmosphere to the environment.
. ,o il 3/4.9.% and 3/4.9.M. WATER LEVEL - REACTOR VESSEL AllD STORAGE POOL '
WATER LEVEL The restrictions on minimum water level ensure that sufficient water depth'is available to remove 99~, of the assumed 10" iodine gap 9 activity released from the rupture of an irradiated fuel assembly. The minimum water depth is consistent with the assumptions of the safety analysis. v B&W-STS B 3/4 9-2 May 7, 1975 e
.b. .I. ? .h.Yh l:
V REFUELING OPERATI0tl5 BASES 'i W. ;q - _ j - ., [ ' Mb'
/4.9.13 STORAGE P0OL RADIATION MONITORING SYSTEM
,', {,. The OPERABILITY of the storage pool radiation monitoring system ., .yf;,' ensures that sufficient radiation monitoring capability is available k to detect excessive radiation levels resulting from 1) the inadvertent owering of the st'orage pool water level or 2) the release of activity
-;g{.] fro . n irradiated fuel assembly.
2% l2. 1 ' 3/4.9.W = d 2 / ; . J . '. : STORAGE POOL VENTILATI0ft SYSTEM - FUEL MOVEMEtlT AllD FUEL STORAGE w.
*3 ,,g?,% .. The requirements on the storage pool ventilation system to be 1 operating or OPERABLE ensure that all radioactive material released j from an irradiated fuel assembly will be filtered through the HEPA filters and charcoal adsorber prior to discharge to the atmosphere.
V The OPERABILITY of tnis system and the resulting iodine removal capacity are consistent with the assumptions of the safety analyses. The 60 day
- decay of irradiated fuel is equivalent to the decontamination factor
- s. 4, of the ventilation system filters and adsorber.
3/4.9.16 FUEL ASSEMBLY HAfi$Tii16 (CYMCMI T _c64% The limitation on handling more than one fuel assembly at a time ermits each fuel handling bridge to be utilized so long as fuel assemb y aration is sufficient to prevent inadvertent criticality. 7 Tf . Y$ , mi 1 i B&W B 3/4 9-3 May 7, 1975 t
i M h h __ N h N. N .hh h f V 3/4.10 SPECIAL TEST EXCEPTI0 tis BASES gk TI,] 3/4.10.1 GROUP HEIGHT A.'lD WITHDRAWAL LIMITS x This special test exception permits individual control rods to be .W/. a positioned outside of their normal group heights and withdrawal 1imits R5 during the performance of such PHYSICS TESTS as those required to 1)- Y,, measure control rod worth and 2) determine the reactor stability index
*-.~.-
. and damping factor under xenon oscillation conditions.
.,k
.~.?"' 3/4.10.2 PRESSURE / TEMPERATURE LIMITS - REACTOR CRITICALITY This special test exception permits the reactor to be critical at
.Q up to 5% of RATED THERMAL POWER during PHYSICS TESTS required to measure s Lt such parameters as control rod worth and SHUTD0'.-iti MARGIt!. 4ft'TO. 3 BOR0fl CONLtuiRATiGu I
/ This special Reactor Coolant eption permits the boron concentration of the.
ud refueling canal to be reduced sufficient to
/
( establish a reac 6 core Keff of no greater than 0.95 during initial
\ fuel loading. This special test exception is required to obtain d@'
7, \ sufficient count rate on the core neutron monitors to adequately plot h verse ruit3 & i - un uu. 6 J ci ly addition. 3/4.10.4 PHYSICS TESTS This special tes't exception permits P'HYSICS TESTS to be performed _. at up to 5"., of RATED THERMAL POWER and is required to verify the
~
{{.g fundamental nuclear characteristics of the reactor core and related
. . , - instrumentation.
i I M l Wa l l v I l l B&W-STS B 3/4 10-1 May 7, 1975 ; l i l l
v 5.0 DC'ilGil FEATURES
~
5.1 SITE EXCLUSI0il AREA 5.1.1 The exclusion area is shown on Figure 5.1-1. L0tl POPULATIOll ZONE 5.1.2 The low population zone is shown on Figure 5.1-2. FLOOD C0flTROL 5.1.3 The flood control provisions (dikes, levees, etc.) shall be designed and inaintained in accordance with the original design pro-visions contained in Section42.4.2.2 / of the FSAR.
-4 5.2 C0i!TAII:I1E;1T-- camiya.,4<w[6-,.,M.,
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,b *. .w % <+uA
- a. t!ominal inside diametef = m feet.
dv4%:- . t. % ,t .
- b. ficminal inside height 3 = gfeet.
- c. Miniir.um thickness of concrete walls = 2.5 feet.
- d. Minimum thickness 'of concrete roof = ?.0 feet.
-e. "" t H i .. :: c' ':r:t: #!::: p1- feet.
r-~ h & . y e +. itominal thickness of steel 44e = I%2. inches. f y. !!et free volume = cubic feet. 2 639 410 B&t!-STS 5-1 00T 1 197i
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\0 0 0" SW 43 47 SE
-7 1970 POPUL ATION SHOWN SSW SSE
- IN0iCATES SEGNENT SUBJ ECT S TO SEASONAL INCREASE DURING SuuuER MONTNS SEE TABLE 2-5
, PERMANENT POP'JLATION PROJECTION lillLS U10. Gli 1331 2210. 20.11 0-1 181 181 !al 211 277 1-2 4I0 4!0 4!0 47 1 506 TOTAL 5 IgI [3I iii Tij FOR SEGMENT DISTRl!UTION IN FUTURE f EARS SEE T ASLE 2-2.
FOR SEASONAL INCREASE IN RE!! DENT 90Pullil0N DURING SCUMER WONTHS. SEE TABLE 2-5.
"' - 211LL 5 ::E '.L':Lt " ' ^ ~:' ' ' S L0# POPUL Ail 0N ZONE FIGURE M I./-2.
$^- N A-A D ' w ' DESIGN FEATURES
~,
- h. , DESIGN PRESSURE AND TEMPERATURE
/f.Q 5.2.2 The reactor containment building is designed and shall be maintained P." for a maximum of internal pressure of 40 psig and a temperature of 21d*f 31 .; f PENETRATIONS
- 5. {'
5.2.3 Penetrations through the reactor containment building are designed '.:IM
'N and shall be maintained in accordance with the original design provisions T contained in Section (6.2.4) of the FSAR with allowance for normal degradation pursuant to the applicable Surveillance Requirement.
P I.h 5.3 REACTOR CORE FUEL ASSEMBLIES 5.3.1 The reactor core shall contain / 77 fuel assemblies with each fuel assembly containingEc5 fuel rods clad with jZircaloy -4f. Each (/ fuel rod shall have a nominal active fuel length of 144 inch 6s and contain a maximum total weight of 25c0 grams uranium. The initial core
. 4.% loading shall have a maximum enrichment of g weight percent U-235.
N% Reload fuel shall be similar in physical design to the initial core loading and shall have a maximum enrichment of M weight percent U-235. The first cycle ' fuel loading shall contain 69 burnable poison rod l assemblies with each assembly containing up to & burnable poison rods of sintered Al203-84C clad with Zircaloy-4.
- CONTROL RODS W 5.3.2 The reactor core shall contain53 safety and regulating and 1 axial power shaping (APSR) control rods. The safety and regulatingThe control rods shall contain a nominal 134 inchee of absorber material.
APSR's shall contain a nominal 36 inches of absorber material at their lower ends. The nominal values of absorber material shall be 80 percent silver,15 percent indium and 5 percent cadmium. All control rods shall
- be clad with stainless steel tubing. The balance of the void length in
" '~'-- '
l the APSR's + ~hh 5 "* khq , 5-4 June 30, 1975 B&W-STS
.,$ ($ I[IONh h' 5 h$.h r V DESIGN FEATURES 5.4 REACTOR COOLANT SYSTEM
. h:l X' Ll DESIGN PRESSURE AND TEMPERATURE 5.4.1 The reactor coolart system is designed and shall be maintained: -Q[C In accordance with the code requirements specified in Section 9fl a. (5.2) of the FSAR, with allowance for normal degradation M pursuant'to the applicable Surveillance Requirement. .M4 i.h ,v 3 ..Q*j' b. For a pressure ofggfqC, psig, and X 4 Forcla
~ ~
- c. a3 temMture of& F, except for the pressurizer and pressurizer surge line which is (g70 F.
,g N ~~
VOLUME 5.4.2 The total water and steam volume of the reactor coolant system is I
//g + 52 cubic feet at a nominal Tavg of 525 F.
V 5.5 EMERGENCY CORE COOLING SYSTEMS EM-
"' The emergency core cooling systems are designed and shall be 5.5.1 maintained in accordance with the original design provisions contained in Section (6.3) of the FSAR, with allowance for normal degradation pursuant to the applicable Surveillance Requirements.
5.6 FUEL STORAGE CRITICALITY The new and spent fuel storage racks are designed and shall be &g> 5.6.1 maintained with sufficient center-to-center distance between i fuel e lent to assemblies placed itt the storage racks to ensure a k -TSc h F
< 0.'5Cwith the storage pool filled with unborated ;
{-4. 90 l DRAINAGE _
- 5.0.2 The spent fuel storage pool is designed and shall be maintained to prevent inadvertent draining of the pool below efestion t _ y, . . .
5-5 June 30, 1975 B&W-STS
? h $ h$b YrY$~ '. $h k hk? ~
DESIGN FEATURES
.y 5.7 SEISMIC CLASSIFICATION O: ,Di 5.7.1 Those structures, systems and components identified as Category I Items in Section (3.2.1) of the FSAR shall be designed and maintained b to the original design provisions contained in Section (3.7) of the FSAR .Rh with allowance for normal degradation pursuant to applicable Surveillance lI ~ :d Requirements. *~s.; -
e .." 5.8 METEOROLOGICAL TOWER LOCATION (p,( The meteorological tower location shall be as shown on Figure W 5.8.1 I (5.1-1). 4 h M.y 5.9 COMPONENT CYCLIC OR TRANSIENT LIMIT 5.9.1 The components identified in Table 5.9-1 are designed and shall be maintained within the cyclic or transient limit of Table 5.9-1. V S ~$h..A _. . ; .t 2 ? w w; ,
. ..a 5-6 June 30, 1975 BiW-3TS
- 4
*i.
a
' j '$ 1 .': ,
l( {. e TABLE 5.9.1 ., i - g COMPONENT CYCLIC OR TRANSIENT LIMITS . s y
' b CYCLIC OR DESIGN CYCLE .
COMPONENT TRANSIENT LIMIT OR TRANSIENT "i c
' Q Reactor %'1D 557 Aid Pressure Vessel (500-heatup and cooldown cycles) (70*F to 464"F to 70 F) .'j
- 48000 s
(+5000 power change cycles) (15% to 100% to 15% RATED "y / R,3 THERMAL POWER)
^ ~
~
- - ~ ~ ~ ~ ~ -
6 t (10 step change cycles) '-(+100psiand+10 Fat 2250} psia and 560 F) . I l /,.kk a co (400 reactor trip cycles) (100% to 0% RATED THERMAL POWER)
.}j.jj s up' l ,,
^
[ 3 6000 ' ( obls V'ru ' 4.e. ( I i'N- . scom m g u , ,t ~ m (i a, a _30 m Jg c~) ch i. <
;i 2
* . 9.
s y , g
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t h' w. F h), if
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4 . 4 1 I r . 1 1 I 13... h-l stall 0ARD ) y- )_ .. Q TECHilICAL SPECIFICATI0tlS 9
* :4 SECTI0tl 6.0 1 ! ADMIT!ISTRAil'/E C0tlTROLS i
4 b September 1, 1974 y,M .
- -r' s
i G 4 v i w s
* ~5_4 .
f' -. i I i l I l s e - -ee-m- 1 ew9 + ,M - ::-- ? -r4m- 4 y qs-+Pr "ww y -gvyTv-- 99 -y y w-51-- T w-.7-- ---ve w' -m+-e a-e e g- ~ teevf'--
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V
.j '. u 6.0 ADMIMSTRATIVE CONTROLS 2C... S ;
6.1 RESP 0NSIBILITY Cs h i I:.*6.1 L_{he g%nt Superintendent} shall be responsible for overall 9.cQacility> operation and shall delegate in writing the succession to this responsibility during his absence.
-]
I !
---,4 l
; 6.2 ORGANIZATION .
i ;
- 0FFSITE e l
.l. ;
Gr*,,a
-m 6.2.1 Theoffsiteorganizationfor(facilitypanagementandtechnical support shall be as shown on Figure 6 D h O . '. - -. J c- , 2 - I c ,
-$ Sca * < J a FACILITY STAFF
- Js h o -
D 6.2.2 The facility crganization shall be as shown on Figure 6.2-2 and: V a. Each on duty shif t shall be composed of at least the minimum shif t crew composition shown in Table 6.2-1.
.G.) *- At least one licensed Operator shall be in the control room b.
when fuel is in the reactor,
- c. At least two licensed Operators shall be present in the control room during reactor start-up, scheduled reactor shutdown and during recovery from reactor trips.
, -. d. An individual qualified in radiation protection procedures shall
- '4p,i ;
- be on site when fuel is in the reactor.
.ya
- e. ALL CORE ALTERATIONS after the initial fuel loading shall be directly supervised by either a licensed Senior Reactor Operator org Senior Reactor Operator Limited to Fuel Handling who has no other concurrent responsibilities during this operation.
t lO.w. Q
- W 2 m, Id ALL STS 6-1
- i 1
i S C O N IL TI S A T O L S N P E E E R I T M I P IL O L C E A F V E D L A G E L S N S I OI T Y S A L R A E N P A O E V I T W T U P N C R P E E E D D X C S N EI F I E A F R FO R P E 6 H C E G T N A I R N E O N V P A I T S R L A E O P R C C e T v S R I N E I S M D A p R S E E MC O V I T E N T S R T E UES C A M R P OO P L R E OV C E D s
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_j ,. 9 FNERGY SUPPLY MISSION VICE PRESIDENT FUEL PURCHASING
)
SUPER VISOR SYSTEM HEAVY OPERATIONS MAINTENANCE
- DEf%RTMENT DEPARTMENT M INTENDENT SUPERINTENDENT
. k mf DAVIS-BESSE BAYSHORE ACME STEAM SERVICES TECHNICAL m $ I- STATION STATION STATION DEPARTMENT SERVICES 34 _ DEPARTMENT Egj TTI SUPERINTENDENT SUPERINTENDENT SUPERINTEPOENT SUPERINTENDENT SUPERINTENDENT
'D rl C r ii '
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;"cnWr -- O; s')- 2 i
FACILITIES - DEVELOPMENT MIS SION vlCE PREBiDENT SYSTEM P L ANNING GROUP VICE PRESIDENT
)
TRANSMISSION QUAUTY D AVIS - BESSE POWER ENGNEERING AND SUBSTATIONS A SSURANCE STATION 8 CONSTRUCTION GROUP D E PAR TMENT (SEE NOTEI GROUP GENERAL SJPT MANAGER SUPERINTENDENT GEfERAL SUPI nDb > f g T 8S PROJECT POWER PLANT POWER d01 ENGINEERING MANAGEMENT CONSTRUCTION ENGINEERING
$r ! DIVISION DEPARTMENT DEPARTMENT DEPARTMENT 3 _
DIRECTOR MANAGER SUPERINTENDENT c) mc C < 0
;U mF To r- r1 Ol a
-- n:)
o Id g NOTE: THE DAVIS-BESSE STATION 3
] SUPERINTENDENT REPORTS L S DE b
]k ' df( g;) CONTROL SYSTEMS DEPARTMENT POWER SYSTEMS DEPARTME N T TRANSMISSION DEPARTMENT ETV E , UNTIL THE STATION BEGNS Ui (3 SUPERIN TENDENT SUPERINTENDDdT MANAGER COMMERCIAL OPERATION. O .4 He THEN REPORTS TO THE O ENERGY SUPPLY VICE PRESIDENT 2 *3P
~ .
E (1) .
$141I44
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f (11 I (Il C (t) E (lu (Il I"III BaltigeneCE 4f FICI 0Ft94Tien! ItChaltat teGletER ENGl#Ett lh g teslutta suPtevisse SL 3L l (1) (s) E (t) E l (1) (1) l (1) l (4) I "I E II & IIIIII" 1986515G $Nif f A5!1ST ARI "U'""' 581 bit #8ECE L 8 ConWL III'ICI C00881 stilt f0BluAs Eu61stit featuas
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' uiteitssaCE bilt!!#40CE STATItt Blit Ri tt SUPtivillNG & COu 8 TEChelCAL pagialist 004801 87t84108 f0pteau AIIIII8"I TElical REPalpett itECipitlant stellCte[4 CLitt L .
to) (6) AUCL( At PL8EI
'IA ' insipputui E G#400111 inclutta I Tf s SPIR 410 A utCH444 C3
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g I C ColltGE C8&OUA11 ( 10 ' Sl 5tulot OPIBatte Lltinst lusFitil0K Ig guirgggg ENGINEtt t ertR4I84 Litte!!
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.in.isins.resilies 9 mmtion L. ,, ,
. i) artCn: ausses er riistas um.sessematteme en I ' IttilalCl Aus - seassessenemonenfeese
/0 ,. fitting teslitens pg 4 g auxillasits I gm
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'..^'"r.'.e ,- - , , - . , . . , -
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- I I
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V ADrittilSTRATIVE C0flTROLS
^ :>.] %
.s. 6.3 M STAFF QUALIFICATI0tlS "alifications for members of the facil~i f my-tc giecTiTe'd l , by usa of an ovt ' alification state.::se-re g AflSI fil8.1-1971
' i or alternately by specify1m ual position qualifications. Gener-
~
i al ly , the f i rst me* ' . s preferableMqe,cond method is
; adaptable o iose facility staffs requiring special qu37tf+edic:: sge- ~ ^ ~ ~-
; i' a cecat se of a unique organizational structure.
6.3.1 JtRi th Each nember of the faciljitytaff shall meet or exceed the minimun
.i qualifications of AriSI Ill8.M7T for comparable positions, d
6.4 TRAlf1IrlG y..-.-~,.-
/ s Sm tue 6.4.1 or the facility A retraining staff shall be maintained and replacement under the direction trainingof programy!
the position title)- and Q shall neet or exceed the requirements and reconr.iendations of Section 5.5 of AilSI til8.1-1971 and Appendix "A" of 10 CFR Part 55. Y
,T.. ,.i 6.5 REVIEW AfiD AUDIT
' - ~
Cr/% Q- ,] !f'L) 6.5.1 I'~ILITV REVIEW 9%P 4:i F-FullCTI0ri O A 8 / /( .. ,,
. 6.5.1.1 The F::"4 ty Review M -shall function to advise the (T ...t -y . SuperintendentJ an all matters related to nuclear safety.
h." / COMPOSITI0ft
- h. ..
2;. ,) 6 . 5 .1. 2 The 'Fac"iy Review c:up'r shall be composed of the:
- n. . ,
Chairman: -Su-et Superintendenty .
., Mante r: (Operations Supc. , sci ) h Au -
.'^, .3 tiember: (Technical Supccvi:or} _ .- ~ ~ _. 5,,G identer : (Maintenance S c - im t) 4. L.- ... - l t*emt.e r : (.~'I ..:- Instrument and ControT Engineer 7 1 Menber: f ' ' iluclear3 Engineer)
- l Memoer. s. $d$11ealth Physici -
v ALL STS 6-5
!,!, A..D.ti.l_fl_ISTRATIVE CutlTROLS V ll ALTERtlATES S R. c ts..
hC.- 6.5.1.3 Alternate members shall be appointed in writing by the -fffff Chairman to serve on a temporary basis; however, no more than two alternates shall participate in Mactivities at any one time. Cf0 MEETIrlG FRE00EtiCY SE R . ,;. 6.5.1.4 The -fGC,f shall meet at least once per calendar month and as convened by the (Hr)r Chairman. g,2.,
-r 21 "
Gil0Ruti~
$pp 6.5.1.5 A quorum of the 4Ahshall consist of the Chairman and four members including alternates.
RESP 0ftSIRILITIES
; . ., +,_-..
&-) .
6.5.1.6 The (facMi-ty. Review C.xn shall be responsible for: 9
- a. Review of 1) all procedures required by Specification 6.8 and V changes thereto, 2) any other proposed procedures or changes thereto as detemined by the Flaut Superintendent}/to affect
' nuclear safety. @, .* -
- b. Review of all proposed tests and experiments that affect -v"- -
nuclear safety.
- c. Review of all proposed changes to the Technical Specifications. .
g-
- d. Review of all proposed chances or modifications to plant systems or equipment that affect nuclear safety. ;
Investiqation of all violations of the Technical Specifications
- e. -
- and shall prepare and fonaard a report covering evaluation and _ - ^
- recommendations to prevent recurrence to the Mcad:nt Of M '*-- ' ,i
1;..R) and to the Cic;;te of the JCompany Nuclear Review
% i Y' ;]":. r.e: .adi t,r,mg
-- ~ ~ , ;
r %c -'A , , ji f. Review of facility ; operations to detect potential safety hazards. v '
- g. Perfomance of special reviews and investigations and reports thereon as requested by the 0 . A. of the (Company Nuclear - t_g Review nd '.dit ". x @. (Cg y .
p.: - (;;.c 4 l 6-6 m d All. STS
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^
7 l O g o.
..y ADMIflISTRATIVE' C0flTROLS
-u-
, RESP 0flSIBILITIES (Continued)
Ef4Gm
- h. Review of the-P+e*t Security Plan and implementing procedures and '
d shall submit recommended changes to the (d., :;nc of the (Company ' j Nuclear Review u,d * .d ' t ".. -4) . *^ vN
- *M ,
' i. Review of the Emergency Plan and implementing procedures ar.d
; shall submit recommended changes to the%%e of the (Company
' Nuclear Revie,w cad 'd
- Cc;#. N'm-1 14 & ,
1
, AUTHORITY I %N
% 0;nsN 6.5.1.7 The (F:-i" ty Review 1 4) shall:
Os'- .
.J a. Pecommend to the "'n+ Superintendentf written app roval or
".' disapproval of items considered under 6.5.1.6(a) through (d) above.
- b. Render, determinations in writing wi th regard to whether or v not each item considered under 6.5.1.6(a) through (e) above constitutes an unreviewed safety question.
_ & c Provide immediate written noti fication to the
'1f.czh. Ed:. , b, .'
'-Sc^-i^ terd:nt ;f - ).y.
G
^
_... 7 m...a, and the yComp uclear Review 2"d 'A4* "
# 2cvM, y
(. of disagreement between th and the ' - Superin tendent); howe ver , the '- ' ..u Superintendent) shallY~'e"_Fesponsibility for resolution obsagreements pursuant to 6.1.1 above. RECORDS bM CtaL%
;. r 6.5.1.8 The (r;;i ty Review ru m) shall maintain written minutes of - _ , .f .<' ty p( # #' Aeach meeting and copies shall be provided to the 'S~i^trder ; f 7 ;..-: c h,e 161- a.- ~: *] W C' .., ~ , -and "--'^ o f the fCompany Nucl ear Review r d * ;di t " w ' .
esan n w) . i l
','43 ; .;d i L
ALL STS 6-7 l l
I ADMINISTRATIVE CONTROLS V L u_d (CMRB) Q'y' 6.5.2 COMPANY NUCLEAR REVIEW ehdhimELGrouP 'C':"CJ v--- i -athod by which independent review and audit of facility operat is accco ~< hed may take one of several forms. The licensee . itner assign this ..
' to an organizational unit separ + independent from the group having nsibility for faci peratiun or may utilize a standing cormittee compose 'ndiv, s from within and outside the licensee's organization. y Irrespective of tdh used, the licensee s ecify the details .-
ew and audit ; '. of each f' .uonal element provided for the independen pre a as illustrated in the following example specificatioliN _ N FUNCTION W'Y
at $ shall function to 6.5.2.1 The 3 Company Nuclear Review -
provide independent review and audit of designated activities in the h~ areas of: , ' AfhNee.-
- a. nuclear power 9 1 st operations @
g b. nuclear engineering
- c. chemistry and radiochemistry i
- d. metallurgy ,
i;-.
- e. instrumentation and control
- f. radiological safety , , ,
- g. mechanical and electrical engineering
- h. quality assurance practices
- 4. b th;c s p. s iete-W4&nsv. . a zu =ui Ur Um muwc - cr,eracter--
, - : t' ~ n ' % d ;; r ; =:: r p h r.t>
l l . I w I
..-a.
,m, u..
ALL STS 6-8
' : .= ' .
3.. .,. ; - c . ,. ; _- . .. y _ ,,: . ,; ;. -. . . . _ , . .
~.
. ,.:y :- ~ '
.~ K. . . . ~ . . "A'.. :
.;;. ., . . c~- . p .Z?.. . . 2 V
T.b.;t
. N.v' ADMIfl!STRATIVE C0flTROLS
.g 5 - COMPOSITIO'ri css 6.5.2.2 The 'C:,T.A^' shall be composed of the:
.'I', i~ * .
LM.vtm Sc tr: (Position Titic) ,< n n' h M m .; W, e
! Member: (Posi tion Title) 'liu.324.d.c._~. , -E 2. -A~ hco-w . .
(Position Title) 4c a ... - .w.QLJ. 5ib-t fuJ_ "?" I Member: Member: (Position Title) Em. rv. 3- wd W. >. ~ : L:-TC% f
. Member: (Pos,itionTitle)s . .. c .' J _ i .d v rM-tWM:m r.._,,.., o %. m . . . a
~ . w. ' ".~N. s; ~
. - .. . ~.f. : > t .e -
. n.q g s . :. ..._u.
s '. - ..v . i +-W. - : r ~eu-~*"^2 f, ALTERflATES
- 7-;., ,.u, a w ---
6.5.2.3 Alternate members shall be appo'inted in writing' "by the (yy
' 7
.,;st C- mu-W ;. to serve on a temporary basis; however, no more than two alter-natives shall participate in (::irX) activities at any one time.
c:MRCs C0flSULTAtlTS C/.)f S
\/ 6.5.2.4 Consultants shall be utilized as determined by the (CliRAu)
-Ci c ter to provide expert advice to the (C ;IA:).
- Y L}Zem ChlIQ M MEETIflG FREQUErlCY c4R2 6.5.2.5 The tC'") shalLmeet at least once per calendar quarter during the initial year of sfacility, operation following fuel loading and at least once per six mon thereafter. 5 0UORUM , - C.nR 5 - x.u ~ 2d . 6.5.2.6 A quorum of (C'""C) shall consist of the Circ tcr or his
'.'f :*< designated alternate and (a sjHt; ;f tS: C" " ) members including al terna tes. flo more than a minority of the quorum shall have line responsibility for operatio\n of the facility.
1
/,
w i, il l
. .a L
ALL STS 6-9
i ADMINISTRATIVE CONTROLS y. REVIEW
. c p i? B csE t s* -
6.5.2.7 The(C."{JC)shallreview;
- a. The safety evaluations for 1) changes to procedures, equipment or systems and 2) tests or experiments completed under the provision of Section 50.59, 10 CFR, to verify that such actions did not constitute an unreviewed safety question.
l n. . , g b Proposed changes to procedures, equipment or systems which involve an unreviewed safety question as defined in Section {{' - 50.59,10 CFR.
- c. Proposed tests or experiments which involve an unreviewed safety question as defined in Section 50.59, 10 CFR.
- d. Proposed changes in Technical Specifications or licenses,
- e. Violations of applicable statutes, codes, regulations, orders, ij.
Technical Specifications, license requirements, or of internal ' ~ ' procedures or instructions having nuclear safety significance. m -
- f. Significant operating abnormalities or deviations from normal Q and expected performance of plant equipment that affect nuclear sa fe ty.
- g. ABNORMAL OCCURRFNCES, as defined in Section 1.0 of these Technical Specifications.
- p .
!j h. Any indication of an unanticipated deficiency in some aspect
': of design or operation of safety. related structures, systems, Ii oc components. I i.
Oth. Reports and meeting minutes of the F="!ty Review 6?- - 4*ewg4.M 9 . i e . 1 i l
'l! '
f.%. f' ALL STS 6-10
.~ -
e , ,.,: ef.. ...
- f. . . .,... . -,.;,., , . , . . , .
u . ,. .
.. . ..> . . . . , . , . . __ .. , m . , . .,,
V
.a s,,, ADMINISTRATIVE CONTROLS 4_ 4d
.cw a. . l AUDITS g, q ,% 6.5.2.8 Audits of facility activities shall be performed under the cognizance of the (CCA07, These audits shall encompass:
~j j . Aj Fic _ ? a i+
I a. The confomance of facility operation to all provisions contained within the Technical Specifications and applicable license condi-
- tinns at least once per year.
- b. The perfomance, training and qualifications of the entire facility staff at least once per year.
-)
s %a r:4 s e unir
- c. The results of all (ctions taken to correct deficiencies occurring in facility equipment, structures, systems or method of operation thiFaffect nuclear safety at least once per six ae months.
, m, *
.4J
- d. The perfomance of all activities required by the Quality Assurance Program to meet the criteria of Appendix "B",10 CFR 50, at least once per two years.
V d*- L .
- e. The recil; # Emergency Plan and implementing procedures at least g '
once per two years. g f. The'9..+A Inility Security Plan and implementing procedures at least
. ..i once per two years.
- g. Any other area of facility ooperation consider'ed appropriate by the ;C:?.AQ c- the '" ice reci der.t Open t i = }.
3 )3 $ 5. AUTHORITY 3p q .)/. . 4. ' v t '..Y 6.5.2.9 The (GPA&) shall -^^-* +- '" advise.sthe (Vice President :- a e' v 7'* Opa had on those areas of responsibility specified in Section 6).5.2.7 e and 6.5.2.8. , s c .
- g, 'Th ; _ & h'~ . .. A D. h [ < ~.'.b . g _.K ;. .,[
/. *%
+ -l .'
*w ALL STS 6-11
ADMlfilSTRATIVE C0flTROLS . V RECORDS > 4 *'S
. cAjFr v hw. -.
l 6.5.2.10 Records of (C""T) activities shall be prepared, approved and distributed as indicated below: c A.f8
- a. Minutes of each (C:AC) meeting shall be prepared, approved and forwarded to the (Vic; I.cali..; f;c.;tix d within 14 days following each meeting. #' p^ sya W'A d..wd d + _ ,,
- b. Reports of reviews encompassed by Section 6.5.2.7 e, f, 9 and h Q.
W-above, shall be prepared, approved and forwarded to the (W(+ ,* Icc;idc^.t 0;r tic.n} within 14 days following completion of the
' ~
- - ' .d / ,,_.._q review. Je .
- c. Audit reports encompassed by Section 6.5.2.8 above, shall be forwarded to the (i;ce Ti es ;Js.o.-Orm e .un ) and to the manage-ment positions responsible for the areas audited within 30 days af ter completion of the audit. -
(" s A'a,..yt j, , y,. ?- t C. 6.6 ABfl0PJtAL OCCURREflCE ACTI0f( , - Q 6.6.1 The following actions shall be taken in the event of an A8f!0RIML OCCURREllCE:
- a. The Conunission shall be notified and/or a report submitted pursuant to the requirements of Specification 6.9. ,,
- s. '
- b. Each Abnormal OccurrencqR,epprt submitted to the Conmission shall be reviewed by the (TT.CPa_nt submitted to the (C"P,A5-) ~e
( : _ r -. ..ms.~s.. vi i v..c 6 Fiau 4 C A/ E3 , ,, h s
- a. -
e~ ALL STS 6-12 l m
.~ . , . - -
T~.
. o ,
v ADMiti!STRATIVE' C0tlTROLS 6.7 SAFETY LIlllT VIOLATI0ft
~
6.7.1 The following actions shall~ be taken in the event a Safety Limit is violated:
- a. The provisions of 10 CFR 50.36(c)(1)(i) shall be complied with immediately.
- b. The Safety Limit violation shall be reported to the Commission, the ( Lpe,i.. d yui J T .m ? ? hg ) and to the ( ~"'"2 } .. _
inur.ediately. ' ' F~~* J; ?" p~ ,
&c
- c. A Safety Limit Violation Report,T.stpil be prepared. The report shall be reviewed by the (fW).- this report shall describe (1) applicable circumstances preceding the violation, (2) effects of the violation upon facility coir.ponents, systems or structures, and (3) corrective action taken to prevent recurrence.
c;. t) R 2.
. d. The Safety Limit /fiolation Deport shall be submitted to t.he -
, .- Commission, the iC".C) and the (C+_c.aaint ;f %er' - ' '4 v A gf . " r' within 10 days or the violation.
V ' 6.8 PROCEDURES , 6.8.1 Written procedures shall be established, implemented and main-tained covering the activities referenced below:
- a. The applicable procedures reccmmended in Appendix "A" of Regulatory Guide 1.33, flovember,1972.
- b. Refueling operations.
- c. Surveillance and test activities of safety related equipment.
- d. Security Plan implementation.
- e. Emergency Plan implementation.
6.8.2 Each procedure and administrative policy off .8.1 above, and _ . . s changes thereto, shall be reviewed by the IrWrP7thi approved by the _ T~dt.fm ("Mc' Superintendent'} prior to implementation and periodically as set forth in each document. ALL STS 6-13 December 30, 1974
n o e , - ~ .. ADMI';ISTPATI'!E CC T00L,5 - . 6.8.3 Tenoorary changes to procedures of 6.3.1 ateve 93y be ~2cc pecviced:
- a. The intent cf the original procedure is not altered.
- b. The change is 3:croved by t..o ethers of the ]lant i 2l' 2"?
staff; a !cas cr: of . co -aic; a :.2niar R:ac:ce C:eri r.r* : Licence On tr: unit affected. T. .,Q
- c. The change is dccu inted, re ie.:ed b/ :Se i~~0e ar.c at: rove:
" y t h e .' ". _ _ c.J. '.. ..r. . . " ' .' r. '.
w . . . i . . '. r.. .'"'a .. ^ #.
. i -- , l e. .' c '. 2.-* ' " .
%w 6.0 : : S '. . '. ". 1 : :". '.: '. . . " _ '.~. 2 Rr V ~ '. ". : .A ".'_ A .'r. ~."Iu' ^..r.f.4. 3 0 ". " r.
C, C. 0 ^ 3 '. a 6.9.1 Infer aticn to re re:orted to :P2 Cc--issi:n, in additicn to "? repcrts re;uired :/ Ti:12 13. ~;;2 f70darai Te: 'atic.s. s.ta'l ~e i- . accordance witn the 22 ulator.y Fcsiticn in Eevision 3 cf :.e:ulat><v . .. Guide 1.16. "Re:orting of Caerating Inferration - A:pendix "A" Technical Spe:ification:." SPECIAL .EFC1TS c .r. t,; ..c.-.c.. .3 ..,.,.
. _ e en.3nir 3 . " . a. r i r. ~, i n 3 n. .- . . . . . ,
..s..-2 av &.... .z^. .=.
- a. c e.1' .14..._..
~ . . --. 4.,.'.
c2.. .r. .2 3 g g 3. . .. _ . . . 4
. . , . . . .f. 3.: ..
,.J.2 .
for each facility an their a"'
'-d su:mittal are designa:
the Technical C-' ' L .: ens. Exarples of shec qcn recor:acs are in sh~ . c .: . N s% 6.9.2 Spe:ici recorts shall be sa:7itted tc t"e Director of the cegulater,. Operations Ee;icnal Office within the time cerio: stacified for eacn report. These .-3corts shall te :st-itted ccverir; :Pe activi:ies iter.:1-fled belcw cursuant to the requirer.ents of the a:;licable reference specification:
- a. InserviceInspectionProgramRevieus,Specificaticns$4.4.10.lf andf4.4.10.2j.
- b. ECCS Actuation, Specifications (3,.S.2j and {,3.5.3j. _
s_..- - ALL STS - 6-14 January 21, 1976
- 4 1r-- ,
- . . . . - - - . . . . . . . , _ . . ,,_ .. . , , , , . , , _ , , , _ , , , , , . _ ,,, ,, , , , . ,, ] ,
e . .
P U ADMIflISTRATIVE CONTROLS 6.10 RECORD RETEilT10t1 6.10.1 The following records shall be retained for at least five years: *
~,g
- a. Recordsandlogsoffh[ilitjoperationcoveringtimeinterval at each power level.
- b. Records and logs of principal maintenance activities, inspections, repair and replacement of principal items of equipment related to nuclear safety.
- c. ABt10RMAL OCCURREf;CE Reports.
- d. Records of surveillance activities, inspections and calibrations required by these Technical Specifications.
- e. Records of reactor tests and experiments.
- f. Records of changes made to Operating Procedures.
U g. Records of radioactive sisipment.s.
- h. Records of scaled source leak tests and results.
- i. Records of annual physical inventory of all sealed source material of record.
6.10.2._ Ilie following records shall be retained for the duration of the .Eacilit'y) Operating License: - Unir -
- a. Record and drawing changes reflectinga facilitj design mcdifi-cations made to systems and equipment described in the Final Safety Analysis Report.
- b. Records of' new and irraduted fuel inventory, fuel transfers and assembly burnup his, tories.
uno#
- c. Records of(acilityy radiation and contamination surveys.
- d. Records of radiation exposure for all individuals entering radiation control areas.
v ALL STS 6-15 December 30, 1974 l l I
e e~ g e. 1 v.' 4 ADM!illSTRATIVE CONTROLS
- e. Records of gaseous and liquid radioactive material released to the environs.
- f. Records of transient or. operational cycles for those facility components designed for a limited number of transients or cycles.
- g. Records of training and qualification for current member of the plant staff.
- h. ~ Records of in-service inspections performed pursuant to these Technical Specifications.
Rccord of Quality Assurance activities required by the QA ~ i.
. Manual.
- j. Records of reviews performed for changes made to procedures or equipment or reviews of tests and experiments pursuant to '
10 CFR 50.59. . , _ SWC C tJ C 3
- y. k. Records of m9etings of the (feet and the (C""'C).
4 . J I b ALL STS 6-16 December 30, 1974 l f - l t i
- .~.-*. ... W .. ".,;.L'-? :'.
f.,.,:? - [. . : *
...~..".'. _*. ;
. . .. y .
V j */ ADMINISTRATIVE CONTROLS 4:,a
- - . l .1 '. :
6.11 RADIATION PROTECTION PROGRAM Procedures for personnel radiation protection shall be prepared consistent
- with the requirements of 10 CFR Part 20 and shall be approved, maintained and adhered to for all operations involving personnel radiation exposure.
. ' , ,i 6.12 RESPIRATORY PROTECTION PROGRAM 7 :^i" ',
l ALLOWANCE
~'
6.12.1 Pursuant to 10 CFR 20.103(c)(1) and (3), allowance may be made cfh& for the use of respiratory protective equipment in conjunction ,wi activities authorized by the operating license for this (acility in deter-
. .~ a mining whether individuals in restricted areas are exposed td concentra-tions in excess of the limit:. specified in Appendix B, Table I, Column 1, of 10 CFR 20, subject to the following conditions and limitations:
- a. The limits provided in Section 20.103(a) and (b) shall not be exceeded.
" b. If the radioactive material is of such form that intake through .. . - f the skin or other additional route is likely, individual exposures
'J,3 to radioactive material shall be controlled so that the radio-active content of any critical organ from all routes of intake averaged over 7 consecutive days does not exceed that which would result from inhaling such radioactive material for 40 hours at the pertinent concentration values provided in Appendix B, Table I, Column 1, of 10 CFR 20.
- c. For radioactive materials designated "Sub" in the " Isotope" column of Appendix B, Table I, Column 1 of 10 CFR 20, the Ny,4 '
concentration value specified shall be based upon exposure to
^'
;- the material as an external radiation source. Individual expo-sures to these materials shall be accounted for as part of the limitation on individual dose in 520.101. These materials shall be subject to applicable process and other engineering controls.
PROTECTION PROGRAM -y 6.12.2 In all operations in which adequate limitation of the inhalation
- A..d of radioactive material by the use of process or other engineering
'F controls is impracticable, the licensee may permit an individual in a restricted area to use respiratory protective equipment to limit the s inhalation of airborne radioactive material, provided:
- a. The limits specified in 6.12.1 above, 'are not exceeded.
ALL STS 6-17 l
. . s .-
Al)MillISTRATIVE C0llTROLS V PROTECTI0fl PROGRAM (Continued) p.y p. v.w
- b. Respiratory protective equipment is selected and used so that the peak concentrations of airborne radioactive material inhaled by an individual waaring the equipment do not exceed the pertinent concentration values specified in Appendix B Tabla I, Column 1, of 10 CFR 20. For the purposes of this subparagraph, the ,
concentration of radioactive material that is inhaled when respirators are worn may be determined by dividing the ambient
.v.
airborne concentration by the protection factor specified in g3y,
. Table 6.12-1 for the respirator protective equipment worn. If 7;'.
the intake of radioactivity is later determined by other measure-ments to have been different than that initially estimated, the later quantity shall be used in evaluating the exposures,
- c. The licensee advises each respirator user that he may leave the area at any time for relief from respirator use in case of equip-ment malfunction, physical or psychological discomfort, or any ,.
other condition that might cause reduction in the protection ' afforded the wearer.
- d. The licensee maintains a respiratory protective program adequate to assure that the requirements above are met and incorporates Q practices for respiratory protection consistent with those recommended by the American flational Standards Institute (AflSI-Z88.2-1969) . Such a program shall include:
- 1. Air sampling and other surveys sufficient to identify the k -
hazard, to evaluate individual exposures, and to permit proper selection of respiratory protective equipment.
- 2. Written procedures to assure proper selection, supervision, and training of personnel using such protective equipment.
- 3. Written procedures to assure the adequate fitting of respirators; and the testing of respiratory protective equip-ment for operability immediately prior to use. .
- 4. Written procedures for maintenance to assure full effective-ness of respiratory protective equipment, including issuance, ,
cleaning and decontamination, inspection, repair, and storage.
- 5. Written operational and administrative procedures for proper use of respiratory protective equipment including provisions 7.,
for planned limitations on working times as necessitated by operational conditions. Eg
,M .~;-
ALL STS 6-18
... . . . . . ,,.~ , '. ,.,
. .~_
w .a .7 ; . . . p.y . . ., . .
. :.~ . . . . . -:
. .. .; . L . -
~. - .
8 O - ADMINISTRATIVE CONTROLS u'
- 6. Bioassays and/or whole body counts of individuals (and other surveys, as appropriate). to evaluate individual exposures and to assess protection actually provided.
- e. The licensee shall use equipment approved by the.U.S. Bureau of Mines under its appropriate Approval Schedules as set forth in
- Table 6.12-1. Equipment not approved under U.S. Bureau of Mines
{ Approval Schedules shall be used only if the licensee has
; evaluated the equipment and can demonstrate by testing, or on the basis of reliable test information, that the material and . ..j perfonnance characteristics of the equipment are at least equal to those afforded.by U.S. Bureau of Mines approved equipment of the same type, as specified in Table 6.12-1.
- f. Unless otherwise authorized by the Comission, the licensee shall not assign protection factors in excess of those specified
.. in Table 6.12-1 in selecting and using respiratory protective 'T equipment.
REVOCATION V 6.12.3 The specifications of Section 6.12 shall be revoked in their - entirety upon adoption of the proposed change to 10 CFR 20, Section m 20.103, which would make such provisions unnecessary. 6.13 HIGH RADIATION AREA (9M40mm) 6.13.1 In lieu of the " control device" or "alann signal" required by paragraph 20.203(c)(2) of 10 CFR 20: C ,
- a. Each High Radiation Area in which the intensity of radiation is greater than 100 mrem /hr but less than 1000 mrem /hr shall be "Q"t
" i, barricaded and conspicuously posted as a High Radiation Area and entrance t' hereto shall be controlled by issuance of a Radiation Work Pennit and any individual or group of individuals permitted to enter such areas shall be provided with a radiation ranitoring device which continuously indicates the radiation dose rate in the area.
.g b. Each High Radiation Area in which the intensity of radiation is
! .lq greater than 1000 mrem /hr shall be subject to the provisions of
-W 6.13.1(a) above, and in addition locked doors shall be provided to prevent unauthorized entry into such areas and the keys shall s bemaintainedundertheadministrativecontrolofthe3 Shift
" pc ' " on duty.
ALL STS '3 %% 6-19
. f . l ?. f\
> TA _6,f-l
, PROTECTI0fl FACTORS FOR RESPIRATORS g .
. PROTECTI0fl FACTORS 2 GUIDES TO SELECTI0ft 0F EQUIPMErlT
,- PARTICULATES AtiD BUREAU OF MIt(ES APPROVAL SCHEDULES
- VAPORS Afl0 GASES FOR EQUIPMErlT CAPABLE OF PROVIDIrlG AT.
EXCEPT TRITIUM LEAST EQUIVALEllT PROTECTI0ft FACTORS DESCRIPTI0tl MODES 1 0XIDE3 *or schedule superseding for equipment of type' listed o a
. I. AIR-PURIFYIflG RESPIRATORS flfM dgj.V odI./ C; locfR ,?c
. Facepiece , hal f-ma sk *7 flP -B'10 810 30 CIR 3 i4.4(M 4F F[dceniece,
. n I,n . e d 17f 1 ,n e & llylp d NP pp 100 t r. r. o _..
21030 CFR -6 %+(d((4h "E -20 II. ATMOSPHERE-SUPPLYIflG i RESPIRATOR T
- 1. Airline res )irator L .' E! Facepiece, 1al f-mask CF -100 1 oo o 198-30-EFR-5--l-BdW(2[ Tw: C"'
.- Facepiece, full CF -1;400 ( t) Om 1 198-3&-CFR 12.2 e (2pTy- C[5!
Facepiece, full 7 D 100 6 0(10 1^S 30 CIR i 12.2 UT p- Cv'jt) .;? Facepiece, full PD -1.r000 i o no O 19B 30 CFR-6-12:2(c.)(ZL Iype_C(iU.). Ilood CF + 1 0 O 6 6 k.it .- V M- k C[ ,g 5,
, 2. Self-contained breathing apparatus (SCBA)
Facepiece, full 7 D -180 So o t ) 13E 30 CFR M i.4(L)(2)(i) Facepiece, full PD -b000JO o<'o 13 E--30-CFR-6-il . 4 ( b) (3-}f44-) Facepiece, full R -100 lo c m -13E 30-CED-4 11 d (h411._ III. COMBIflAT10tl RESPIRATOR ) Any combination of air- Protection factor 190-GFR S 12.2(e} ne annlicable purifying and atmosphere- for type and mode s chedes--+14si nd ahnve-j supplying respirator of operation as listed above k ,' 1, 2, 3, 4, 5 , r, , 7 [Thes notes are on the following pages] i,. f, . i )
;. '7G
.a, .
[' , a ~
'OT sgr ??.?T
- 4. rt
.]:,l, [-
_ ", ; : . ' ' _' _' ' a . '. .., . . ~-
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- h~["
G. -
;~ v .
4,sJ TABLE 6.9-1 (Continued) 1.. . I See the following symbols: CF: continuous flow D: demand l flP: negative pressure (i.e., negative phase during inhalation) PD: pressure demand (i.e., always l R recirculating (closed circuit) positive pressure)
. W:* F~n.::G. -<. pm.~ :
2(a) For purposes of this specification the protection factor is a measure of the degree of protection afforded by a respirator, defined as the ratio of the concentration of airborne radio-
-- I active material outside the respiratory protective equipment to that inside the equipment (usually inside the facepiece) under conditions of use. It is applied to the ambient airborne
. y, concentration to estimate the concentration inhaled by the u -. wearer according to the following formula:
Ambient Airborne Concentration Concentration Inhaled = Protection Factor V (b) The protection factors apply:
- V (i) only for trained individuals wearing properly fitted respirators used and maintained under supervision in
"'d; a well-planned respiratory protective program.
(ii) for air-purifying respirators only when'high efficiency [above 99.97. removal efficiency by U.S. Bureau of Mines type dioctyl phthalate (DOP) test] particulate filters and/or sorbents appropriate to the hazard are used in atmospheres not deficient in oxygen, dN * (iii) for atmosphere-supplying respirators only when supplied j with adequate respirable air. 3 Excluding radioactive contaminants that present an absorption or submersion hazard. For tritium oxide approximately half of the intake occurs by absorption through the skin so that an overall protection factor of not more than approximately 2 is appropriate wnen atmosphere-supplying respirators are used to protect against tritium oxide. Air-purifying respirators are not recommended for
% use against tritium oxide. See also footnote s, below, concerning
.,1..,.y supplied-air suits and hoods.
4 Under chin type only. .flot reccmmended for use where it might be possible for the amoient airborne concentration to reach instan-taneous values greater than 50 times the pertinent values in
- . Appendix B, Table I, Column 1 of 10 CFR Part 20.
ALL STS 6-21}}