ML20237E997
| ML20237E997 | |
| Person / Time | |
|---|---|
| Site: | Byron, Braidwood  |
| Issue date: | 07/21/1998 |
| From: | COMMONWEALTH EDISON CO. |
| To: | |
| Shared Package | |
| ML20237E991 | List: |
| References | |
| NUDOCS 9809010385 | |
| Download: ML20237E997 (138) | |
Text
{{#Wiki_filter:- _ - _ _ _ 1 ENCLOSURE 1 ITS REVISION I ITS SECTION 3.9 O t l l i l 1 O 9E09010385 980811 PDR ADOCK 05000454 p PDR , f
Boron Con <.entration 3.9.1 l.3'.9 fREFUELIN OPERATIONS
' l. { f c3.9;1: Boron. Concentration
, .LC0' 3.9.1 - Boron concentrations:of the Reactor Coolant System. the refueling canal. and the refueling cavity shall be
. maintained within the limit specified in the COLR.
1 I APPLICABILITY: MODE 6. I f cv , 4 ACTIONS-
-NOTE - - --
;Whi.le this.LC0 is not met.. entry into MODE 6 from MODE 5 is not permitted.
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- CONDITION' ! REQUIRED ACTION. COMPLETION TIME -
. A. . Boron concentration A.1 Suspend CORE. Immediately f] - not'within-limit. ALTERATIONS.
;V AND-A.2 Suspend: positive Immediately-reactivity additions.
AND A'. 3 Initiate action to ~Immediately
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-restore boron-concentration to within limit.
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.N/g;- l r BYRON - UN'ITS 1 &'2- 3.9.1 - 1 7/21/98 Revision I .._ h
Boron Concentration 3.9.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY
. 1 SR - 3.9.1.1 Verify boron concentration is within the 72 hours
' limit specified in the COLR.
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- BYRON - UNITS 1 & 2 3. 9.1 - 2 7/16/98 Revision A 4
Unborated Water Source Isolation Valves 3.9.2
- 3.9, REFUELI'GN OPERATIONS 3.9.2 Unborated Water Source Isolation Valves-LCO 3.9.2 Each valve used'to isolate unborated water sources shall be secured'in the closed position.
APPLICABILITY: MODE 6. ACTIONS
.- NOTE-----
. Separate Condition entry is allowed for each.unborated water source isolation valve.
. CONDITION REQUIRED ACTION COMPLETION TIME A. ---
NOTE- A.1 Suspend CORE Immediately Required Action A.3
.O- ' .must be completed ALTERATIONS.
whenever Condition A AND
.is entered. .
A.2 Initiate actions to Immediately secure valve in One or more valves not. closed position. secured:in closed position.- ., AND 1 i A.3 Perform SR 3.9.1.1. 4 hours : i l p.c i: BYRON . UNITS'l'&'2 , 3.9. 2 - 1 7/16/98 Revision A : L
Unborated Water Source Isolation Valves 3.9.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY 1 SR -3.9.2.1 Verify each valve that isolates unborated 31 days I water sources is secured in the closed position. 1 1 10 V
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s C . BYRON -. UNITS 1 & 2 3.9.2 - 2 7/16/98RevisionI
Nuclear Instrumentation I 3.9.3
'3.9 ' REFUELING OPERATIONS j .3.9.3 Nuclear' Instrumentation LC0 3.9.3 Two source range neutron flux monitors shall be OPERABLE.
APPLICABILITY: MODE 6. ACTIONS-CONDITION ~ REQUIRED ACTION COMPLETION TIME A. One source range A.1 Suspend CORE Immediately neutron flux monitor ALTERATIONS. inoperable AND A.2 Suspend positive Immediately reactivity. additions.. B. Two source range B.1 Initiate action to Immediately neutron flux monitors restore one source inoperable. range neutron flux monitor to OPERABLE status. AND B.2 Perform SR 3.9.1.1. Once per 12 hours i l O . BYRON - UNITS 1.& 2 3.9. 3 - 1 7/16/98 Revision A J
Nuclear Instrumentation 3.9.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY
-SR 3.9.3.1 Perform CHANNEL CHECK. 12 hours SR 3.9.3.2 NOTE -
Neutron detectors are excluded from CHANNEL CALIBRATION. Perform CHANNEL CALIBRATION. 18 months l. i l:
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Containment Penetrations
... 3.9.4 tin 3.9 REFUELING.0PERATIONS-(_)-- l 3, : '3.9.4' Co'ntainment Penetrations-g.-
i LC02 3.9.4 -The containment' penetrations shall be in the following. ni ? status: x lf a.
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One' door in.the' 3ersonnel air lock closed and the ' equipment hatch leid in place by 2 4 bolts:
.b. One door..in the emergency air lock closed: 'and
- c. Each penetration providing direct access from.the containment atmosphere to the outside atmosphere either:
- 1. ' Closed by a manual or automatic isolation valve.
-l' blind fl.ange, or equivalent, or 4 l1
. 2 ._ Capable of being closed by an OPERABLE ' Containment
- Ventilation Isolation System.
-................._.......--NOTE----------------------------
2 Item a. only required when the Fuel _ Handling Building Exhaust Filter Plenum Ventilation System is not in X compliance with LCO 3.7.13. " Fuel Handling Building Exhaust jj : Filter _ Plenum (FHB) Ventilation System." l 1 3.: w ni' DuringCOREjALTERATIONS..
'H (APPLICABILITY: During movement of irradiated -fuel- assemblies within p containment.
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Containment Penetrations 3.9.4 77 ACTIONS V CONDITION REQUIRED ACTION COMPLETION TIME A. One or more A.1 Suspend CORE Immediately containment ALTERATIONS. penetrations not in required statu'.s AND A.2 Suspend movement of Immediately irradiated fuel assemblies within containment. SURVEILLANCE REQUIREMENTS i SURVEILLANCE FREQUENCY i 1 s SR 3.9.4.1 Verify each required containment 7 days penetration is in the required status. f (v) SR 3.9.4.2 Verify each required containment purge 18 months valve actuates to the isolation position on an actual or simulated actuation signal. SR 3.9.4.3 Verify the isolation time of each required In accordance
- containment purge valve is within limits. with the Inservice Testing Program o
V . l BYRON - UNITS 1 & 2 3.9.4 - 2 7/16/98 Revision A l
RHR and Coolant Circulation-High Water Level 3.9.5 3.9 REFUELING OPERATIONS 3.9.5' Residual Heat Removal (RHR) and Coolant Circulation-High Water Level LC0 3.9.5 One RHR loop shall be OPERABLE and in operation. NOTE The required RHR loop may be removed from operation for s 1 hour per 8 hour period, provided no o3erations are ermitted that would cause reduction of t1e Reactor Coolant ystem boron concentration. APPLICABILITY: MODE 6 with the water level 2 23 ft above the top of reactor vessel fluge. ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. RHR loop requirements A.1 Suspend operations Immediately O not met. involving a reduction in reactor coolant boron concentration. AND A.2 Suspend loading Immediately irradiated fuel assemblies in the core. AND A.3 Initiate action to Immediately satisfy RHR loop requirements. AND . (continued) i O ' BYRON - UNITS 1 & 2 39.5 7/16/98 Revision A 9
p. u L RHR and Coolant-Circulation-High Water. Level q 3.9.5 ACTIONS'
- l. ~' CONDITION ' REQUIRED ACTION ~ COMPLETION TIME i-i .
~A. (continued) A.4 Close all containment 4-hours- !
penetrations providing direct l access from ' containment
' atmosphere to outside atmosphere, SURVEILLANCE REQUIREMENTS j
- l4 SURVEILLANCE FREQUENCY 9( . ,
'T ;SR 3.9.5.1 Verif one RHR-loop.is in operation and-
" 12 hours circu ating reactor coolant at a flow rate 7 ..M of-a 1000 gpm.
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RHR and Coolant Circulation-Low Water Level 3.9.6 L3.9, REFUELING OPERATIONS (
3.9.6 Residual
Heat Removalf (RHR) and Coolant' Circulation-Low Water Level LC0 3.;9.6 L Two RHRl loops shall be.0PERABLE, and one RHR loop shall be Lin operation.
. . NOTE-One required RHR loop may be removed from operation-and '
considered OPERABLE:
- a. 'To s'upport filling and draining the reactor cavity when aligned to, or during transitioning to or from, the refueling water storage tank provided the required RHR loop is capable of being realigned to the Reactor Coolant System (RCS): or-b; -To support-required testing provided the. required RHR-loop is capable of being. realigned to the RCS.
APPLICABILITY MODE'6 with the water level .< 23 ft above the top of reactor n- vessel flange. AJ ACTIONS - NOTE
'While this LC0_is'not met, entry into MODE-6 with the water level < 23 ft above the top of'the reactor vessel' flange is not permitted.
CONDITION' - REQUIRED ACTION COMPLETION TIME A. One or more RHR loops A,1 Initiate action to Immediately inoperable, restore RHR loop (s) o to OPERABLE status. DB
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RHR and Coolant Circulation-Low Water Level 3.9.6
, ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.2 Initiate action to Immediately estabhsh a 23 ft of water above the top of reactor vessel flange.
B. No RHR loop in B.1 Suspend operations Immediately operation. involving a reduction in reactor coolant boron concentration. AND B.2 Initiate action to Immediately restore one RHR loop to operation. AND e j B.3 Close all containment 4 hours a penetrations l providing direct access from containment . atmosphere to outside l atmosphere. l l i
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BYRON - UNITS 1 & 2 3.9.6 - 2 7/16/98 Revision A
RHR and_ Coolant Circulation-Low Water Level 3.9;6 {; ' SURVEILLANCE-REQUIREMENTS SURVEILLANCE FREQUENCY d' iSR 3.9.6.1 ' Verify one RHR loop is in operation and 12 hours
? circulating reactor coolant at a flow rate i of a 1000 gpm.
t W SR 3.9.6.2. Verify correct breaker alignment and- 7 days indicated )ower available to the required RHR pump tlat is not.in operation. O L
' BYRON - UNITS 1 8 2 3. 9. 6 - 3 7/21/98 Revision I
o i Refueling Cavity. Water Level;; 3,9.7 3 i REFUELING' OPERATIONS-13.9.71 Refueling Cavity Water Level F
- LCOh3.9.7c
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Refueling cavity' water. level shall be maintained a 23 ft
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-o-above:the top of reactor vessel. flange.
g. M .; APPLICABILITY: During CORE ALTERATIONS' except during latching'and
; l? ' unlatching .of.. control- rod drive shafts.
Y During movement of irradiated fuel assemblies within. containment. ACTIONS CONDITION ~ . REQUIRED ACTION COMPLETION TIME
.-A. Refueling cavity water A.1 ' Sus)end CORE Immediately.
Llevel not within- ALTERATIONS.
- t limit..
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A.2 Suspend movement of Immediately irradiated fuel:
. assemblies ~within containment.
V SURVEILLANCE' REQUIREMENTS SURVEILLANCE . FREQUENCY SR 3.9.7.1 Verify refueling cavity water level is 24 hours a 23 ft above the top of reactor vessel.
' flange.
i L-BYRON - UNITSl-' & 2 3.9.7 - 1 7/17/98 Revision I E-__ _ _ _ _ _ _ _ - - - -
l Boron Concentration B 3.9.1 B 3.9 REFUELING OPERATIONS (v] . 3 B 3.9.1 Baron Concentration BASES BACKGROUND The limit on the boron concentration ensures the reactor remains subcritical during MODE 6. Refueling boron concentration is the soluble boron concentration in the filled portions of the Reactor Coolant System (RCS).. the refueling canal, and the refueling cavity that are hydraulically coupled to the reactor core during refueling. The soluble boron concentration offsets the core reactivity and is measured by chemical analysis of a representative sample of the coolant in each of the volumes. The refueling i boron concentration limit is specified in the COLR. The specified boron concentration is controlled by plant procedures to maintain an overall core reactivity of k 5 0.95 during fuel handling, with control rods and fuel,, assemblies assumed to be in the most adverse configuration (least negative reactivity). ' 3 GDC 26 of 10 CFR 50. Appendix A. requires that two (-) v independent reactivity control systems of different design 3rinciples be 3rovided (Ref.1). One of these systems must 3e capable of 1olding the reactor core subcritical under cold conditions. The Chemical and Volume Control System (CVCS) is the system capable of maintaining the reactor subcritical in cold conditions by maintaining the boron concentration (Ref. 2).
~
. The reactor is brought to shutdown conditions before beginning operations to'open the reactor vessel for
. refueling. After the RCS is cooled and depressurized, the vessel head is unbolted, and removed. The refueling cavity is then flooded with borated water from the refueling water storage tank through the 03en reactor vessel by gravity feeding or by the use of t7e Residual Heat Removal (RHR) !
System pumps. l t I l O BYRON - UNITS 1 & 2 B 3.9.1 - 1 7/16/98 Revision A
Boron Concentration B 3.9.1
.( BASES.
BACKGROUND (continued) The pumping action of the RHR System in the RCS and the natural circulation due to thermal driving heads in the reactor vessel and refueling cavity, ensure adequate mixing of the borated water. The RHR System is in operation during refueling (see LC0 3.9.5. " Residual Heat Removal (RHR) and Coolant Circulation-High Water Level." and LC0 3.9.6. .
" Residual Heat Removal (RHR) and Coolant Circulation-Low Water Level") to provide forced circulation in the RCS and assist in maintaining the boron concentration in the RCS.
the refueling canal, and the refueling cavity above the COLR limit. APPLICABLE During refueling operations, the reactivity condition of the SAFETY ANALYSES core is consistent with the initial conditions assumed for the boron dilution accident in the accident analysis (Ref. 3) and is conservative for MODE 6. The boron concentration limit specified in the COLR is based on the core reactivity at the beginning of each fuel cycle (the end of refueling) and includes an uncertainty allowance. r~~ V3 The required boron concentration and the plant refueling procedures that verify the correct fuel loading plan
] (including full core mapping) ensure that the k of the g core will remain s 0.95 during the refueling op,e,r,ation.
ql Hence, at least a 5% Ak/k margin of safety is established g during refueling. During refueling, all filled portions of the RCS, the water volume in the spent fuel pool, the transfer tube, the refueling canal, the refueling cavity, and th'e reactor vessel form a single mass. As a result, the soluble boron concentration is relatively the same in each of these volumes. The limiting boron dilution accident analyzed occurs in MODE 5 (Ref. 3). A detailed discussion of this. event is provided in Bases B 3.1.1. " SHUTDOWN MARGIN (SDM)." The RCS boron concentration satisvies Criterion-2 of 10 CFR 50.36(c)(2)(ii). O 1 G 1 B RON - UNITS 1 & 2 B 3.9.1 - 2 7/17/98 Revision I 1 l i
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Boron Concentration B 3.9.1 i _' BASES LCO The LCO' requires.that a minimum b'oron concentration be-maintained in all. filled portions of the RCS. the refueling-
' canal, and the refueling cavity, that are hydraulically
. coupled.to the-reactor. core, while-in MODE 6. The boron
-concentration limit specified in the COLR ensures'that'a
. core.ke , of s 0.95 is maintained during' fuel' handling
. operations. ' Violation of the LC0 could lead to an
; inadvertent criticality during MODE 6.
-APPLICABILITY- : This LC0 is applicable in MODE 6 to ensure that the fuel' in- !
'the' reactor vessel will remain .subcritical. The required s0 In MODES 1 and 2 boron with k concentration 2 1.0. LC0 3.1.4.ensures
" Rod a k,'droup.95.
Alicjnment Limits," LC03.1er.5. " Shutdown Bank Insertion Limits.," and LC0 3.1.6.
" Control: Bank ~ Insertion Limits." ensure an adequate' amount of: negative reactivity is available to shutdown the reactor.
In MODE 2 with k,,7 < 1.0 and MODES 3, 4.'and 5. LC0 3.1.1.
" SHUTDOWN MARGIN (SDM).." ensures that an adequate amount of negative reactivity is available to shut down the reactor and maintain.-it subcritical .
p, Al ACTIONS The ' ACTIONS are mo'dified by a Note stating that entry into the MODE 6 from tt0DE 5 is?not permitted while the LCO is not-met. This is an exception to LC0.3.0.4 and precludes .
' detensioning the head when the refueling' boron concentration limit specified :in the COLR is not met.
w . A.11 A.2. and A.3-9-
? ' Continuation of CORE ALTERATIONS or positive reactivity i .U additions (including actions to reduce boron concentration)
;"4 is contingent-.upon maintaining the unit in compliance with the LCO.
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4 BYRON ' UNITS 1~&-2' B 3.9.1 - 3 7/17/98 Revision I l 9
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9 Boron Concentration B 3.9.1 BASES ACTIONS-(continued)
'If the boron concentration of any coolant volume.in the filled portions of the RCS. the refueling canal, or the j7 refueling cavity is less than its-limit, an inadvertent 1 criticality may occur due to an incorrect fuel loading. To
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minimize the potential of an inadvertent criticality resulting from a fuel loading error, all operations-H involving CORE ALTERATIONS and y must be suspended immediately. positive reactivity additions Suspension-of CORE ALTERATIONS and positive reactivity additions shall.not preclude moving a. component to a safe position or norual heatup/cooldown of the coolant volume for the purpose of system temperature control. In addition to immediately suspending CORE. ALTERATIONS and positive reactivity additions, action to restore the boron concentration must be initiated immediately. There are no safety analysis assumptions of boration flow rate and concentration that'must be satisfied. The only requiremen',. is to restore the boron concentration-to its , '.W required value as soon~as possible. In. order to raise the !~ U' boron concentration as soon as possible the operator should begin boration with the best source available for unit i' conditions. Once actions have been initiated.-they must be continued ; until the boron concentration is restored. The restoration ; time de3 ends on the. amount of boron that must be injected to- .
-reach tie required concentration, '
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Boron Concentration B 3.9.1
"'c BASES (O
SURVEILLANCE- SR 3.9.1.1 REQUIREMENTS This SR ensures that the coolant boron concentration in all filled portions of the RCS. the refueling canal, and the refueling cavity, that are hydraulically coupled with the reactor core, is within the COLR limits. The boron concentration of the coolant is determined periodically by chemical analysis. A Frequency of once every 72 hours is a reasonable amount of time to verify the boron concentration of representative samples. The Frequency is based on operating experience, which has shown /2 hours to be adequate to detect slow trends in boror, concentration in these volumes prior to significant reduction. REFERENCES 1. 10 CFR 50. Appendix A. GDC 26.
- 2. UFSAR. Section 9.3.4.
- 3. UFSAR. Section 15.4.6.
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e l a/ I l BYRON - UNITS 1 & 2 B 3.9.1 - 5 7/16/98 Revision A J e , l 1 l l E_ _. _ - _ - _ - - - _ _ _-- 1
Unborated Water Source Isolation Valves B 3.9.2 (] B.3.9 -REFUELING OPERATIONS
.B 3.9.2 Unborated Water Source Isolation Valves
. BASES BACKGROUND During MODE'6 operations, all isolation valves for reactor -
makeup water sources containing.unborated water that are connected to'the Reactor' Coolant System (RCS) must be closed to prevent unplanned boron dilution of the reactor coolant. The isolation valves (CV111B. CV8428. CV8441. CV8435, and CV8439) must be secured in the closed position. The Chemical and Volume Control Systeu is capable of
- supplying borated and unborated water to the RCS through
- various flow paths. Since a positive reactivity addition made by reducing the boron concentration is inappropriate during MODE 6.1 solation of all unborated water sources prevents an unplanned boron dilution.
The Refueling Water Storage Tank (RWST) is assumed to be a boration source. -With the RWST boron concentration not l satisfying these assumptions, the RWST becomes. a potential ir~)
~j dilution source and valves CV112D and CV112E are considered unborated water source isolation valves. These valves must be secured in the closed position.
APPLICABLE The possibility of an uncontrolled boron dilution event SAFETY ANALYSES (Ref.1) occurring during MODE 6 refueling operations .is
. precluded by adherence to this LCO. which requires that potential. dilution sources be isolated. Closing the required valves during refueling operations prevents the flow of unborated water to the filled portion of the RCS, The valves are used to isolate unborated water sources.
These valves have the potential to indirectly allow dilution of the RCS boron concentration in MODE 6. By isolating unborated water sources, a safety analysis for an uncontrolled boron dilution accident in accordance with the Standard Review Plan (Ref. 2) is not required for MODE 6. The RCS unborated water source isolation valves satisfy Criterion 2 of 10 CFR 50.36(c)(2)(ii). i q L.) BYRON - UNITS 1 & 2 B 3.9.2 - 1 7/17/98 Revision I L 4
oil Unborated Water Source Isolation Valves
,# B 3.9.2
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h % % BASES
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'This LCO requires that flow paths to the RCS from unborated H water sources-be isolated!to prevent unplanned boron
- 'r 7 dilution during MODE 6 and thus avoid a reduction in SDM.
APPLICABILITY.. -In MODE 6. this LC0 is applicable'to prevent an-inadvertent boron dilution event by-ensuring ~ isolation of all sources of-
.unborated water to the RCS, For all other MODES. the boron-dilution accident was analyzed and was found to be capable of being mitigated.
1 ACTIONS The' ACTIONS table has been modified by a _ Note that allows separate Condition entry for.each unborated water source isolation valve.- ' A.1. A.2. and A.3 Continuation of CORE. ALTERATIONS is contingent upon maintaining the unit in compliance with this LCO. With any g valve used to isolate'unborated water sources not secured in
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the closed position- all operations < involving CORE ALTERATIONS'must be suspended immediately. The Completion
. Time of "immediately" for performance of Required' Action A.1
'shall not preclude completion of movement of'a component:to
-a safe position.
Preventing inadvertent dilution of the reactor coolant boron
, concentration is dependent'on maintaining the unborated water isolation valves s'ecured closed. Securing the valves
. in the' closed position' ensures _that the valves cannot be inadvertently opened. The Completion Time of "immediately" requires an operator to' initiate-actions to close an open
-valve' and secure the isolation valve in the closed position
.without delay. Once actions are initiated, they must be continued until the valves are secured in the closed
, position. - L p n BYRON _ . UNITS:1 & 2- B 3.9.2 '- 2 7/17/98 Revision I L t
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Unborated Water Source Isolation Valves
, .B 3.9.2' t
Q BASES ACTIONS (continued) Due'to.the potential of having.dilut'ed the boron concentration of.the reactor coolant. SR 3.9.1.1 (verification of boron concentration) must be performed whenever' Condition A is. entered to demonstrate that the-required boron concentration exists. The Completion Time of. 4' hours is sufficient to obtain and analyze a' reactor coolant sample for boron concentration. Condition A has been-modified by a Note to recuire that Required Action A.3 be. completed whenever Concition A is entered; SURVEILLANCE- SR 3 9.2.1 REQUIREMENTS M. These valves ~are to be. secured closed to isolate possible 0 dilution paths. The . likelihood of a significant reduction
% in the. boron concentration during MODE 6 operations is G -
remote due to the large mass of borated water in the-A refueling cavity and the fact that all unborated' water rj 3 - sources are isolated, precluding a dilution. The boron-concentration is checked every 72 hours during MODE 6.under d . SR - 3. 9.1.1.. This SR demonstrates-that valves CV111B, CV8428..CV8441. CV8435, and CV8439 are secured closed by.the-use of~ mechanical stops.' removal of air, or removal of electrical power. Verification of the secured valve . position through a system'walkdown ensures the' isolation of possible dilution paths. . The 31 day Frequency is based on engineering judgment.and is considered reasonable in view of other administrative controls that will ensure that the-
. valve. opening is an unlikely possibility.
REFERENCES 1. UFSAR Section 15.4.6.
- 2. NUREG-0800. Section 15.4.6.
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-BYfkON - UNITS 1 & 2 ' B 3.9.2 - 3 7/17/98 Revision I
j Nuclear Instrumentation l B 3.9.3 .j m. h 'B 3.9 REFUELING OP'ERATIONS; E B 3.9.3- Nuclear Instrumentation-p 'BASESj u q [ L BACKGROUND:. LThe source range' neutron flux monitors are used during'- refueling o condition. perations to monitor the
. The installed-source core range reactivity.
neutron flux monitors 1
' are part of the Nuclear Instrumentation System (NIS). These' l 4 '
detectors are. located external to the reactor vessel and : detect neutrons leaking from the core. 'The use of I D k Jortable detectors is permitted; provided the! _C0 requirements'are met. j o ( The-installed source range neutron flux monitors are boron 1 r :l :trifluoride detectors operating .in the proportional region i
. of the gas' filled detector characteristic curve. The detectors monitor the neutron ~ flux in counts per second. a The instrument range covers six decades (1E+6 cps) with a 7% i instrument accuracy..(Ref,,1). 'The detectors.also. provide
. continuous ~ visual indication in the control.' room to alert
- operato'rs to a possible dilution accident, 'The' NIS-is
. designed in accordance with the criteria presented in dpn ' ~
Reference 2. If used, portable detectors must be- { 2-functionally equivalent to the installed NIS source rang'e ! a' " monitors.
.: APPLICABLE' . .
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zTwo OPERABLE source range neutron flux monitors are recuired 1
; SAFETY; ANALYSES to provide;a1 signal to alert the operator to unexpectec- R changes.in core reactivity such as with a boron dilution ;
accident -(Ref. 3) or an improperly loaded fuel assembly. The need -for a safety analysis.for an uncontrolled' boron dilution accident is eliminated by isolating.all unborated l -water sources.as required by LCO 3.9.2, "Unborated Water ;
.. Source Isolation Valves."
.The source range neutron flux monitors satisfy Criterion 3-i of 10 CFR 50.36(c)(2)(ii).
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EASES LCO This LCO requires that two source range neutron flux monitors be OPERABLE.to ensure that redundant monitoring capability is available to detect changes in core reactivity. To be OPERABLE each monitor must provide visual indication. APPLICABILITY In MODE 6. the source range neutron flux monitors must be OPERABLE to determine changes in core reactivity. There are no other direct means available to check core reactivity. levels. In MODE 2 below the intermediate range neutron flux interlock setpoint (P-6). and in MODES 3. 4. and 5 with the Rod Control System capable of rod withdrawal or with all rods not fully inserted, the installed source range neutron flux monitors are required to be OPERABLE by LC0 3.3.1.
" Reactor Trip System (RTS) Instrumentation."
ACTIONS A.1 and A.2 s With only one source range neutron flux monitor OPERABLE. b) ( redundancy has been lost. Since these instruments are the only direct means of monitoring core reactivity conditions. CORE ALTERATIONS and positive reactivity additions must be suspended immediately. Performance of Required Action A.1 or A.2 shall not preclude completion of movement of a component to a safe position or normal heatup/cooldown of ' the coolant volume for the purpose of system temperature control. B.1 and B.2 With no source range neutron flux monitor OPERABLE there are no direct means of detecting changes in cure reactivity. Therefore, action to restore a monitor to OPERABLE status shall be initiated immediately and continued until a source range neutron flux monitor is restored to OPERABLE status. 1 l l
.O LJ BYRON - UNITS 1 & 2 B 3.9.3 - 2 7/16/98 Revision A l
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Nuclear Instrumentation B 3.9.3 4 IO J BASES
- . ACTIONS (continued)
Since CORE ALTERATIONS and positive reactivity additions are not to be made, the core reactivity condition is stabilized until the source range neutron flux monitors are OPERABLE. This stabilized condition is determined by performing SR 3.9.1.1 to ensure that the required boron concentration I exists. , The Completion Time of once per 12 hours is sufficient to obtain and analyze a reactor coolant sample for boron concentration and ensures that unplanned changes in boron concentration would be identified. The 12 hour Frequency is reasonable, considering the iow probability of a change in core reactivity during this time period. SURVEILLANCE SR 3.9 ~ REQUIREMENTS SR 3.9.3.1 is the performance of a CHANNEL CHECK which is a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the r] assumption that the two indication channels should be a consistent with core conditions. Changes in fuel loading and core geometry can result in significant differences between source range channels, but each channel should be consistent with its local conditions. The Frequency of 12 hours is consistent with the CHANNEL CHECK Frequency specified similarly for the same instruments in LCO 3.3.1. j ( SR 3.9 3.2
]
SR 3.9.3.2 is the 3erformance of a CHANNEL CALIBRATION every g 18 months. This S1 is modified by a Note stating that
~
o neutron detectors are excluded from the CHANNEL CALIBRATION. 4 The CHANNEL CALIBRATION for the source range neutron flux 4 f monitors consists of obtaining the detector discriminator y curves, evaluating those curves, and comparing the curves to the manufacturer's data. The 18 month Frequency is based on kj the need to perform this Surveillance under the conditions tp that ap)1y during a plant outage. Operating experience has shown t1ese components usually pass the Surveillance when l performed at the 18 month Frequency.
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BYRON - UNITS 1 & 2 B 3.9.3 - 3 7/17/98 Revision I L_________.____ _ _ _ . _ _ _ . . _ _ . . _ _ _ _ _ _ .
Nuclear Instrumentation B 3.9.3
- /~'i -BASES (f
REFERENCES _ 1. UFSAR. Table 7.5-2.
- 2. 10 CFR 50. Appendix A. GDC 13. GDC 26. GDC 28. and GDC 29.
- 3. UFSAR. Section 15.4.6.
l i I I k BYRON - UNITS 1 & 2 B 3.9.3 - 4 7/16/98 Revision A
g . Containment Penetrations B 3.9.4 B 3.9 ' REFUELING OPERATIONS' l B 3.9.4 Containment Penetrations. BASES-BACKGROUND During CORE ALTERATIONS or movement of irradiated fuel assemblies within containment, a release of fission product radioactivity within containment.will be restricted from escaping to the environment when the LCO requirements'are met. In MODES 1. 2. 3. and 4. this is accomplished by maintaining containment OPERABLE as described in LC0 3.6.1.
" Containment;"' In MODES 5 and 6. the potential for containment pressurization as a result of an accident is not elikely: therefore. requirements to isolate the containment from the outside atmosphere can be less stringent. The LCO: requirements are referred to as " containment closure" {
rather than " containment OPERABILITY." Containment closure l means that all potential escape paths are filtered, closed. !
.or capable of being. closed. Since there is no significant i potential for containment _3 pressurization, the 1
~ 10 CFR 50, Appendix J. lea cage criteria and testsi are'not j required.
t
- t The containment serves to contain fission product
-radioactivity that may be released from the reactor core following an accident, such that offsite radiation exposures are maintained well within the requirements of 10 CFR 100.
In addition, the containment provides radiation shielding
. from the -fission products that may be present in the
' containment: atmosphere following accident conditions.
[ . I l BYRON -LUNITS 1.& 2: B 3.9.4 - 1 7/16/98 Revision A um______.__________.._________m.._. _ _ _ _ _ _ _ _ _ _ _ _ . _ . _ _ _ _ _ . _ _ _ _ _ _ _ . _ _ _ . _ . . _ _ _ _ . _ _ _ _ _ _ . _ _ _ . . . _ _ . _ . _ _ _ _ _ _ _ ._ _. _ _ _ . _ _ _ . _ _ _ , .
i Containment Penetrations B 3.9.4 (] v BASES BACKGROUND (continued) l The containment equipment hatch, which is part of the containment pressure boundary, provides a means for moving large equipment and components into and out of containment. During CORE ALTERATIONS or movement of irradiated fuel assemblies within containment with the equipment hatch installed, the equipment hatch must be held in place by at least four bolts. Good engineering practice dictates that
.the bolts be approximately equally spaced. During CORE I l ALTERATIONS or movements of irradiated fuel assemblies i' within containment and the equipment hatch not intact, the i OPERABILITY requirements of the Fuel Handling Building Exhaust Filter Plenum (FHB) Ventilation System must be met.
The OPERABILITY requirements of the FHB Ventilation System , are provided in LCO 3.7.13. " Fuel Handling Building Exhaust Filter Plenum (FHB) Ventilation System." The containment air locks. which are also part of the containment pressure boundary, provide a means for personnel access during MODES 1. 2. 3. and 4 in accordance with LC0 3.6.2. " Containment Air Locks." The two air locks are the personnel air lock and the emergency air lock. Each air ' pT lock has a door at both ends. The doors are normally i
- t. interlocked to prevent simultaneous opening when containment '
OPERABILITY is required. During periods of unit shutdown when containment closure is not required, the door interlock mechanism may be disabled, allowing both doors of an air lock to remain open for extended aeriods when frequent containment entry is necessary. Juring CORE ALTERATIONS or ' movement of irradiated fuel assemblies within containment. containment closure is required: therefore, the door
, interlock mechanism may remain disabled, but one air lock door must always remain closed. An exception, however. is
. 3rovided for the personnel air lock. It is acceptable to 1 ave both doors of the personnel air lock opened simultaneously provided the FHB Ventilation Syst.em is in compliance with LCO 3.7.13.
The closure restrictions are sufficient to restrict unfiltered fission product radioactivity releases from containment to the environment due to a fuel handling accident during refueling. ] l ) o BYRON - UNITS 1 & 2 B 3.9.4 - 2 i 7/16/98 Revision A l l l l
Containment Penetrations I
, B 3.9.4 BASES BACKGROUND (continued) j 1
The Containment Ventilation Isolation System consists of the normal purge subsystem, the mini purge subsystem. and the post Loss Of Coolant Accident purge subsystem. These three ! subsystems contain penetrations which provide direct access ' from the containment to the outside atmosphere. In MODE 6. the minipurge subsystem is normally used to exchange large volumes of containment air to support refueling operations. i Each penetration contains inside and outside containment j isolation valves which close automatically on an actuation signal. During CORE ALTERATIONS or movement of irradiated fuel within containment, all required valves within a j subsystem must be capable of being closed by a containment j ventilation isolation signal whenever the' associated i subsystem is in operation. A list of the instrumentation l which functions to isolate the valves in these penetrations I is provided in LCO 3.3.6. " Containment Ventilation Isolation Instrumentation." The other containment penetrations that provide direct access from containment atmosphere to outside atmosphere must be isolated on at least one side. Isolation may be achieved by a closed automatic isolation valve. a manual 1 ( isolation valve, blind flange, or equivalent. Equivalent isolation methods allowed under the provisions of 10 CFR 50.59 may include use of a material that can provide ! a temporary atmospheric pressure ventilation barrier during ! CORE ALTERATIONS or movement of irradiated fuel within the - 4 containment. I i i l O . BYRON - UNITS 1 & 2 83.9.4-3 7/16/98 Revision A
i Containment Penetrations B 3.9.4 l A BASES. V APPLICABLE During CORE ALTERATIONS or movement of irradiated fuel i SAFETY ANALYSES assemblies within containment. the most severe radiological ! consequences result from a fuel handling accident. The fuel ! handling accident is a postulated event that involves damage l to irradiated fuel (Ref.1). Fuel handling accidents. 3 analyzed in Reference 2. include dropping a single l irradiated fuel assembly and handling tool or a heavy object i onto other irradiated fuel assemblies. The requirements of l LC0 3.9.7. " Refueling Cavity Water Level." and the minimum ' decay time of 100 hours prior to CORE ALTERATIONS ensure that the release of fission product radioactivity, subsequent to a fuel handling accident in containment. , results in doses that are well within the guideline values i specified in 10 CFR 100. Reference 2 defines "well within" ! 10 CFR 100 to be 25% or less of the 10 CFR 100 values. The { acceptance limits.for offsite radiation ex30sure for the j fuel handling accident will be 25% of 10 C R.100 values or l the NRC staff approved licensing basis (e.g. , a specified l fraction of-10 CFR 100 limits). < Containment penetrations satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii). O j LCO This LCO limits the consequences of a fuel handling accident i in containment by limiting the potential escape paths for j fission product radioactivity released within containment. ! The LC0 requires any penetration providing direct access i from the containment atmosphere to the outside atmosphere to ; be closed except for the OPERABLE containment purge.(supply l and exhaust) penetrations. For the OPERABLE containment ; purge penetrations'.-this LCO ensures that unisolated i penetrations are isolable by the Containment Ventilation ' Isolation System. The OPERABILITY requirements for this LC0 ensure the automatic purge valve closure times specified i in'the UFSAR can be achieved and. therefore, meet the assumptions used in the safety analysis to ensure that l releases through the valves are terminated, such that , radiological doses are within the acceptance limit. ' The LC0 is modified by a Note which allows both personnel air lock' doors to be open or the equipment hatch not intact l When the FHB Ventilation System is in compliance with LC0 3.7.13. When the equipment hatch is installed it serves to contain fission product radioactivity that may be released following a fuel handling accident in the O ' BYRON - UNITS 1 & 2 B 3.9.4 -4 7/16/98 Revi.sion A
Containment Penetrations B 3.9.4 (3 BASES
%-)
LC0 (continued) I containment. When the equipment hatch is not intact. or when both doors of the personnel air lock are simultaneously 1 opened, the internal containment pressure is essentially equal to the internal pressure of the fuel handling building. In the event of a fuel handling accident in the containment, realigning of the fuel handling building i ventilation system creates a negative pressure in the containment and fuel handling building relative to the j auxiliary building and outside atmosphere. The negative pressure ensures that any radioactivity released to the containment atmosphere will either remain in the containment or be filtered through a FHB Ventilation System train. As such, with the equipment hatch not intact, or with both 3ersonnel air lock doors open. the consequences of a fuel landling accident in containment would not exceed those calculated for a fuel handling accident in the fuel handling building. APPLICABILITY The containment penetration requirements are applicable during CORE ALTERATIONS or movement of irradiated fuel assemblies within containment because this is when there is ON a potential for a fuel handling accident. In MODES 1. 2. 3. ! and 4, containment penetration requirements are addressed by ' LCO 3.6.1. In MODE 5, and in MODE 6 when CORE ALTERATIONS or movement of irradiated fuel assemblies within containment are not being conducted, the potential for a fuel handling ! accident does not exist. Therefore, under these conditions no requirements are placed on containment penetration status, i i
. BYRON - UNITS 1 & 2 B 3.9.4 - 5 7/16/98 Revision A
( l
! Containment Penetrations B 3.9.4 BASES ACTIONS A.1 and A.2 If the containment equipment hatch, air lock doors. or any containment penetration that provides direct access from the containment atmosphere to the outside atmosphere is not in { 1 the required status, the unit must be placed .in a condition where containment closure is not need2d. This is accomplished by immediately suspending CORE ALTERATIONS and movement of irradiated fuel assemblies within containment. Performance of these actions shall not preclude completion of movement of a component to a safe position. SURVEILLANCE SR 3.9.4.1 REQUIREMENTS This Surveillance demonstrates that each of the containment penetrations required to be isolated is isolated. This Surveillance for the open purge valves demonstrates that the valves are not blocked from closing. Also the Surveillance will demonstrate that each valve operator has motive power which will ensure that each valve is capable of being closed by an OPERABLE automatic Containment Ventilation Isolation ( signal. The Surveillance is performed every 7 days during CORE ALTERATIONS or movement of irradiated fuel assemblies within containment. The Surveillance interval is selected to be commensurate with the normal duration of time to complete i fuel handling operations. As such. this Surveillance ' ensures that a postulated fuel handling accident that releases fission product radioactivity within the containment will not result in a release of fission product radioactivity to the environment. i 4 BYRON - UNT'S 1 & 2 B 3.9.4 - 6 7/16/98 Revision A
Containment Penetrations n , .B 3.9.4
] BASESi 1 SURVEILLANCE REQUIREMENTS (continued)'
m
,m , .SR 3.9.4.2
~
4 This Surveillance. demonstrates that each re uired containment purge valve actuates to its; iso ation position
'on an actual or simulated high radiation signal. The 18 month Frequency maintains consistency with other similar Engineered Safety. Feature Actuation System instrumentation and valve-testing-requirements. .In LCO 3.3.6. the 4 .
. Containment Ventilation Isolation instrumentation requires a i
w. CHANNEL' CHECK every 12 hours and a COT every.92. days to.
- ensure the channel OPERABILITY during refueling o erations.
- g. Every 18 months a CHANNEL' CALIBRATION is performe 4 SR 3.9 4.3 demonstrates that the. isolation time of'each
-valve'is in accordance with the Inservice' Testing Program.
- H requirements. These Surveillance performed during. MODE 6
? :
will ensure that the valves are capable of closing after a _ postulated fuel handling accident to limit a release of fission product radioactivity from the containment. SR 3.9.4.3 fi This Surveillance demonstrates'that'the isolation time'of (,i - each required containment purge valve providing direct access from the containment atmosphere to the'outside atmosphere is in accordance with the' Inservice Testing' Program requirements. This SR along with SR 3.9.4.2. ensures the containment purge valves in penetrations which-provide direct access from the containment atmosphere to the. outside. atmosphere are capable of. closing after a postulated ifuel. handling accident to limit the release of fission product radioactivity from the containment. REFERENCES 1. UFSAR. Section 15.7.4. f 2. NUREG-0800. Section 15.7.4. Rev. 1. July 1981. 4 L fy , aQ , L .- BYRON'- UNITS l'& 2: B 3.9.4 - 7 7/17/98 Revision I L+
.-t
E , t RHR and Coolant' Circulation-High Water. Level B 3.9.5 B~3 9; REFUELING OPERATIONAL B 3.9.5 Residual Heat Removal :(RHR) and Coolant Circulation-High Water l Level l BASES. ! BACKGROUND -The purpose of the RHR System in MODE 6 is to remove decay heat and. sensible heat from the Reactor Coolant System l (RCS) as required by GDC 34; to provide mixing of borated L coolant and to prevent' boron. stratification (Ref.1). Heat
.is removed from the RCS by circulating reactor coolant through the RHR heat exchanger (s), where the heat is transferred to the Component Cooling Water System. The-coolant is then returned to the RCS via the RCS. cold leg (s).
f Operation of the RHR System for normal cooldown or decay L heat removal is manually accomplished from the control room. The heat removal rate.is adjusted by controlling the flow of reactor coolant through the RHR heat exchanger (s) and bypass line(s). . Mixing of the reactor coolant is maintained by i this continuous circulation of reactor coolant through the RHR System. APPLICABLE While there is no explicit analysis assumption for the decay l SAFETY ANALYSIS heat removal function of the RHR System in MODE 6. if the-reactor coolant temperature is not maintained below 200 F, L boiling of the reactor coolant could result. This could i lead to a=1oss of coolant in the reactor vessel. In i addition boiling of.the reactor coolant could lead to a L reduction in boron. concentration in the coolant due to boron l plating out on components near the areas of the-boiling activity. ,The loss of. reactor coolant and the reduction'of boron concentration in the reactor coolant would eventually , challenge the integrity of the fuel cladding, which is a i fission product barrier. One train of the RHR System is required to be OPERABLE and in operation in MODE 6. with the water level 2 23 ft above the top of the reactor vessel l -flange. to prevent this challenge. The LC0 does permit l' de-energizing the RHR pump for short durations, under the L condition that the boron concentration is not reduced.. This conditional de-energizing of the RHR pump does not result in a challenge to the' fission product barrier. RHR and Coolant Circulation-High Water Level satisfies Criterion 4 of 10 CFR 50.36(c)(2)(ii). I p. L BYRON --UNITS 1 & 2 , B 3.9.5 - 1 7/16/98 Revision A n
RHR and Coolant Circulation-High Water Level B 3.9.5 [') v
. ' BASES LCO Only one_RHR* loop is required for decay heat removal in MODE 6 with the water level 2: 23 ft above the top of the reactor vessel flange because the volume of water above the reactor vessel flange provides backup decay heat removal 1
capability. One RHR loop is required to be in operation and OPERABLE to provide: r3
? a. Removal of decay heat:
W in d b. Mixing of borated coolant to minimize the possibility ; of criticality; and
- c. Indication of reactor coolant temperature.
An OPERABLE RHR loop includes an RHR pump, a heat exchanger valves, piping, instruments, and controls to ensure an OPERABLE flow path. The flow path starts in one.of the RCS hot legs and is returned to the RCS cold legs. The LCO is modified by a Note-that allows the required operating RHR loop to be removed from service for up to 1 hour per 8 hour period, provided no operations are permitted that would cause a reduction of the RCS boron
.(._n) concentration. Boron concentration reduction is prohibited becauseu~niform concentration distribution cannot be ensured '
without forced circulation. This permits operations such as core ma] ping or' alterations in the vicinity of the reactor
. vessel lot leg nozzles and RCS to RHR isolation valve testing. During this 1 hour period, decay heat is removed by natural convection to the large mass of water in the refueling cavity.
J i l I
)
I x; BYRON - UNITS 1 & 2 B 3.9.5- 2 7/17/98 Revision I
RHR and Coolant Circulation-High Water Level B 3.9.5 A BASES U APPLICABILITY One RHR loop must be OPERABLE and in operation in MODE 6, with the water level a 23 ft above the top of the reactor vessel flange, to provide decay heat removal and mixing of the borated coolant. The 23 ft water level was selected because it corresponds to the 23 ft requirement established for fuel movement in LCO 3.9.7. " Refueling Cavity Water Level." Requirements for the RHR System in MODES 1, 2. 3.
- 4. and 5 are covered by LCO 3.4.6. "RCS Loops-MODE 4."
LC0 3.4.7. "RCS Loops -MODE 5. Loops Filled." LCO 3.4.8.
"RCS Loops-MODE 5.' Loops Not Filled." LCO 3.5 2.
"ECCS -Operating. " and LCO 3.5.3 "ECCS - Shutdown. " RHR loop requirements in MODE 6 with the water level < 23 ft are located in LC0 3.9.6. " Residual Heat Removal (RHR) and Coolant Circulation-Low Water Level . "
ACTIONS A.1. A.2. A.3. and'A.4 If RHR loop requirements are not met, there will be no forced circulation to provide mixing to establish uniform boron concentrations. Reduced boron concentrations'can
, , , occur by the addition of water with a lower boron ;] ' concentration than that contained in the RCS. Therefore, actions that could result in a reduction in the coolant boron concentration must be suspended immediately.
With no forced circulation cooling, decay heat removal from the core occurs by natural convection to the heat sink provided by the water above.the core. A minimum refueling
' water level of 23 ft above the reactor vessel flange
, provides an adequate available heat sink. Suspending any operation that would increase decay heat load, such as
. loading a fuel assembly, is a prudent action under this condition. Therefore, actions shall be taken immediately to suspend loading of irradiated fuel assemblies in the core.
Suspension of these activities shall not preclude completion of movement of a component to a safe condition. With the unit in MODE 6 and the refueling water level a 23 ft above the top of the reactor vessel flange, removal of decay heat is by ambient losses only. Therefore. corrective actions shall ba initiated immediately and shall continue until the RHR loop requirements are met. (~) V BYRON - UNITS 1 & 2 B 3.9.5 - 3 7/16/98 Revision A 9
RHR and Coolant Circulation-High Water Level B 3.9.5
/ BASES ACTIONS (continued)
With the RHR loop requirements not met, the potential exists for the coolant to boil and release radioactive gas to the containment atmosphere. Therefore, all containment penetrations providing direct access from the containment atmosphere to the outside atmosphere must be closed within 4 hours. Closing containment penetrations that are open to the outside atmosphere ensures dose limits are not exceeded. j The Com;letion Time of 4 hours is reasonable, based on the ! low pro] ability of the coolant boiling in that time. j l SURVEILLANCE SR 3.9.5.1 . REQUIREMENTS This Surveillance demonstrates that the RHR loop is in operation and circulating reactor coolant. The flow rate is determined by the flow rate necessary to provide sufficient decay heat removal capability and to provide mixing of the 1 borated coolant to prevent thermal and boron stratification in the core. The Frequency of 12 hours is sufficient. considering the flow, temperature, pump control, and alarm O indications available to the operator in the control room U for monitoring the RHR System. l REFERENCES 1. UFSAR Section 5.4.7 t BYRON - UNITS 1 & 2 B 3.9.5 - 4 7/16/98 Revision A
RHR and Coolant Circulation-Low Water Level B 3.9.6 B 3.9 REFUELING OPERATIONS B 3.9.6 Residual Heat Removal (RHR) and Coolant Circulation-Low Water Level BASES BACKGROUND The purpose of the RHR System in MODE 6 is to remove decay heat and sensible heat from the Reactor Coolant System . (RCS), as required by GDC 34, to provide mixing of borated coolant and to prevent boron stratification (Ref.1). Heat is removed from the RCS by circulating reactor coolant through the RHR heat exchangers where the heat is transferred to'the Component Cooling Water System. The coolant is then returned to the RCS via the RCS cold leg (s). Operation of the RHR System for normal cooldown decay heat removal is manually accomplished from the control room. The heat removal rate is adjusted by controlling the flow of reactor coolant through the RHR heat exchanger (s) and bypass line(s). Mixing of the reactor coolant is maintained by 5 this continuous circulation of reactor coolant through the RHR System. O APPLICABLE SAFETY ANALYSIS While there is no explicit analysis assumption for the decay but removal function of the RHR System in MODE 6, if the reactor coolant temperature is not maintained below 200 F, t;okily of the reac+.or coolant could result. This couid leaa to e loss of coolant in the reactor vessel. In addition, boiling of the reactor coolant could lead to a reduction in boron concentration in the coolant due to the boron plating out on components near the areas of the boiling activity. The loss of reactor coolant and the reduction of boron concentration in the reactor coolant will eventually challenge the integrity of the fuel cladding, which is a fission product barrier. Two trains of the RHR System are required to be OPERABLE, and one train in operation, in order to prevent this challenge. RHR and Coolant Circulation-Low Water Level satisfies Criterion 4 of 10 CFR E0.36(c)(2)(ii). O BYRON - UNITS 1 & 2 B 3.9.6 - 1 7/16/98 Revision A
' RHR and. Coolant Circulation-Low Water. Level-B 3.9.6
} BASES =
LCO. Both.:RHR loops 1must be OPERABLE in MODE 6, with the water level < 23 ft above the. top of the reactor vessel flange. LIn addition, one RHR loop must be in operation.in order to
= provide:
- ra J9 a. Removal of: decay. heat:
-e M b. ; Mixing of borated coolant to minimize the possibility
'** of-criticality: and-q %@ c. Indication of reactor coolant. temperature.
An' OPERABLE RHR loo
' exchanger:, valves, p consists of an RHR pump, a heat-piping. instru ensure an OPERABLE-flow path. -The flow 3ath-starts in one of the RCS. hot legs and is' returned to tie.RCS cold legs.
However. the LCO is modified by a Note that permits the required RHR loop to be removed from operation and-considered OPERABLE when aligned to, or.during transitioning to or from.- the Refueling Water Storage Tank (RWST) to.
. support filling or draining the' refueling cavity, or.to support required testing, if capable of being. realigned to the RCS.
APPLICABILITY _ Two RHR loops are required to be OPERABLE, and.one RHR loop
, must be in operation in MODE 6, with.the water level: < 23 ft
' above the top of the reactor vessel flange. to provide decay heat removal and mixing of the borated coolant.
Requirements for-the RHR System in MODES 1, 2, 3. 4. and 5
> are covered by' LC0 3.4.6. "RCS Loops -MODE :4," LC0 3.4.7.
"RCS Loops -MODE 5. Loops Filled." LC0.3.4.8. "RCS
' Loops -MODE 5. Loops Not Filled." LC0 3.5.2, "ECCS-Operating," and LC0 3.5.3.' "ECCS-Shutdown." RHR-loop requirements'in MODE 6 with the water level a 23 ft are located in LCO 3.9.5. " Residual Heat Removal (RHR) and Coolant Circulation-High Water. Level."
.1 I
h BVRONi- UNITSt1 & 2. B 3,9.6 - 2 7/17/98 Revirion I-m , i
~
RHR and Coolant Circulation-Low Water Level B 3.9.6
? BASES-ACTIONS The ACTIONS are modified by a Note stating that entry into
.the Applicability-is not permitted while the LCO is not met.
This is.an exce) tion to LCO 3.0.4 and arecludes transition
'into-MODE 6 wit 1 water level < 23 ft w111e the LCO is not met.
A.1 and A.2 With one or more RHR. loops inoperable, the RHR System may not be capable of removing decay heat.and mixing the borated coolant. Therefore.. action shall be immediately initiated and continued until the required number of RHR loops are' restored to OPERABLE status or until.a 23 ft of water level is established abovc: the reactor vessel flange. When-the-water level-is a 23 ft above the reactor vessel flange, the
-Applicability changes to that of LC0 349.5, and only one RHR
. loop'is requi-red to be 0PERABLE and. in operation. An
, immediate' Completion Time.is necessary for an operator to initiate corrective actions.
B.1 B.2. and B.3 e1 If no RHR loop'is in operation there will be no forced
~
~
circulation to provide mixing to establish uniform boron concentrations. . Reduced boron concentrations can occur by L the addition of water with a lower boron concentration than L that contained in the RCS. Therefore actions-that would-
-result in a reduction in the coolant boron-concentration
- j. .
must be suspended immediately. E ;In addition, with no forced circulation, any decay heat removal occurs by ambient losses only. Therefore, action L' shall be initiated immediately to restore one RHR loop to L . caeration. .Once initiated, actions shall continue until one p RiR loop has been restored to operation. o L, L ( BYRON - UNITS 1 & 2 B 3.9.6 - 3 7/17/98 Revision I~
RHR and Coolant Circulation-Low Water Level B 3.9.6 BASES ACTIONS (continued) i With no RHR loop in operation the potential exists for the l coolant to boil and release radioactive gas to the containment atmosphere. Therefore, all containment penetrations providing direct access from the containment atmosphere to the outside atmosphere must be closed within 4 hours. Closing containment penetrations that are open to the outside atmosphere ensures that dose limits are not exceeded. The Com)letion Time of 4 hours is reasonable, based on the low pro] ability of the coolant boiling in that time. I SURVEILLANCE SR 3.9.6.1 REQUIREMENTS This Surveillance demonstrates that one RHR loop is in operation and circulating reactor coolant. The flow rate is determined by the flow rate necessary to provide sufficient decay heat removal capability and to provide mixing of the borated coolant to prevent thermal and boron stratification in the core. The Fre considering the flow quency of 12 pump temperature, hourscontrol, is sufficient, and alarm indications available to the operator for monitoring the RHR , System in the control room. l SR 3.9.6.2 Verification that the required pump is OPERABLE ensures a RHR pump can be placed in operation, if needed, to maintain
, decay heat removal and borated coolant circulation. 4 Verification is performed by verifying proper breaker i alignment and power available to the pump. The Frequency of '
7 days is considered reasonable in view of other administrative controls available and has been shown to be acceptable by operating experience. REFERENCES 1. UFSAR. Section 5.4.7. l l O V BYRON - UNITS 1 & 2 B 3.9.6 -4 7/16/98 Revision A w _ _-_ _ ___-_____-_
J Refueling Cavity Water Level B 3.9.7 {; B 3.9 REFUELING OPERATIONS B 3.9.7 Refueling Cavity Water Level
. BASES
. BACKGROUND The movement of irradiated fuel assemblies within containment or performance of CORE ALTERATIONS except during latching and unlatching .of control rod drive shafts, requires a minimum water level of 23 ft above the top of the reactor vessel flange. This requirement ensures a sufficient level of water is maintained in the refueling cavity to retain iodine fission product activity resulting from a fuel handling accident in containment (Refs. 1 and 2). Sufficient iodine activity would'be retained to limit offsite doses from the accident to < 25% of 10 CFR 100 limits. as provided by the guidance of Reference 3.
APPLICABLE During CORE ALTERATIONS and movement of irradiated fuel SAFETY ANALYSES assemblies the water level in the refueling cavity is an ed initial condition design parameter in the analysis of a fuel o? handling accident in containment, as postulated by C- Regulatory Guide 1.25 (Ref. 1). A minimum water level of (~) C 23 ft (Regulatory Position C.1.c of Ref. 1) allows a decontamination factor of 100 (Regulatory Position C.1.g of d Ref.1) to be used in the accident analysis for iodine. This relates to the assumption that 99% of the total iodine released from the pellet to cladding gap of all the dropped fuel assembly rods is retained by the refueling cavity water. The fuel pellet to cladding gap is assumed to contain 10% of the total fuel rod iodine and noble gas inventory, with the exception of 30% for Kr-85 (Ref. 2). The fuel handling accident analysis inside containment is described in Reference 2. With a minimum water level of 23 ft and a minimum decay time of 100 hours prior to fuel l handling, the analysis and test programs demonstrate that the iodine release due to a postulated fuel handling accident is adequately captured by the water and offsite doses are maintained within allowable limits (Refs. 4 ! and 5). -1 Refueling cavity water level satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).
, l
'V> !
BYRON - UNITS 1 & 2 B 3.9.7 - 1 7/17/98 Revision I N_-
Refueling Cavity Water Level B 3.9.7 BASES d(' l LCO A minimum refueling cavity water level of 23 ft above the reactor vessel flange is required to ensure that the 1 radiological consequences of a postulated fuel handling accident inside containment are within acceptable limits. i APPLICABILITY LCO 3.9.7 is applicable during CORE ALTERATIONS. exce3t during latching and unlatching of control rod drive slafts, and when moving irradiated fuel assemblies within containment. The LC0 ensures.a sufficient level of water is present in the refueling cavity to minimize the radiological consequences of a fuel handling accident in containment. If irradiated fuel assemblies are not present in containment, there can be no significant radioactivity release as a result of a postulated fuel handling accident. Requirements for fuel handling accidents in the spent fuel pool are covered by LCO 3.7.14. " Spent Fuel Pool Water Level." 1 ACTIONS A.1 and A.2 !
/,] \ With a water level of < 23 ft above the top of the reactor vessel flange, all operations involving CORE ALTERATIONS or movement of irradiated fuel assemblies within the containment shall be suspended immediately to ensure that a fuel handling accident cannot occur.
The suspension of CORE ALTERATIONS and fuel movement shall not preclude completion of movement of a component to a safe position. i I i l (O v' BYRON - UNITS 1 & 2 B 3.9.7 - 2 7/16/98 Revision A
l Refueling Cavity Water Level B 3.9.7 ) i
/ BASES i
SURVEILLANCE SR 3.9.7.1 REQUIREMENTS I l Verification of a minimum water level of 23 ft above the top of the reactor vessel flange ensures that the design basis for the analysis of the postulated fuel handling accident during refueling operations is met. Water at the required level above the top of the reactor vessel flange limits the consequences of damaged fuel rods that are postulated to result from a fuel handling accident inside containment (Ref. 2). The Frequency of 24 hours is based on engineering judgment ; and is considered adequate in view of the large volume of i water and the normal procedural controls of valve positions. which make significant unplanned level changes unlikely ; REFERENCES 1. Regulatory Guide 1.25. March 23. 1972. I
- 2. UFSAR, Section 15.7.4.
- 3. NUREG-0800. Section 15.7.4.
I
- 4. 10 CFR 100.10. l
- 5. Malinowski. D. D., Bell. M. J., Duhn. E. and Locante. J. , WCAP-7828. Radiological Consequences of a Fuel Handling Accident. December 1971.
I
/% 'd BYRON - UNITS 1 & 2 B 3.9.7 - 3 7/16/98 Revision A
I Boron Concentration-3.9.1 r( :3:94 REFUELING OPERATIONS ~ N^ i- , r . . .
-3.9;1 .: Boro. n Concentration-a 4
- LC043.9.1; : Boron concentrations of the Reactor Coolant System, the
. refueling canal, and the refueling cavity shall be j
, . maintained within the limit specified.in the COLRJ j q
- )a
?-APPLICABILITY.:' MODE:6.
l 1 iACTIONS-. ' NOTE . 1'
.While this LCOLis not-met:, entry into MODE:6 from MODE 5.is~not permitted.
c y
- CONDITION' 1 REQUIRED' ACTION- COMPLETION TIME ~
A 5 Boron: concentration A.1- Sus)end CORE Immediately A not within limit. . ALT ERATIONS. AND: A.2- Suspend positive -Immediately:: Li reactivity' additions. 1
]l g:
0;
. A.3- Initiate action to Immediately !
2
' restore-boron'-
concentration.to '
? .
within limit.' ' ,y . f i 1 x FBRAIDWOOD.--[ UNITS 11&i2 3.9.1 - 1 7/21/98 Revision I i
-)
, e,.
= ' _ ~ _ -:___. . _ - . _ _ _ . - . _ _ - _ . _ - __ . _ ~ . _______b
i Boron Concentration l 3.9.1 t .
/ SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR. 3.9.1.1 Verify boron concentration is within the 72 hours I limit specified in the COLR.
L l L A. l O , i l l i i I I l 1 i [Q \
\
BRAIDWOOD - UNITS 1 & 2 3.9.1 - 2 7/16/98 Revision A
I Unborated Water Source Isolation Valves l 3.9.2 l L3.9 REFUELING OPERATIONS , 3.9.2 Unborated Water Source Isolation Valves LCO -3.9.2 Each valve used to isolate unborated water sources shall be secured in the closed position. APPLICABILITY: - MODE 6.
-ACTIONS.
NOTE- d Separate. Condition entry is allowed for each unborated water source isolation ! valve.
. CONDITION- REQUIRED ACTION COMPLETION TIME A. .- . NOTE- A.1 Suspend CORE Immediately Required Action A.3 ALTERATIONS.
O - must be completed whenever Condition A AND is entered. _ A.2 Initiate actions to Immediately 1 secure valve in . One or more valves not closed position.
. secured in closed position. AND A.3 Perform SR 3.9.1.1. 4 hours l
l
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BRAIDWOOD - UNITS-I & 2 3.9. 2 - 1 7/16/98 Revision A __-___.__.-_----.__u__m.__--._--___.___.-.---__m____.-___.__-_._._-_____-__--_-.-____-_-_m _ . - - _ . -
Unborated Water Source Isolation Valves 3.9.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR '3.9.2.1 Verify each valve that isolates unborated 31 days water sources is secured in the closed position. (v3 ( l
'BRAIDWOOD - UNITS 1 & 2 3.9.2 - 2 7/16/98 Revision A '
I
]
Nuclear Instrumentation 3.9.3 3.9 REFUELING OPERATIONS 3.9.3 Nuclear Instrumentation LCO ~3.9.3'- Two source range neutron flux monitors shall be OPERABLE. APPLICABILITY: MODE 6. A_CTIONS CONDITION' REQUIRED ACTION l -- COMPLETION TIME-
'A. One source range A.1 SusSend CORE Immediately neutron flux monitor ALTERATIONS.
inoperable-. AND A.2 Suspend positive Immediately reactivity additions. 'O B. Two source range B.1 Initiate action to Immediately neutron flux monitors- restore one source
- inoperable, range neutron flux monitor to OPERABLE L status.
AND s B.2 Perform SR 3.9.1.1. Once per 12 hours l i __.r-
. U ;
BRAIDWOOD : UNITS 1 & 2' 3.9.3 - 1 7/16/98Revisiond a I
Nuclear Instrumentation
, 3.9.3 SURVEILLANCE REQUIREMENTS
-SURVEILLANCE FREQUENCY SR 3.9.3.1 Perform CHANNEL CHECK. 12 hours SR 3.9.3.2 NOTE Neutron detectors are excluded from CHANNEL CALIBRATION.
Perform CHANNEL CALIBRATION. 18 months l O . l l l
-h BRdlDWOOD-UNITS 1&2 3.9.3 - 2 7/16/98 Revision A
Containment Penetrations 3.9.4 3.9 REFUELING OPERATIONS 3.9.4 Containment Penetrations I
)6 LCO '3.9.4 The containment penetrations shall be in the following status:
H a. One door in the 3ersonnel air lock closed and the 6l equipment hatch 1 eld in place by a 4 bolts:
- b. One door in the emergency air lock closed: and
- c. Each penetration providing direct access from the containment atmosphere to the outside atmosphere either:
l 1. Closed by a manual or automatic isolation valve, blind flange, or equivalent. or l 2. Capable of being closed by an OPERABLE Containment Ventilation Isolation System. NOTE Item a. only required when the Fuel Handling Building Exhaust Filter Plenum Ventilation System is not in compliance with LC0 3.7.13. " Fuel Handling Building Exhaust Q Filter Plenum (FHB) Ventilation System." 8 I [l APPLICABILITY: During CORE ALTERATIONS, g During movement of irradiated fuel assemblies within g containment. 4 O BRAIDWOOD - UNITS 1 & 2 3.9.4 - 1 7/17/98 Revision I
Containment Penetrations l 3.9.4 j ( ('~'T ACTIONS l CONDITION REQUIRED ACTION COMPLETION TIME l A. One or more A.1 Suspend CORE Immediately containment ALTERATIONS. penetrations not in required status. AND A.2 Suspend movement of Immediately irradiated fuel assemblies within containment. l l SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY
,o SR 3.9.4.1 Verify each required containment 7 days Q penetration is in the required status. ,
SR 3.9.4.2 Verify each required containment purge 18 months valve actuates to the isolation position on an actual or simulated actuation signal. SR 3.9.4.3' Verify the is51ation time of each required In accordance containment purge valve is within limits. with the Inservice' Testing Program A N..] BRAIDWOOD - UNITS 1 & 2 3.9.4 - 2 7/16/98 Revision A
RHR' and Coolant Circulation-High Water Level 3.9.5
~
3.9- REFUELING OPERATIONS' 3.9.5 Residual Heat Removal (RHR) and Coolant Circulation-High Water Level LCO- 3.9.5' One RHR loop shall be OPERABLE and in operation. NOTE The-required RHR loop may be removed from operation for s 1 hour per 8 hour period, provided no o>erations are permitted that would cause reduction of t1e Reactor Coolant System boron concentration. APPLICABILITY: MODE- 6 with the water level = 23 ft above the top of reactor vessel flange.
~
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. RHR loop requirements A.1 Suspend operations- Immediately O not met. involving a reduction in reactor coolant baron concentration. i AND A.2 Suspend loading Immediately irradiated fuel assemblies in the core. , l AND A.3 Initiate action-to Immediately-satisfy RHR loop requirements. AND (continued) [b +
- BRAIDWOOD-- UNITS 1 & 2- 3.9.5 - 1 7/16/98 Revision A
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7 RHR and Coolant Circulation-High Water Level , 3.9.5 l ACTIONS-CONDITION- REQUIRED ACTION ~ COMPLETION TIME
'A.
(continued) A.4 'Close.all containment 4 hours penetrations l providing direct-access from i containment atmosphere ta outside atmosphere. SURVEILLANCE RE00IREMENTS
- h SURVEILLANCE' FREQUENCY 'I 4
ki SR 3.9.5.1 Erify 'one RHR loop is in operation' and 12 ho'rs u 4 circulating reactor coolant at a flow rate
^
g of a 1000.gpm.
.)
i 1 i
-I 1
l l
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BRAIDWOOD - UNITS 1 & 2 3.9.5 - 2 8/3/98 Revision I
l-P RHR and Coolant ' Circulation-Low Water Level 3.9.6
] .3.9 : REFUELING OPERATIONS.
tl' 13.9.6 Residual Heat Removal (RHR) and Coolant Circulation-Low Water Level LCO :3.9.6 Two' RHR loops shall be OPERABLE, and one RHR loop shall- be in operation.
----NOTE---
One required RHR loop may be. removed from operation and-
. considered OPERABLE:
.[ a. To support filling and draining the reactor cavity when aligned to. .or during transitioning to or from, the -
-refueling water storage. tank provided the' required RHR v
' loop is capable of being realigned to the Reactor-Coolant System (RCS): or
~ b'. To support required testing provided the.requi_ red RHR L loop is capable 'of being realigned to the.RCS.
4 4 1 APPLICABILITY: ' MODE 6 with the water level < 23 ft above the top'of reactor i p vessel flange. U. ACTIONS NOTE LWhile this LC0 is.not' met, entry into MODE 6 with the water level;< 23 ft
~
(.; abo've the top of the reactor vessel flange is:not permitted,
, :CONblTION REQUIRED ACTION ~ COMPLETION TIME.
-A. One~or more RHR loops A.1 Initiate action to Immediately
. inoperable. restore RHR loop (s) to OPERABLE status.
DB (continued) M(% . lBRAIDW000'- UNITS 1 & 2: 3.9.6 - 1 7/21/98 Revision I c, l c _ _ _ - 2 _ l
RHR and Coolant Circulation-Low Water Level 3.9.6 , -ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.2 Initiate action to Immediately establish a 23.ft of water above the top ; of reactor vessel ; flange.
~
B. No RHR loop in B.1 Suspend operations. Immediately operation. involving a reduction in reactor coolant ' boron concentration. AND B.2 Initiate action to Immediately ; restore one RHR loop i to operation. .j AND h B.3 Close all containment
. penetrations 4 hours providing direct access,from 1
I containment i atmosphere to outside l atmosphere. A V BRAIDWOOD - UNITS 1 & 2 3.9.6 - 2 7/16/98 Revision A l
RHR and. Coolant Circulation-Low Water Level 3.9.6
-h
~
. SURVEILLANCE REQUIREMENTS
- X.) -
. SURVEILLANCE- FREQUENCY k _. .
a.
.SR 3.9.6.1: Verify one RHR loop is in operation and 12 hours e . circulating reactor coolant at a flow rate
'of a 1000 gpm.
sri 3.9.6.2 -Verify correct breaker? alignment and 7 days indicated )ower availablesto the re RHR pump taat is not in operation. quired' I [f}..
.BRAIDWOOD_-: UNITS 1&2
. 3.9.6 - 3 7/21/98 Revision I j
_ _ _ _ _ _ _ _ . = _ _ _ _ _ - . - _ _ __. _ . _ . o
1- . Refueling Cavity Water Level
' ~ ~
3.9.7 3,9.~. REFUELING 0PERATIONS
, .{; s ~ .. .
- 3.9;7; Refueling Cavity' Water Level .
' : LCO- 3.9.7 . Refueling-' cavity. water level:shall be maintained 2 23 ft above the. top of reactor vessel flange.
APPLICABILITY: DuringL CORE ALTERATIONS, except during latching and
' unlatching of control rod drive shafts -
, f l' f .During movement of. irradiated fuel ' assemblies within p
.containnient.
" ACTIONS CONDITION- REQUIREDLACTION' COMPLETION-TIME A. Refueling' cavity water;. A.1- Sus)end CORE Immediately
' level not<within < ALTERATIONS.
- limit.
AND: Ih 'A.2 Suspend movement of-irradiated fuel. Inunediately
- assemblies within -
-- containment.
1 SURVEILL'ANCE. REQUIREMENTS . SURVEILLANCE- FREQUENCY SR13.9'.7.1 Verify refueling cavity water level is 24 hours
=-23 ft above.the-top of reactor vessel:
flange. i u a - ff;
, A_)
1BRhlDWOODI-' UNI'TS1&2 3.9.7 - 1 7/17/98 Revision I L O_ i = - ._
Boron Concentration B 3.9 1 D V B 3,9 REFUELING OPERATIONS B 3.9.1 Boron Concentration BASES BACKGROUND The limit on the boron concentration ensures the reactor remains subcritical.during MODE 6. Refueling boron concentration is the soluble boron concentration in the filled portions of the Reactor Coolant System (RCS) the refueling canal, and the refueling cavity that are hydraulically coupled to the reactor core during refueling. The soluble boron concentration offsets the core reactivity and is measured by chemical analysis of a representative sample of the coolant in each of:the volumes. The refueling boron concentration limit'is specified in the COLR. The specified boron concentration is controlled by plant procedures to maintain an overall core reactivity of k f ' s 0,95 during fuel handling, with control rods and fuelf assemblies assumed to be in the most adverse configuration (least negative reactivity).
,s GDC 26 of'10 CFR 50. Appendix A. requires that two' Q- independent reactivity control' systems of different design 3rinciples be )rovided (Ref. 1). .One of these systems must 3e capable of 1olding the reactor core subcritical under j cold conditions. The Chemical and Volume Control. System (CVCS) is the system capable of maintaining the reactor subcritical in cold conditions by maintaining the boron concentration-(Ref. 2).
The reactor is' brought to shutdown conditions before beginning operations to open the reactor vessel for refueling. After the RCS is cooled and depressurized. the vessel head is unbolted. and removed. The refueling cavity i is then flooded with borated water from the refueling water ' storage tank through the o)en reactor vessel by gravity feeding or by the use of t1e Residual Heat Removal (RHR) , System pumps. j l O '
- BRAIDV- N -' UNITS 1 & 2 ,
B 3.9.1.- 1 7/16/98 Revision A L u
Boron Concentration B 3~9.1 p g BASES BACKGROUND.(continued) The pumping action of the RHR System in the RCS. and the natural circulation due to thermal driving heads in the . reactor vessel and refueling cavity, ensure adequate mixing of the borated ~ water. The RHR System is in operation during refueling (see.LC0 3.9.5. " Residual. Heat Removal-(RHR) and Coolant Circulation-High Water, Level." and LCO 3.9.6.-
" Residual Heat- Removal (RHR) and Coolant Circulation-Low Water Level") to provide. forced ' circulation in the RCS and
. assist in maintaining'the boron concentration in the RCS~.
.the refueling canal _ and.the refueling cavity above.the COLR
. limit.
' APPLICABLE During refueling operations, the reactivity condition of the SAFETY ANALYSES
^ core is consistent'with the initial conditions assumed for;
'the boron dilution accident'in the accident analysis-(Ref. 3) and _is conservative ~ for MODE 6. The boron concentration limit specified in the COLR is leased on the. -
core reactivity.at the beginning of each fuel cycle (the end,
'of refueling) and includes an uncertainty allowance.
v h's Lo The required ~ boron concentration and the plant refueling
. procedures. that verify the correct fuel loading plan (including-full core mapping) ensure that the km of the di core will remain s 0.95 during the refueling operation.
t Hence, at least a 5% Ak/k margin of safety is established
]l Y
during refueling. During refueling. all filled portions of the RCS. the water I
. volume in the spent fuel pool the transfer tube, the refueling canal. the refueling cavity, and the reactor vessel form a single mass. As a result, the soluble boron concentration is relatively the same in each of these volumes.
The limiting boron dilution accident analyzed occurs in
-MODE 5 (Ref 3). A detailed discussion of this event'is provided in Bases'_B 3.1.1 " SHUTDOWN MARGIN (SDM)."
The RCS boron concentration satisfies ' Criterion 2 of - 10 CFR 50 36(c)(2)(ii). ' BRAIDWOOD - UNITS 1 & 2 B 3.9.1 - 2 7/17/98 Revision I ! =___ L _ _ _ _ = _
Boron-Concentration I 1 B 3.9.1
- BASES d>
A- - LCO The LCO requires that a minimum boron concentration be
- maintained in all filled portions of the'RCS. the . refueling 3 canal.fand the. refueling cavity that are hydraulically coupled to the reactor core, while in MODE 6. The boron concentration limit specified in the COLR ensures that a ;
core k,,r of s 0.95 is maintained during fuel handling operations. . Violation of the LCO.could lead to an inadvertent criticality during MODE 6. 1 APPLICABILITY. This LC0 is applicable in MODE 6 to ensure that the fuel in the-reactor vessel will remain subcritical. The required s0 In MODES 1 and 2 boronconcentrationensuresakdroup.95: with k Alignment Limits."
,5. " Shutdown l Bank Insertion LC03.Y.2-1.0.LC03.1.4." Rod Limits." and LC0 3.1.6.
" Control Bank Insertion Limits." ensure an. adequate amount
.~
_of negative reactivity;is available to shutdown-the reactor. In MODE ~2 with k 1.0 and MODES 3. 4. and 5. LC0 3.1.1. !
~" SHUTDOWN MARGIN,r(f <SDM)," ensures that an. adequate amo negative reactivity is available to shut down the reactor and maintain it subcritical.
ACTIONS The ACTIONS are modified by a Note stating that entry into the MODE 6 from MODE 5 is not aermitted while the LC0 is not
-met. This is an exception to C0 3.0.4 and precludes detensioning the head when the refueling boron concentration limit.specified in the COLR is not met.
a A.1. A.2. and A 3 g ' Continuation of CORE ALTERATIONS.or positive reactivity j additions (including actions to reduce boron concentration) is contingent upon maintaining the unit in compliance with j9 the LCO h
- BRAIDWOOD - UNITS 1 & 2 B 3.9.1 - 3 7/17/98 Revision I l ,
. .<
- 9'
~
Boron Concentration.
.c B 3.9.1 TN .BASESL u ~
EACTIONSL(continuedh
;If the' boron. concentration of any coolant volume in the-filled portions of the RCS. the refueling canal. or.the.-
refuel 1ng' cavity isLless than its ' limit.- an inadvertent
&. criticality _ may occur due'to an : incorrect. fuel loading. -To E minimize the' potential'of-an: inadvertent criticality 1 g ; resulting from:a. fuel loading' error, all operations M ; involving CORE ALTERATIONS and positive reactivity additions must be suspended immediately; j
- Suspension of CORE ALTERATIONS and. positive reactivity-additions shall not preclude moving a component ~to a safe a
position or normal heatup/cooldown of the coolant volume for the purpose of system temperature control. In addition to immediately suspending CORE ALTERATIONS and positive reactivity-additions. action to restore the boron 2 concentration must'be initiated immediately. There are no' safety analysis assumptions of boration flow rate and concentration that must be. satisfied. The only n ~ requirement is to restore tne boron concentration to.its 7 ' required value'as soon_as paesible. In order to raise the
. h..;
boron-concentration as'.soen as/possible. the operator should begin boration with the best source available for unit
~
conditions.
.Once actions have been initiated. they must be-continued-until the boron concentration is restored.- The restoration
- time de) ends on the' amount of boron that must be injected to reach tie required contsntration.'
.e r
D m r
.;BRAIDWOOD - UNITS 1-&;2 B 3.9.1 - 4 7/17/98 Revision I
Boron Concentration B 3.9.1 :
)
("'N BASES' I I\--] 1
. SURVEILLANCE SR 3,9.1.1 REQUIREMENTS This SR ensures that the coolant boron ~ concentration in all filled portions of the RCS. the refueling. canal, and the i refueling cavity. that are hydraulically coupled with the l reactor core, is within the COLR limits. The boron '
concentration of the coolant is determined periodically by chemical analysis. 4 A Frequency of once every 72 hours is a reasonable amount of time to verify the boron concentration of representative samples. The Frequency is based on operating experience, which has shown 72 hours to be adequate to detect slow trends in boron concentration in these volumes prior to significant reduction. REFERENCES 1. 10 CFR 50. Appendix A. GDC 26.
- 2. UFSAR. Section 9.3.4.
.- 3. UFSAR. Section 15.4.6. .v l
l l l (nJ
-BRAIDWOOD - UNITS 1 & 2 B 3.9.1 - 5 7/16/98 Revision A
_ - . _ ________.-______._m__..________-__d
.x s , 1r M.- Unborated Water Source Isolation' Valves B 3.9.'2-f -.B 3.9c REFUELING OPERATIONS-1)(! '
. 1
'B 3.9.2.':Unborated Water-Source Isolation Valves
.g
# l BASES- ,
s 5ACKGROUNDL (During-MODE 6 operations, all': isolation valves for ' reactor.
- makeup water sources'.containing.unborated water that are i
? connected:to the Reactor Coolant System-(RCS)'must be closed.
to prevent unplanned boron dilution of:the reactor coolant.
,The isolation valves (CV1118. CV8428. CV8441. CV8435, and
-CV8439).must be secured in 1.he closed. position.
The'. Chemical and Volume Control System is capable.of supplying borated and unborated-water to the RCS throughL b : various flow paths. Since a positive reactivity addition made by reducing the-boron concentrationzis inappropriate
'during MODE 6: 1 solation of all' unborated water sources-prevents an unplanned' boron dilution.
I The. Refueling. Water Storage Tank (RWST)Lis assumedLto.be a-
'boration source. With the:RWST boron concentration not satisfying these assumptions, the RWST becomes a potential <
-- : dilution source and valves CV1120 and CV112E are considered 7Y unborated water source isolation valves. These valves must 1
. [l _ ,
-be~ secured in the closed position. l m
? APPLICABLE ~ The possibility of-an uncontrolled boron dilution event t SAFETY ANALYSES' '(Ref. 1): occurring during MODE o refueling. operations.is-
~
J
' : preclude.d by adherence to this LCO..which requires that
. potential dilution sources be' isolated. . Closing the
- required valves during re' fueling operations prevents the
. flow lof unborated water to the filled portion.of the RCS.
N The valves are used.to isolate unborated water sources.. These' valves have'the potential to indirectly allow dilution of the RCS' boron concentration.in MODE 6.. By isolating unborated water.' sources. a safety analysis for an . o uncontrolled-boron dilution accident in accordance with the ! Standard Review Plan (Ref. 2) is not required for MODE-6. 1 j 1 - The RCS unborated water source. isolation valves satisfy
, Criterion 2 of 10 CFR 50.36(c)(2.)(ii).
a h) e4 . {
!BRAIDWOODl-; UNITS 1 & 2. -
B ' 3.9 2 - 1 7/17/98 Revision I l v m
. . , , I-y ,
Unborated Water Source Isolat' ion Valves
,. B 3.9.2
' BASES-M LC0- This LC0 requires that flow paths to the'RCS from unborated
?- water sources'be' isolated to prevent unplanned boron M
' dilution during MODE.6 and thus avoid a-reduction in SDM.
te APPLICABILITY' In' MODE'6. this'LC0 is applicable to prevent an inadvertent
; boron dilution' event by ensuring isolation of all sources of unborated water to the RCS.
For all'other MODES, the boron dilution accident was analyzed and was found to be capable of being mitigated. LACTIONS The ACTIONS table has been modified by a Note that allows separate Condition entry for each unborated water source isolation valve.- A'.1'.'A.'2. and A.3 Continuation of CORE ALTERATIONS is contingent upon y maintaining the unit in compliance with this LCO. With'any (-- valve used to isolate unborated water sources not secured in the' closed position, all operations involving CORE ALTERATIONS must be suspended immediately. The Completion Time of "immediately" for performance of Required Action A.1
'shall not preclude completion of movement of a component to -
La safe-position. Preventing inadvertent dilution of-the reactor' coolant boron
. concentration is dependent.on maintaining the unborated water isolation valves secured closed. Securing the valves in the closed position ensures that the valves cannot be inadvertently opened. The Completion Time of "immediately" requires an operator to initiate actions to close an open valve and secure the isolation valve in the closed position without delay. Once actions are initiated, they must be continued until the valves are secured in the closed position, p
-BRAIDWOOD - UNITS.1 & 2 B 3.9.2 - 2 7/17/98 Revision I l
u . .
r___--_--- - . _. . _ _ _ . . Unborated Water Source Isolation Valves , B 3.9.2 ! BASES (] v ACTIONS (continued) Due to the potential of having diluted the boron concentration of the reactor coolant. SR 3.9.1.1 (verification of boron concentration) must be performed whenever Condition A is entered to demonstrate that the required boron concentration exists. The Completion Time of 4 hours is sufficient to obtain and analyze a reactor coolant sample for boron concentration. Condition A has been modified by a Note to recuire that Required Action A.3 be completed whenever Concition A is entered. SURVEILLANCE SR 3.9.2.1 REQUIREMENTS These valves are to be secured closed to isolate possible I m dilution paths. The likelihood of a significant reduction '
? in the boron concentration during MODE 6 operations is N remote due to the large mass of borated water in the e refueling cavity and the fact that all unborated water sources are isolated precluding a dilution. The boron H concentration is checked every 72 hours during MODE 6 under ~[] 2 SR 3.9.1.1. This SR demonstrates that valves CV1118.
CVP428, CV8441. CV8435, and CV8439 are secured closed by the u>e of mechanical stops, removal of air. or removal of electrical power. Verification of the secured valve position through a' system walkdown ensures the isolation of possible dilution paths. The 31 day Frequency is based on engineering judgment and is considered reasonable in view of I other administrative controls that will ensure that the ! valve opening is an unlikely possibility. REFERENCES 1. UFSAR. Section 15.4.6.
- 2. NUREG-0800. Section 15.4.6. I I
O l V !
- BRAIDWOOD - UNITS 1 & 2 B ' 3.9. 2 - 3 7/17/98 Revision I
" Nuclear Instrumentation B 3.9.3 (). B'3.9 ~ REFUELING OPERATIONS
'B 3.9.3L Nuclear Instrumentation
-BASES BACKGROUND The source range neutron flux' monitors are used during refueling operations to' monitor the core reactivity condition. The installed source range neutron flux monitors are part of the Nuclear Instrumentation System (NIS). These detectors are located external to the reactor vessel and detect neutrons leaking from the core. The use of
)ortable detectors is permitted, provided the C0 requirements are met.
The installed source range neutron flux monitors are boron [ trifluoride detectors operating in the proportional region of.the gas filled detector characteristic curve. The detectors monitor the neutron flux in counts per second. The instrument range covers six decades (1E+6 cpsi with a 7% instrument accuracy (Ref...'1). The detectors also provide continuous visual indication in the control room to alert operators to a possible dilution accident. The NIS is o- ' designed in accordance with the criteria presented in V Reference'2. If used, portable detectors must be functionally equivalent to the installed NIS source range monitors. APPLICABLE Two OPERABLE source range ~ neutron flux monitors are required SAFETY ANALYSES to provide a signal to alert the operator to unexpected changes in core reactivity such as with a boron dilution accident (Ref 3) or an improperly loaded fuel assembly. The need for a safety analysis for an uncontrolled boron dilution accident is eliminated by isolating all unborated water sources as required by LCO 3.9.2. "Unborated Water Source Isolation Valves." The source range neutron flux monitors satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii). I-
,Y '
F . . C' *
.BRAIDWOOD - UNITS 1 & 2- B 3.9.3 - 1 7/17/98 Revision I w.i-.- - - . . . _ -
Nuclear Instrumentation B 3.9.3 D BASES U LC0 This LCO requires that two source range neutron flux monitors be OPERABLE to ensure that redundant monitoring capability is available to detect changes in core reactivity. To be OPERABLE. each monitor must provide visual indication. APPLICABILITY In MODE 6. the source range neutron flux monitors must be OPERABLE to determine changes in core reactivity. There are no other direct means available to check core reactivity levels. In MODE 2 below the intermediate range neutron flux interlock setpoint (P-6), and in MODES 3. 4. and 5 with the Rod Control System capable of rod withdrawal or with all i rods not fully inserted, the installed source range neutron flux monitors are required to be OPERABLE by LCO 3.3.1.
" Reactor Trip System (RTS) Instrumentation."
ACTIONS A.1 and A.2 With only one source range neutron flux monitor OPERABLE. O redundancy has been lost. Since these instruments are the 'V only direct means of monitoring core reactivity conditions. CORE ALTERATIONS and positive reactivity additions must be suspended immediately. Performance of Required Action A.1 or A.2- shall not preclude completion of movement of a component to a safe ~ position or normal heatup/cooldown of the coolant volume for the purpose of system temperature ! control ' B.1 and B.2 With no source range neutron flux monitor OPERABLE. there are no direct means of detecting changes in core reactivity. Therefore, action to restore a monitor to OPERABLE status shall be initiated immediately and continued until a source range neutron flux monitor is restored to OPERABLE status. l l BRAIDWOOD - UNITS 1 & 2 B 3.9.3 - 2 7/16/98 Revision A
Nuclear Instrumentation B 3.9.3
/T V
BASES ACTIONS (continued) Since CORE ALTER TIONS and positive reactivity additions are not to be made, the core reactivity condition is stabilized until the source range neutron flux monitors are OPERABLE. This stabilized condition is determined by' performing SR 3.9.1.1 to ensure that the required boron concentration exists. The Completion Time of once per 12 hours is sufficient to obtain and analyze a reactor coolant sample for boron concentration and ensures that unplanned changes in boron concentration would be identified. The 12 hour Frequency is reasonable, considering the low probability of a change in core reactivity during this time period. SURVEILLANCE SR 3.9,3.1 REQUIREMENTS SR 3.9.3.1 is the performance of a CHANNEL CHECK which is a comparison of the parameter indicated on one channel to a similar parameter on other channels. .It is based on the assumption that the two indication channels should be (v,) consistent with core conditions. Changes in fuel loading and core geometry can result in significant differences between source range channels. but each channel should be consistent with its local conditions. The Frequency of 12 hours is consistent with the CHANNEL CHECK Frequency specified similarly for the same instruments in LC0 3.3.1. SR 3.9.3.2 g SR 3.9.3.2 is the performance of a CHANNEL CALIBRATION every o 18 months. This SR is modified by a Note stating that , e neutron detectors are excluded from the CHANNEL CALIBRATION. l 9 The CHANNEL CALIBRATION for the source range neutron flux R monitors consists of obtaining the detector discriminator M curves, evaluating those curves, and comparing the curves to
~y the manufacturer's data. The 18 month Frequency is based on q the need to perform this Surveillance under the conditions y that ap)1y during a plant outage. Operating ' experience has shown t7ese components usually pass the Surveillance when performed at the 18 month Frequency.
C BRAIDWOOD - UNITS 1 & 2 B 3.9.3 - 3 7/17/98 Rdvision I
Nuclear Instrumentation B 3.9.3 BASES REFERENCES 1. UFSAR. Table 7.5-2.
- 2. 10 CFR 50. Appendix A. GDC 13. GDC 26. GDC 28. and GDC 29.
3; UFSAR. Section 15.4.6. I l l I i I i l O BRAIDWOOD - UNITS 1 & 2 B 3.9.3 -4 7/16/98 Revision A
Containment' Penetrations B 3.9.4 B 3.9 ~ REFUELING OPERATIONS l B 3.9.4 Containment Penetrations BASES-BACKGROUND During CORE ALTERATIONS or movement of irradiated fuel assemblies' within containment, a release of fission product radioactivity within containment will be restricted from escaping to the environment when the LC0 requirements are met. 'In MODES 1. 2. 3. and 4. this is accomplished by maintaining containment-0PERABLE as described in LCO 3.6.1.
" Containment." In MODES 5 and 6, the potential for containment pressurization as a result of an accident is not-likely:.therefore, requirements to isolate the containment from the'outside atmosphere can be less striagent. -The .
LCO requirements are referred to as " containment closure" rather than " containment OPERABILITY." Containment closure means that all potential escape paths are filtered, closed, or capable of being closed. Since there is no significant
~ potential for containment 3 pressurization, the 10 CFR 50. Appendix' J. leacage criteria and tests are not required.
'( ) The containment serves to contain fission product radioactivity that may be released from the reactor core' following an accident, such that offsite radiation exposures are. maintained well within the requirements of 10 CFR 100. In addition, the containment.provides radiation shielding ! from the fission' products that may be present in the ! containment ~ atmosphere following accident conditions. ! l I i I i 7,7- 3..
-BRAIDWOOD'- UNITS 1 & 2 B 3.9.4 - 1 7/16/98 Revision A e
Containment Penetrations i B 3.9.4 I (Q BASES V BACKGROUND (continued) The containment equipment hatch, which is part of the containment pressure boundary, provides a means for moving large equi 3 ment and components into and out of containment. During CORE ALTERATIONS or movement of irradiated fuel assemblies within containment with the equipment hatch . installed, the equipment hatch must be held in place by at least four bolts. Good engineering practice dictates that I the bolts be approximately equally spated. During CORE l ALTERATIONS or movements of irradiated fuel assemblies within containment and the equipment hatch not intact, the OPERABILITY requirements of the Fuel Handling Building Exhaust Filter Plenum (FHB) Ventilation System must be met. The OPERABILITY requirements of the FHB Ventilation System are provided in LCO 3.7.13. " Fuel Handling Building Exhaust Filter Plenum (FHB) Ventilation System." l The containment air locks, which are also part of the ! containment pressure boundary, provide a means for personnel ' access during MODES 1. 2. 3. and 4 in accordance with LCO 3.6.2. " Containment Air Locks." The two air locks are the personnel air lock and the emergency air lock. Each air lock has a door at both ends. The doors are normally b -h interlocked to prevent simultaneous-opening when containment OPERABILITY is required. During periods of unit shutdown I when containment closure is not required, the door interlock mechanism may be disabled, allowing both doors of an air lock to remain open for extended periods when frequent containment entry is necessary. During CORE ALTERATIONS or movement of irradiated fuel assemblies within containment. ! ' containment closure is required: therefore, the door ' interlock mechanism may remain disabled, but one air lock door must always remain closed. An exception however, is 3rovided for the personnel air lock. It is acceptable to lave both doors of the personnel air lock opened simultaneously provided the FHB Ventilation System is in l compliance with LC0 3.7.13. 1 The closure restrictions are sufficient to restrict ! unfiltered fission product radioactivity releases from containment to the environment due to a fuel handling accident during refueling. l A > U i BRAIDWOOD - UNITS 1 & 2 B 3.9.4 - 2 7/16/98 Revision A 1
Containment Penetrations B 3.9.4 BASES BACKGROUND (continued) The Containment Ventilation Isolation System consists of the normal purge subsystem, the mini purge subsystem. and the ) post Loss Of Coolant Accident purge subsystem. These three { subsystems contain penetrations which provide direct access from the containment to the outside atmosphere. In MODE 6. the minipurge subsystem is normally used to exchange large volumes of containment air to support refueling operations. Each penetration contains inside and outside containment isolation valves which close automatically on an actuation signal. During CORE ALTERATIONS or movement of irradiated fuel within containment, all required valves within a subsystem must be capable of being closed by a containment ventilation isolation signal whenever the associated subsystem is in operation. A list of the instrumentation which functions to isolate the valves in these penetrations is provided in LC0 3.3.6, " Containment Ventilation Isolation Instrumentation." i The other containment penetrations that provide d'irect i access from containment atmosphere to outside atmosphere must be isolated on at least one side. Isolation may be g achieved by a closed automatic isolation valve a manual isolation valve. blind flange, or equivalent. E uivalent isolation methods allowed under the provisions o 10 CFR 50.59 may include use of a material that can provide a temporary atmospheric pressure ventilation barrier during CORE' ALTERATIONS or movement of irradiated fuel within the containment. I 1 BRAIDWOOD - UNITS 1 & 2 B 3.9.4 - 3 7/16/98 Revision A
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Containment Penetrations B 3.9.4 l i 4 j BASES APPLICABLE During CORE ALTERATIONS or movement of irradiated fuel l SAFETY ANALYSES assemblies within containment, the most severe radiological l consequences result from a fuel handling accident. The fuel handling accident is a postu;ated event that involves damage I to irradiated fuel (Ref.1). Fuel handling accidents. l analyzed in Reference 2, include dropping a single irradiated fuel assembly and handling tool or a heavy object I onto other irradiated fuel assemblies. The requirements of LC0 3.9.7, " Refueling Cavity Water Level." and the minimum decay time of 100 hours prior to CORE ALTERATIONS ensure that the release of fission product radioactivity, subsequent to a fuel handling accident in containment. results in doses that are well within the guideline values specified in 10 CFR 100. Reference 2 defines "well within" 10 CFR 100 to be 25% or less of the 10 CFR 100 values. The acceptance limits for offsite radiation exposure for the fuel handling accident will be 25% of 10 CFR 100 values or i the NRC staff approved licensing basis (e.g. , a specifie'd fraction of 10 CFR 100 limits). 1 Containment penetrations satisfy Criterion 3 of 10 CFR 50.36(c)(2)(ii).
\
LCO This LCO limits the consequences of a fuel handling accident in containment by limiting the potential escape paths for fission product radioactivity released within containment. - The LCO requires any penetration providing direct access from the containment atmosphere to the outside atmosphere to be closed except for the OPERABLE containment purge (supply and exhaust) penetrations. For the OPERABLE containment purge penetrations, this LCO ensures that unisolated l penetrations are isolable by the Containment Ventilation Isolation System. The OPERABILITY requirements for this l LC0 ensure the automatic purge valve closure times specified l in the UFSAR can be achieved and, therefore, meet the assumptions used in the safety analysis to ensure that releases through the valves are terminated, such that radiological doses are within the acceptance limit. The LCO is modified by a Note which allows both personnel air lock doors to be open or the equipment hatch not intact l when the FHB Ventilation System is in compliance with LCO 3.7.13. When the equipment hatch is installed it serves to contain fission product radioactivity that may be l released following a fuel handling accident in the l-BRAIDWOOD - UNITS 1 & 2 B 3.9.4 - 4 7/16/98 Revision A 1 l l
Containment Penetrations B 3.9.4 4 f) v BASES LCO (continued) containment. When the equipment hatch is not intact, or when both doors of the personnel air . lock are simultaneously opened, the internal containment pressure is essentially equal to the internal pressure of the fuel handling building. In the event of a fuel handling accident in the containment, realigning of the fuel handling building . ventilation system creates a negative pressure in the containment and fuel handling building relative to the auxiliary building and outside atmosphere. The negative pressure ensures that any radioactivity released to the containment atmosphere will either remain in the containment or be filtered through a FHB Ventilation System train. As such, with the equipment hatch not intact, or with both l Jersonnel air lock doors open, the consequences of a fuel landling accident in containment would not exceed those calculated for a fuel handling accident in the fuel handling building. APPLICABILITY The containment penetration requirements are applicable
. during CORE ALTERATIONS or movement of irradiated fuel
(~S assemblies within containment because this is when there is a potential for a fuel handling accident. In MODES 1, 2. 3. and 4, containment penetration requirements are addressed by LCO 3.6.1. In MODE 5 and in MODE 6 when CORE ALTERATIONS or movement of irradiated fuel assemblies within containment are not being conducted, the potential for a fuel handling accident does not exist. Therefore, under these conditions no requirements are placed on containment penetration status. l 1 l m V BRdIDWOOD-UNITS 1&2 B 3.9.4 - 5 7/16/98 Revision A t i l
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Containment Penetrations B 3.9.4 O v BASES 1 1 ACTIONS All and A.2 If the containment equipment hatch, air lock doors, or any containment penetration that provides direct access from the l containment atmosphere to the outside atmosphere is not in i the required status, the unit must be placed in a condition , where containment closure is not needed. This is I accomplished by immediately suspending CORE ALTERATIONS and movement of irradiated fuel assemblies within containment. Performance of these actions shall not preclude completion of movement of a component to a safe position. SURVEILLANCE SR 3.9.4.1 R.'.0VIREMENTS This Surveillance demonstrates that each of the containment penetrations required to be isolated is isolated. This l Surveillance for the open purge valves demonstrates that the valves are not blocked from closing. Also the Surveillance , will demonstrate that each valve operator has motive power which will ensure that each valve is capable of being closed by an OPERABLE automatic Containment Ventilation Isolation signal. j The Surveillance is performed every 7 days during CORE ALTERATIONS or movement of irradiated fuel assemblies within containment. The Surveillance interval is selected to be commensurate with the normal duration of time to complete fuel handling operations. As such, this-Surveillance ensures that a postulated fuel handling accident that
. releases fission product radioactivity within the '
, containment will not result in a release of fission product radioactivity to the environment.
G k) BRAIDWOOD - UNITS 1 & 2 B 3.9.4 - 6 7/16/98 Revision A l
z i Containment Penetrations ! B 3.9.4 ) p BASES. LJ . i
' SURVEILLANCE REQUIREMENTS (continued) '
SR 3.9.4.2 This Survei.llance demonstrates that each required containment purge valve actuates to its isolation position on an actual or simulated high radiation signal. The 18 month Frequency maintains consistency with other similar Engineered Safety Feature Actuation System instrumentation
' r\J ' and valve testing requirements. In LCO 3.3.6. the Q' Containment Ventilation Isolation instrumentation requires a I
-i . CHANNEL CHECK every 12 hours and a COT every 92 days to w ensure the channel OPERABILITY during refueling operations.
. x Every 18 months a CHANNEL CALIBRATION is performed.
SR 3.9.4.3 demonstrates that the isolati'on time of each
' D- valve is in accordance with the Inservice Testing Program H requirements. .These Surveillance performed during MODE 6-T will ensure that-the valves are capable of closing after a OC postulated fuel handling accident to limit a release of fission product radioactivity from the containment.
SR 3.9.4.3 This Surveillance demonstrates that the isolation time'of-
' l T: each required containment purge valve providing direct
-access from-the containment atmosphere to the outside atmosphere is in accordance with the Inservice Testing.
-Program requirements. This SR, along with SR 3,9.4.2, ensures the containment purge valves in penetrations which provide direct access from the containment atmosphere to the outside atmosphere are capable of closing after a postulated fuel handling accident to limit the release of fission Tproduct radioactivity from the containment, i
REFERENCES 1. UFSAR. Section 15.7.4. l 2. NUREG-0800. Section 15.7.4, Rev. 1, July 1981. 1 r. p j BRAIDWOOD - UNITS 1 & 2 B 3.9.4 - 7 7/17/98 Revision I i
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RHR and _ Coolant _ Circulation-High Water Level B 3.9.5 B 3 9 . REFUELING'0PERATIONS
,B.3.9.5 Residual Heat Removal;(RHR) and Coolant Circulation-High Water
. Level BASES-BACKGROUND- ~ The purpose of the RHR System in MODE 6 is to remove decay heat-and sensible heat from the Reactor Coolant System (RCS). as required by GDC 34. to provide mixing of borated coolant and to prevent boron stratification (Ref.1). Heat is removed from the RCS by circulating. reactor coolant through the RHR heat exchanger (s) where the heat-is transferred to the Component Cooling Water System. The coolant is'then returned to the RCS'via the RCS cold leg (s).
Operation of the RHR System for normal cooldown or decay heat removal is manually accomplished from the control room. The heat removal rate is-adjusted by-controlling the flow of reactor coolant through the RHR heat exchanger (s) and bypass-line(s). Mixing of the reactor coolant is maintained by this continuous circulation of reactor coolant through the
.RHR System.
APPLICABLE- While there is no explicit analysis assumption for the decay SAFETY ANALYSIS heat removal function of the RHR System in MODE 6. if the reactor coolant temperature is not maintained below 200 F. boiling of the reactor coolant could result. This could-lead to a loss of coolant in the reactor vessel. In addition, boiling of the reactor coolant.could lead to a reduction'in boron concentration in the coolant due to boron plating.out on components near the areas of the boiling activity. The loss of reactor coolant and the reduction of boron concentration in the reactor coolant would eventually challenge-the integrity of the fuel cladding, which is a fission product barrier. One train of the RHR System is required to be OPERABLE and-in operation in MODE 6, with the water level a 23 ft above the top of the reactor vessel-flange. to prevent this challenge The LC0 does permit de-energizing the RHR pump for short durations, under the condition that the boron concentration.is not reduced. This
' conditional de-energizing of the RHR pump does not result in a challenge to the fission product barrier.
f RHR and Coolant Circulation-High Water Level satisfies Criterion 4 of 10 CFR 50.36(c)(2)(ii). BRAIDWOOD UNITS 1 & 2 B 3.9.5 - 1 7/16/98 Revision A
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L RHR and Coolant Circulation-High Water Level B 3.9.5 O BASES
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LCO Only one RHR loop is required for decay heat removal in MODE 6, with the water level a 23 ft above the top of the reactor vessel flange because the volume of water above the reactor vessel flange provides backup decay heat removal capability., One RHR loop is required to be in operation and j OPERABLE to provide:
- a. Removal of decay heat:
i b. Mixing of borated coolant to minimize the possibility y of criticality; and D c. Indication of reactor coolant temperature. An OPERABLE RHR loop includes an RHR pump, a heat exchanger. valves, piping. instruments, and controls to ensure an OPERABLE flow path. The flow path starts in one of the RCS hot legs and is returned to the RCS cold legs.. The LC0 is modified by a Note that allows the required operating RHR loop to be removed from service for up to 1 hour per 8 hour period, provided no operations are n permitted that would cause a reduction of the RCS boron (") concentration. Boron concentration. reduction is prohibited because uniform concentration distribution cannot be ensured without forced circulation. This permits operations such as core ma] ping or alterations in the vicinity of the reactor vessel lot leg nozzles and RCS to RHR isolation valve testing. During this I hour period, decay heat is removed by natural convection to the large mass of water in the refueling cavity. { i l t i i (l v-BRAIDWOOD - UNITS 1 & 2 B 3.9.5 - 2 7/17/98 Revision I
l RriR and Coolant Circulation-High Water Level B 3.9.5
' BASES l
APPLICABILITY One RHR loop must be OPERABLE and in operation in MODE 6. with the water level a 23 ft above the top of the reactor i ' vessel flange, to provide decay heat removal and mixing of the borated coolant. The 23 ft water level was selected because it corresponds to the 23 ft requirement established for fuel movement in LCO 3.9.7. " Refueling Cavity Water
- Level." Requirements for the RHR System in MODES 1. 2. 3.
- 4. and 5 are covered by LC0 3.4.6, "RCS Loops-MODE 4."
LC0 3.4.7. "RCS Loops -MODE 5. Loops Filled." LCO 3.4.8.
"RCS Loops-MODE 5 Loo)s Not Filled." LCO 3.5.2.
"ECCS-Operating. " and C0 3.5.3. "ECCS -Shutdown. " RHR loop requirements in MODE 6 with the water level < 23 ft are located in LCO 3.9.6, " Residual Heat Removal (RHR) and Coolant Circulation-Low Water Level."
ACTIONS A.1. A.2. A.3. and A.4 If RHR loop requirements are not met, there will be no forced circulation to provide mixing to establish uniform boron concentrations. Reduced boron concentrations can , occur by the addition of water with a lower boron ! I concentration than that contained in the RCS. Therefore.. O actions that could result in a reduction in the coolant boron concentration must be suspended immediately. With no forced circulation cooling, decay heat removal from the core occurs by natural convection to the heat sink provided by the water above the core. A minimum refueling water level of 23 ft above the reactor vessel flange provides an adequate available heat sink. Suspending any operation that would increase decay heat load, such as loading a fuel assembly, is a prudent action under this condition. Therefore, actions shall be taken immediately to suspend loading of irradiated fuel assemblies in the core. Suspension of these activities shall not preclude completion of movement of a component to a safe condition. With the unit in MODE 6 and the refueling water level a 23 ft above the top of the reactor vessel flange removal of decay heat is by ambient losses only. Therefore, corrective actions shall be initiated immediately and shall continue until the RHR loop requirements are met.
'bD l BRAIDWOOD - UNITS 1 & 2 B 3.9.5 - 3 7/16/98 Revision A
RHR and Coolant Circulation-High Water Level B 3.9.5 , ('"v} BASES ACTIONS (continued) With the RHR loop requirements not met, the potential exists for the coolant to boil and release radioactive gas to the containment atmosphere. Therefore, all containment penetrations providing direct access from the containment atmosphere to the outside atmosphere must be closed within 4 hours. Closing containment penetrations that are open to the outside atmosphere ensures dose limits are not exceeded. The Com31etion Time of 4 hours is reasonable, based on the low pro) ability of the coolant boiling in that time. SURVEILLANCE SR 3.9.5.1 REQUIREMENTS This Surveillance demonstrates that the RHR loop is in operation and circulating reactor coolant. The flow rate is determined by the flow rate necessary to provide sufficient decay heat removal capability and to provide mixing of the borated coolant to prevent thermal and boron stratification in the core. The Fre considering the flow,quency of 12 pump temperature, hourscontrol, is sufficient. and alarm
. f) . indications available to the operator in the controi room
' \- / for monitoring the RHR System. REFERENCES 1. UFSAR. Section 5.4.7. n U. . BRAIDWOOD.- UNITS 1 & 2 B 3.9.5 - 4 7/16/98 Revision A
l RHR and Coolant Circulation-Low Water Level
, B 3.9.6 /7 B 3.9 -REFUELING OPERATIONS G
B 3.9.6 Residual Heat Removal (RHR) and Coolant Circulation-Low Water Level BASES BACKGROUND The purpose of the.RHR System in MODE 6 is to remove decay heat and sensible heat from the Reactor Coolant System ; (RCS) as required bfGDC 34, to provide mixing of borated ' coolant, and to prevent $horon stratification (Ref.1). Heat , is removed from the RCS by circulating reactor coolant ' through the RHR heat exchangers where the heat is transferred to the Component Cooling Water System. The coolant is then returned to the RCS via the RCS cold leg (s). Operation of the RHR System for normal cooldown decay heat < removal is manually accomplished from the control room. The heat removal rate is adjusted by controlling the flow of i reactor coolant through the RHR heat exchanger (s) and bypass line(s). Mixing of the reactor coolant is maintained by. this continuous circulation of reactor coolant through the RHR System. ; ( r). APPLICABLE SAFETY ANALYSIS While there is no explicit analysis assumption for the decay heat removal function of the RHR System in MODE 6, if the reactor coolant temperature is not maintained below 1 200 F, boiling of the reactor coolant could result. This could lead to a loss of coolant in the reactor vessel. In . addition, boiling of the reactor coolant could lead to a reduction in boron concentration in the coolant due to the boron plating out on components ne.r the areas of the boiling activity. The loss of reactor coolant and the , reduction of boron concentration in the reactor coolant will eventually challenge the integrity of the fuel cladding, which is a fission product barrier. Two trains of the RHR System are required to be OPERABLE, and one train in operation, in order to. prevent this challenge. . RHR and Coolant Circulation-Low Water Level satisfies Criterion 4 of 10 CFR 50.36(c)(2)(ii). j .
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- t. BRAIDWOOD - UNITS 1 & 2 B 3.9.6 - 1 7/16/98 Revision A
RHR and Coolant' Circulation-Low Water Level B 3.9.6
< BASES-LC0 Both RHR loop's must'be OPERABLE in MODE 6. with the water
-level < 23 ft above the top of the reactor- vessel flange.
In= addition..one RHR loop must be'in operation in order to
].
provide;- o
.s
'a. Removal of decay heat 4 b '. Mixing of borated coolant to minimize the possibility D
of criticality; and H T c. . Indication of-reactor coolant temperature. .
'An OPERABLE RHR loop consists of an RHR pump, a' heat' exchanger. valves, piping, instruments and controls to ensure an OPERABLE flow path. The flow Jath starts'in one of. the RCS hot. legs and is returned to tle RCS cold legs.
However, the LCO-is modified by a Note that permits the-required RHR 1000 to be. removed from' operation and . r considered OPERABLE when. aligned to, or:during transitioning to'or.from; the Refueling Water Storage Tank (RWST) to support filling:or draining the refueling cavity, or to. support required testing, if. capable of-being realigned to q .the RCS. w 1 APPLICABILITY. Two RHR loops' are required to be OPERABLE. and one RHR loop
.must be in operation in MODE 6.'with the water level <'23 ft aboveL the top of:the reactor vessel' flange. -to provide decay heat removal and mixing of the borated coolant.
.,. . Requirements for the RHR-System'in-MODES 1. 2. 3. 4. and 5
.are covered by LCO 3.4.6. "RCS Loeps-MODE 4." LCO 3,4.7 "RCS Loops -' MODE 5. Loop's Filled. " LC0 3.4.8.1 "RCS
. Loops-MODE- 5. Loops Not Filled ~ " LC0 3.5.2.
"ECCS-Operating. " and LC0 3.5.3. "ECCS-Shutdown. " RHR loop requirements in MODE 6 with the water level a 23 ft are located in LC0 3.9.5 " Residual Heat Removal (RHR) and Coolant Circulation-High Water Level.'"
g . l; ;BRAIDWOOD - UNITSL 1 & 2 B 3.9. 6 - 2 7/17/98 Revision I
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. RHR and Coolant- Circulation-Low Water Level B 3.9.6 I PY . BASES.
U _ ACTIONS- . The ACTIONS are modified by a Note stating that entry into the Applicability is not permitted while the LCO is not met.
-This-is an exce) tion to LCO 3.0.4 and arecludes transition
, into MODE 6 wit 1 water level <'23 ft w111e the LCO is not met. A.1 and'A.2-With one or more RHR' loops inoperable, the RHR System may not be capable of removing decay heat and mixing the borated coolant. LTherefore. action shall be immediately initiated
- and continued until the required number of RHR loops are
- restored to OPERABLE status or until a 23 ft of water level is' established above the reactor vessel flange. When the water level is 'a 23 ft above the reactor vessel flange, the Applicability changes to that of LC0 3.9.5, and only one RHR loop is required to be OPERABLE and in operation. An immediate Completion Time is necessary for an operator to initiate corrective actions.
B.1. B.2. and B.3 If no RHR loop is in operation, there will be.no _ forced b_ circulation to provide mixing to establish uniform boron'.
. concentrations. Reduced boron concentrations can occur by - the~ addition of water with a lower boron concentration than that contained in.the RCS. Therefore.-actions that would result in a reduction in the coolant boron concentration .
must-be suspended immediately. In addition', with no. forced circulation, any decay heat removal occurs by ambient losses only. -Therefore, action
.shall be initiated immediately to restore one RHR loop to
! operation. Once: initiated, actions shall continue until one RHR loop has been restored to operation. l'
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. BRAIDWOOD -- UNITS 1-& 2 B 3.9.6 7/17/98 Revision I
.l._L____l__N_-_'_-- - ~ . . . -
RHR and Coolant Circulation-Low Water Level B 3.9.6 BASES ACTIONS (continued) With no RHR loop in operation, the potential exists for the coolant to boil and release radioactive gas to the containment atmosphere. Therefore, all containment penetrations providing direct access from the containment l atmosphere to the outside atmosphere must be closed within i 4 hours. Closing containment penetrations that are open to. l the outside atmosphere ensures that dose limits are not " exceeded. The Com)letion Time of 4 hours is reasonable, based on the low pro) ability of the coolant boiling in that time. 1 i SURVEILLANCE SR 3.9.6.1 REQUIREMENTS This Surveillance demonstrates that one RHR loop is in operation and circulating reactor coolant. The flow rate is determined by the flow rate necessary to provide sufficient decay heat removal capability and to provide mixing of the i borated coolant to prevent thermal and boron stratification 1 in the core The Fre ! considering the flow,quency of 12 pump temperature, hourscontrol, is sufficient. and alarm indications available to the operator for monitoring the RHR System in the control room. 1 SR 3.9.6.2 Verification that the required pump is OPERABLE ensures a i RHR pump can be placed in operation, if needed, to maintain i decay heat removal and borated coolant circulation. Verification is performed by verifying proper breaker l alignment and power available to the pump. The Frequency of ' 7 days is considered reasonable in view of other ; administrative controls available and has been shown to be ' acceptable by operating experience. I REFERENCES 1. UFSAR.,Section 5.4.7. ! 1 l n V BRAIDWOOD - UNITS 1 & 2 B 3.9.6 -4 7/16/98 Revision A _ - - - _ _ - _ _ - _ _ _ _ _ _ _. _ __ _ . _ - _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ._a
Refueling Cavity Water Level B 3.9.7 7 B.3.9 REFUELING OPERATIONS B 3.9.7 Refueling Cavity Water Level' BASES BACKGROUND The movement of irradiated fuel assemblies within containment or performance of CORE ALTERATIONS except . during latching and unlatching of control rod drive shafts, requires a. minimum water level of 23 ft above the top of the reactor vessel flange. This requirement ensures a sufficient level of water is maintained in the refueling
~ cavity to retain iodine fission product activity resulting from a. fuel handling accident in containment (Refs. 1 and 2). Sufficient iodine activity would be retained to
? limit offsite doses from~the accident to < 25% of 10 CFR 100 limits. as provided by the guidance of Reference 3. APPLICABLE During CORE ALTERATIONS and movement of irradiated fuel SAFETY ANALYSES assemblies, the water level in the refueling cavity is an
'M initial condition design parameter in the analysis of a fuel
,, ? .
handling accident in containment, as postulated by _/ ;F Regulatory Guide 1.25 (Ref. 1). A minimum water level of i '( 23 ft (Regulatory Position C.1.c of Ref. 1) allows a g decontamination factor of 100 (Regulatory Position C.1.9 of
< Ref.1) to be used in the accident analysis for iodine.
This relates to the assumption that 99% of the total iodine released from the pellet to cladding gap of all the dropped fuel assembly rods is retained by the refueling cavity water. The fuel pellet to cladding gap is assumed to contain 10% of the total fuel rod iodine and noble gas inventory. with the exception'of 30% for Kr-85 (Ref. 2). The fuel handling accident analysis inside containment is described in Reference 2. With a minimum water level of 23'ft and a minimum decay time of-100 hours prior to fuel handling, the analysis'and test programs demonstrate that the iodine release due to a postulated fuel handling accident is adequately captured by the water and offsite doses are maintained within allowable limits (Refs. 4 and 5) Refueling cavity water level satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii). g BRAIDWOOD - UNITS 1 & 2 B 3.9.7 - 1 7/)7/98 Revision I i L_m__. _ _ _ _ _ _ _ _ - - - - -
=
Refueling Cavity Water Level B 3.9.7 l BASES l C) L . LCO A minimum refueling cavity water level of 23 ft above the , reactor vessel flange is required to ensure that the radiological consequences of a postulated fuel handling accident inside containment are within acceptable limits. APPLICABILITY LCO 3.9.7 is applicable doring CORE ALTERATIONS exceat during latching and unlatching of control rod drive slafts, and when moving irradiated fuel assemblies within containment. The LCO ensures a sufficient level of water is present in the refueling cavity to minimize the radiological consequences of a fuel handling accident in containment. If irradiated fuel assemblies are not present in containment. there can be no significant radioactivity release as a
~
result of a postulated fuel handling accident. Requirements for fuel handling accidents in the spent fuel'3001 are covered by LCO 3.7.14. " Spent Fuel Pool Water _evel."
. ACTIONS A.1 and A.2
( ~) i With a water level of < 23 ft above the top of the reactor U vessel flange, all operations involving CORE ALTERATIONS or movement of irradiated fuel assemblies within the containment shall be suspended immediately to ensure that a fuel handling accident cannot occur. The suspension of CORE ALTERATIONS and fuel movement shall not preclude completion of movement of a component to a safe
. position.
F l t O V BRAIDWOOD - UNITS 1 & 2 B 3.9.7 - 2 7/16/98 Revision A u__.. _ _ . - . _ _
Refueling Cavity Water Level B 3.9.7 BASES SURVEILLANCE SR 3.9.7.1 REQUIREMENTS Verification of a minimum water level of 23 ft above the top of the reactor vessel flange ensures that the design basis for the analysis of the postulated fuel handling accident during refueling operations is met. Water at the required level above the top of the reactor vessel flange limits the consequences of damaged fuel rods that are postulated to result from a fuel handling accident inside containment (Ref. 2). The Frequency of 24 hours is based on engineering judgment and is considered adequate in view of the large volume of water and the normal procedural controls'of valve positions. which make significant unplanned level changes unlikely. REFERENCES 1. Regulatory Guide 1.25, March 23, 1972.
- 2. UFSAR, Section 15.7.4.
. , , 3. NUREG-0800 Section 15.7.4.
\
- 4. 10 CFR 100.10.
- 5. Malinowski. D. D., Bell, M. J., Duhn, E. and Locante J., WCAP-7828. Radiological Consequences of a Fuel Handling Accident, December 1971.
I BRAIDWOOD - UNITS 1 & 2 B 3.9.7 - 3 7/16/98 Revision A
,3 ej, g cie.0. 2 BORON CONCENTRATION tr o 3 M .2.
LIMITING CONDITION FOR OPERATION "d
- N# i'" C"d'Y
~
1.CO 3.9.1 The boron concentration r 1. r ::.-d " = , he Reactor Coolant System and the refueling canah shall be maintained f
.... ... .. . . icient 1 \-
ginurethetth:
. ;;r re:tri-tiv: f the f;11;dn; re::tivity tenditi;n: i: 1 i !w.ht & l.;i.4 .5aceLI i,i & Coal Q (;. *
,.Y.,,,of0.95orless,or} h
'L
- i. I f A bere.. cencentr;ti;n of grc;ter than or cau:1 to 2000 ppm.I b C " b;r;n cen ntre. tion ;f ;r :ter th:n ;r cau:: 10 /300 ;;r.)
--APPH CABIHTYi-MODL.
H { t Nok
- fin:lfur not en(18.
~"
(s om gs no cerrm d. s
. n ac%n % tes%e boron coneen+1cnkn \
g & +a J 4 h e., l &d+-s . J With the requirements o the above specification not satisfied, immediately suspend all operations involving CORE ALTERATIONS or positive reactivity ,, chances and initiate /;nd centinu ter;ti:n :t gre:ter than er q :1 t 20 gp.t na-oreater-than or eaual to 7000 com boron or its {of : :Clution-contain"
.cuivalerfun-. 1. m fer :rd in :en thin er ::: to 0. n r --- 12 .= E icon n;Tellen :: r55!07 !O gre !:r !r. r. er QU .1 !0 000. pp. G /:UU ;Df@ ,'
d.ichever i -th =cre restric4ive./ - A3 Ai my p SURVEILLANCE RE0VIREMENTS V deYr n$or$t f [0. R:mov4 t g-er-enboltin;; the-reacEor-vessel--head,-and- [b. Wit rawa 'of any full- ngth ontro ' rod ~ n ex 'ss o 57 eps ( proxi tely feet .from s fu y in rte posi on thi the e actor vessel n 2.q , . , w -% reheu-s cui+o Aq
==aa The boron concentration of the Reactor Coolant System and the
. refueling canales all be iM:r' rd h;. " 'r:1 =lytid at least once per 72 hours.
se .3 9. t. : Q i i 9, g,.. & li,ihncGcl s'<>-Be CotR N Valves a-tv8441.- CV84E. and cve4391shall be verified er er r- r ! r " t - r?
\closedandsecuredinpositionev=:n=cester:
1::tri: 1 ::::rlat least once per 1 days. g '
' Lco 3.9.2 A cnous NOTL CONh- A
( _T.MSE RT 3.9- l A , @
*The eact shal be aff intaifed infl0DE 6j5theney6r f4dl is fn thf reaftor f ve sel w' h th vessel hearclosve bolt 4 lesVthan4ullygenst6ned 4r wi4h p i t a he remn ed. I V Q Not applicable to Unit 1. Applicable to Unit 2 until the completion of cycle 5. -
A3
, ** Applicable to Unit 1. Not applicable to Unit 2 until after cycle 5.
BYRON - UNITS 1 & 2 3/4 9-1 AMENDMENT NO. 65 Eca r t
l7 Lco an.W ~ L c o 3.9. #
, Q ,$ REFUELING OPERATIONS; -
(jf J.'t.4"". r.e CONTAINMENT AM333mm PENETRATIONS i LIMITING CONDITION:FOR OPERATION EThe containment buil'ing d penetrations shall.be in the following status:
. Law tac.k 4 et Am )
Qcugg ' . .The . personnel Qgtg5)(shou' d)have a minimum of one door closed at any oneLtime'and the equipment hatch shall'be in place and held by a minimum of four boltsler th: c';uip :nt 5:tch r :=:d ;;r:::nt Q dco 3,M ge--{Sur;;i' ?:nt: Receirc::nt t.0.4.2.1 1_ 43.,,4 3 q 4A)
- LLo19.4.k . jr. ~ A' minimum 'of one door. in :thelee++enee+) emergency erit h:tch) i
. closed. and' wgg LICOM 4.c. rd Each penetration providing; direct access from the containment atmosphere to the outside atmosphere shallsbe either.:
for ewivelan+ L,
- 1) Closedbyanfisolat'ionvalve,blindflange,ormanualvalve, lor
% +tr d c
. -2): Capable of-being closed by an OPERABLE automatic containment purge: isolation valve.
i1.co 19'3 - APPLICABILITY: Ouring CORE; ALTERATIONS or movement of irradiated fuel within
-the: containment.
_ ACTION: Coad ALWith:the requirements of.the above specification not satisfied, immediately suspend.all: operations involving CORC ALTERATIONS or movement of' irradiated fuel'in the. containment building. 15URVEli.LANCEREQUIREMENTS i
~
- Each of'the above required containment building penetrations shall be
'f determined toLb_e eit_her in its closed / isolated condition /fF capable of being)
. (closed by an OPERABLE automatic containment purge isolation _ valveE m..,
' .-Jat l east once per 7 daysT duri no CORE
' ALTERATIONS or. movement'of irradiated ifuel in the containment building by:
L2 SR L8i.v.i M M rm,dreA Siah
' 523 cL4.1 L e Verifying.the penetrations;are in their O = :d/i:i n d : =diticr), eq or- ' [ Ac+us k to tW Mom b ert posiHm en =
tackat ers:nwt 4edachauen sTA not ( N$
> .y
+ AR .3 '14.2. p. ' Testing.thecontainmentpurgeisolationvalvesbertheapplicable 1 INi wortions of Specification 4.6.3.2. o' 'W h ee re- IB ** A ; "!
,o: N .19l4 3 & TksE,tT .3.9 48 &- u, i./ ha=m=ht-unit 2=ic-ta-iti:1critsaliteJ6 l
, BYRON - UNITS 1"&'2 3/4 9-4 gI
LCO 3.9 5
.19 REFUELING ' OPERATIONS .
pj3,5;d. .=-- RESIDUAL HEAT REMOVAL' AND COOLANT CIRCULATION - v l
!(HIGHWATERLEVELy
@~ LIMITING CONDITION FOR OPERATION Lco =-J.9.5 - _:f At. least one residual heat removal (RHR) loop shall be OPERABLE and
;in operation."
APPLICABILITY: ~ MODE 6, when the water level above the top of the reactor
. vessel. flange is greater than or equal to 23 feet.
ACTION: f 6),y mgg y ,g ,., g CWA /Immedialcly M With no RHRtloop OPERABLE and in operation,/ suspend all operations involving en =n.-.:::: " tr.: r ::: r ::::y . :: :: q or.a reduction in boron concentra-tion of the Reactor. Coolant System and immediately initiate corrective action i
~to return the required'RHR loop' to 0.PERABLE and operating ~ status as.soon as '
possible. Close all containment penetrations providing direct access from the containment-atmosphere to the outside atmosphere within 4 hours. (4 5 vE V [ -SURVEILLANCE REQUIREMENTS i.
' S R 3.9 .5.1
- r- - At'least once per 12 hours, one RHR loop shall be verified in opera on H, and ' circulating coolant:at a flowrate of areater than or equal to 1000 gpm '
'.905 t::::r:tur: 1::: thr. or :;;;l t: li O"F.) g!
e prw0ed no opetohon1 are perm'oded lhrdwouId raase. I h redAbn d & Reu*> Ce%+.%s k wakhans : l- , L Lto Ste.*The RHR loop may be removed from operation for up to I hour per 8-hour periodo [during _ m. ... ,~_...
,. , -.... . .... .... ........., in the vicinity ef the reectm _)
f' L3 x .
! BYRON - UNITS 1 & 2 3/4 9-9 AMENDMENT NO. 38 . b
. Leo 3.9.(o
.3.9. REFUELING OPERATIONS s [RuideI b+ Re-el fuebd Coded C.,.61.g, - } l
' O 3,3 (,/ LOW WATER LEVEL g LIMITING CONDITION FOR OPERATION-Lco.3%t. nx- - Two residual heat removal (RHR) loops shall be OPERABLE, and at least one-RHR loop shall be in operation. l L.C.D N* n xmu.:r 3.9-ras I
-APPLICABILITY: , MODE 6, when the water level above the top of the reactor
. vessel. flange is less than 9't faat:
^
ON. NOTE I g bEnceri E .9 -- 10C}.h GaA g. With.less than the required RHR loops-OPERABLE, immediately initiate corrective actica to return the required RHR loops to OPERABLE status, or, establish greater than or equal to 23 feet of water above the reactoA vessel flange >2: rrrr 21 per--- - - limned u M n h k w k 4n m, (anoa,*ael4 Cond 0 #. With no RHR loop in opera (tWn,/ suspend all opera ions involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required RHR loop to operation. Close all containment penetrations providing-direct access from the containment atmosphere to the outside atmosphere within 4 hours. O r V 1 SURVEILLANCE REQUIREMENTS ff m SR 3.9.(. I - 42FN:R At least once per 12 hours one RHR loop shall be verified in operation y and circulating coolant at a flowrate of areater than or equal to 1000 gpm@ j (RCS :::::r:tur: 10:: th:r er :: :1 to 140?f Mi
& .i LAio SR .3 9.L . 2. .
( nts se.r .3.9- s o A l Q l .O q V' -
)
i BYR0'N - UNITS 1 & 2 3/4 9-10 AMENDMENT N0. 38 Revr
; CTS INSERT (S)
SECTION 3.9 LC0 3.9.6-IN5 ult 3.9 10A (Ms) SURVEILLN4CE FREQUENCY ,
.SRL 3.9.6.2 . Verify correct breaker- alignment and 7 days indicated sower available to the required RHR pump t1at is not in operation.
t - t 4 INSERT 3.9-108 (L ). ii h.
--NOTE
' One required RHR loop may be removed from operation and considered OPERABLE:
- a. 'To'. support filling and'. draining the reactor. cavity when aligned to, or during transitioning to or from, the
, refueling water storage tank provided the required RHR loop is capable of being realigned to the Reactor Coolant-System (RCS): or
- b. To support required testing provided the required RHR loop is capable.of being realigned to the RCS.
LINSERT 3.9 10C: (A6)
-NOTE
- ~ While this LC0 is.not met, entry into MODE 6 with the water level < 23 ft L ,
above the top of..the reactor' vessel flange is not permitted. es
- N 8/6/98 Revision I 7_
S e.dicrn ~$.9
'\REFUELINGOPERATIONS\ gco 3,7. gaf 3 9.11 WATER LEVEL / BORON CONCENTRATION - STORAGE POOL O LIMITINo ONDITION FOR OPERATION' /
/
/
3.9.11 At le t 23 feet of water shall be. maintained over the top of irradiated fuel ssemblies. seated-in the storage racks. ThedissolvedDo/ ron concentration of he water in the storage pool shall be maintained at/ greater than or equal to 2 0 ppm. l APPLICABILITY: Wheneve irradiated fuel assemblies are in t storage pool.
)-
ACTION:
- a. With the water level r uirements of the ove specification not satisfied, suspend all vement of fuel ssemblies.and crane operations with loads in e fuel sto ge areas and restore the water leve.1 to within its limit w hin 4 urs.
- b. With the boron concentration re irements of the above specification not satisfied, suspend all mo m t of fuel assemblies and crane 4 operations with loads in th fuel orage areas and imediately take l action to restore t.e dis lved boro concentration to within its limit as soon as possib ..
- c. The provisions of 5 cification 3.0.3 are et applicable.
1 l 4-SURVEILLANCE RE TREMENTS 4.9.11 T water levei in the storage pool shall be determined to b at least its. min' um required depth 'at least once per 7 days when irradiated fu assem ies are in the fuel storage pool. 4 .ll.a Boron concentration in the storage pool shall be determined to be reater than or equal to 2000 ppm at least once per 48 hours. Adhessed 4 Seeke ?.9 Se.e. Doc -Gr s'ec +lcm 3 7 i BYRON - UNITS 1 & 2 3/4 9-13 AMENDMENT NO. 94 Red [
Lco 39. I v.4 O / i .O REFUELINC '0PERATIONS LOO #"' A 5,4.1 OM.^.D BORON CONCENTRATION A9 LIMITING CONDITION FOR OPERATION N * * '3
/ u, LCO 3.9.1 The boron concentration c' en fill:" : ^' ,1)o the Reactor Coolant System and the refueling canal"shall be maintained /,;ntf:= =c ef+tetent te
':=en th:t th u n = trt:ti= cf th: f:ll::ir.; :ntivit; :: fiti:= i:
a 0% +Le lWit 4pdfned'.a & N lax rx ._ . . , %- .. . ..
.... . -. ,. .. . .. ., .... J- tn t Q e. :: 9 n a; = = ut ati = ;f s a:ter th e er e;r-, te see0 --- a q- _
u, -Et - ,, here unentreti;n e'._ii._r.eeter thn Or :;rd t: 2300 ;- . ADPL KARILITV- MODE .
- ~kke' $E!.$. bl&%Yot%",*h (A$ th * * +o rekt O W e&*M*i Cot 4P A m . s u " *s. 3 -, 'y With the requirements of t e above specification not satisfied, immediately suspend all operations involving CORE ALTERATIONS or positive reactivity v changes and initiatefte: ::=t:n ::ntten at ; rester thee er r; 21 t: 30 ;--
ef : =letin ertet de;; ;rnt:r thr er ete 1 te ?000 --- 5: = :r it: : cir j,alenJt-t P _ , i : rd rrd te ' err m- -- - -" - ' + = a -- - -
-- - = = :r -
concenu:auen u renere: 50 ;reate- : t er eers ! g' <, whicheve- is the -- e restrictive _f ?^^ -- ")G= := , Lk3 b Us SURVEILLANCE REOUTREMENTS a<._.._..._____u....,.,.. C,,,<________.,_,m. ....2,m...__.._
....... .... .. . .. u i u . v i u.. . .. . . .... ... . . . . . . . .
determ+ned-prior-tet b 4 ;. "; ;Via;; er steltin the rent:r ::n:1 had, rf) )
, __. , ,_ _ ____ _, ,, _ i.
Cc . . u. u._ >. o . . . . ,i
. oi . . ., . . h, , ._ ._. y_ m, u .....__, __2.... ... ...
..,,, 1 L
y3 fappr-ou4mately-1-feet)-from-its--fu14y i=:rt:d puit4erHdthin the t 7:=ter Vesset _-.--.._. .. g 3 q,I,l a4 *e rehem cavrig G. '.D The boron concentration of the Reactor Coolant System and the refueling canales all bec........... .. -..sical n;1y-12 at least once per 72 hours.
- W h & w.+ ep.4,A ,% h r_cr o to 4 i S U.%2 L...I. , ,) Valves ..,....
. -m..
o m .. ...m i- -- - -- __2
-... m. . ....)shall
~ . be verified,
, closed and secured in position ey --ri-f ra! ste;;_0- by - 12! ef af- r- /
< ',electrial nn J at least once per 31 days. Lco 3. % LA Actio,i$ e tc laascr,T' 3.9-lh 4#N P A th The enttifiEC' M .21 stat =c -tn :-m e e. never ?:e: 1: nn r :cter vessel-with the vessel hnd clesure-bo1# '^w the fu14; t:=i=:d : with 42 b= h:14 - :sL 1 s.__, j ppi.6.c,_ __ ui. .._.._u..
. . . ....._u..
. . . .._u...
. . . c. W. e ,+ , .:
D a.;;1 tub 1: te Unit I rd Unit ? Startin; eith tyrle S. t
. Unit 1 - Amendment No. 56 BRAIDWOOD - UNITS 1 & 2 3/4 9 1 Unit 2 - Amendment No. 55 9_cJ 'I-
LCO J. 7. ( 3 L C D 3 9.'{
,3,9 REFUELING OPERATIONS 3,q,4 - G/4. ^ . 4) CONTAINMENTGJI LD;IG PENETRATIONS
- f')
. '. LIMITING CONDITION FOR OPERATION 4
(++4) The containment building penetration shall be in the folloding status: l CO 3.4,4.a, ev Cate \
'The personnel 4% sht\ 3 m " w have a minimum of one door closed at any one time and the equipment hatch shall be in place and held by a minimum of four boltsf:- th: quip :nt h:t:F r==:0 pur;uant tel t.6 0 5.f.4. /bte. Gur.;i1';r. ; R quir;. ;nt 4. O.4.2,/ V Lt.4 J.f, V. k, A A minimum of one door in the(eeneme4) emergency (em t natchlis closed, and c.;c #w 9 Leo J 4,V. c.. er Each penetration providing direct access from the containment LT
, atmosphere to the outside atmosphere shall be either:
Q.adete%3A4 C ) (or 6&Wivel6aq ,
- 1) Closed by an+[ isolation valve, blind flange, @ manual valve', l 2)' Capable of being closed by an OPERABLE automatic containment purge isolation valve.
LCo 'svl . APPLICABILITY: During CORE ALTERATIONS or movement of irradiated fuel within the containment. ACTION: Got4P ll With the requirements of the above specification not satisfied, immediately suspend all operations involving CORE ALTERATIONS or movement of irradiated fuel in the containment building. SURVEILLANCE REQUIREMENTS
. 4
- a. C. 4. " Each of the above required containment building penetration shall be determined to be either in its closed / isolated condition /o~r capable of being]
(closed by an OPERABLE automatic containment purge isolation valveM 6 ID L 000 t.ours g. .o. ;; tr.; start of f M T. ieast once per 7 daysLdurino CORE ALTERATIONS or movement-of irradiated fuel in the containment building by: sn 1.4,v.! _ _ g,g
. sg y,9,q, l e- Verifyina the penetrations are in their(cle: d/icchied conditicQ g
l,8 *Y 4 l twl e.stmulaMacWkn ae%+< so su naeumshpant enium on u ) W ee &ed %ttas l h H@ SR 3.4.'f. 2 b- [ Testing the containment purge isolation valvesJper the applicable i
@ac' (portionsofSpecification 4.6.3.2. 64 le *
- art P" 88 m A ki
$R. 3.4.'I. 3 , (Imer4 .u-18)
("Not-epp+teetde to the persennet-hatch p6 to initial c. iticelity un Cycic 1] ! V BRAIDWOOD - UNITS 1 & 2 3/4 9-4 As-L REh/T , L
S.T. REFUELING OPERATIONS E1,f a/4.2.0 RESIDUAL HEAT REMOVAL AND COOLANT CIRCULATION
@(HIGH WATER LEVEL]
LIMITING CONDITION FOR OPERATION L.CO 3.48 . . N .At least one residual heat removal (RHR) loop shall be OPERABLE and in operation.* APPLICABILITY: MODE 6, when the w'ater level above the top of the reactor ! vessel flange'is' greater than or equal to 23 feet. A7 ACTION: ( loadhg 'graamA Gl awe =h A he b CoHP A - S ** d d* With no-RHR3-loop OPERABLEgand in operation,Isuspend all operations involving-nn incr rs: the rrrier d:::. 50:t 10:d?or a reduction in boron concentra-tion of the Reactor Coolant System and immediately initiate corrective action to return the required RHR loop to OPERABLE and operating status as soon as possible. Close all containment penetrations providing direct access from the containment atmosphere to the outside atmosphere within 4 hours, j
)
I ,O u '
-SURVEILLANCE REQUIREMENTS SR3.A4.8 s
P,.0.0.13 At least once per 12 hours, one RHR loop shall be verified in operation H0 and circulating coolant at a flowrate of greater than or equal to 1000 gpm @ g% C'c t ;;;reter: u;; than-or-ectua i T.e au i .) 4 e$ LA: gravide.A no opordich4 ud parm'thrA f4at would mM rekuSim 4 4Ae. Ac_he New sp+cm Lorm c.w<.eea+i%. J Lc.o e flate. >The RHR loop may be removed from operati_on for up to I hour per 8-hour period v (der' n;; ... 7. . ... . . .. . ... .....,... ~ i- th: .icinity cf the rc :ter)
' V :::1 F.ct 1 ;;;.J
~
L4
- BR IDWOOD - UNITS 1 & 2 3/4 9-9
~
AMENDMENT NO. 25 kG\/ ~C
LCo 3 % 3.q REFUELING OPERATIONS-
- L'Ret La( %- Movat (Ueb oi Golut C:,rM < Hen }
3.9.6/LOWWATERLEVEL LIMITINGCONDITIOSFOROPERATION l uco s.e.<, l GCEB~ib Two residual heat removal (RHR) loops shall be OPERABLE, and at least i one wg,RHR loop shall be in operation. Metr M-tos 3 qs l APPLICABILITY: MODE 6, when the water level e the top cf the reactor R ! vessel ACTsc>M flange is e. N ot less
< thayJncer 23 fee"Y a h -to k
r4 A 10m immediately %* ate _ achiasfa I CoNp r .a'. With less han the required RHR loops OPERABLE, immediately initiate correctiv action to return the required RHR loops to OPERABLE status, o establish greater than or equal to 23 feet of water above
.the reactor vessel flanaet .. ..... . ____. .
fot@ 6 E With no RHR loop in ope r%f (rition,/suspen$wedi,+ d all operations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required RHR loop to operation. Close all containment penetrations providing direct access from the containment atmosphere to the outside 7 atmosphere within 4 hours. s SURVEILLANCE REQUIREMENTS bi
%4 W ti n
CR 3 A.6. l G.10.D At least once per 12 hours one RHR loop shall be verified in operation H and circulating coolant at a flowrate of creater than or equal to 1000 gpm @ (200 te ,erature less th;r, ;r : =:1 to M 0 F] g l oc N sa, 3.4.4. 2 ~--trasu.r 2.4-ic e 0 bq BRAIDWOOD - UNITS 1 & 2 3/4 9-10 AMENDMENT NO. 25 REV I
CTS INSERT (S) f5 SECTION 3.9
%.) .
LC0 3.9.6 INSERT 3.910A :(M 3) . .
' SURVEILLANCE FREQUENCY
'SR .3.9.6.2-
. Verify correct breaker alignment and 7 days indicated )ower available to the required RHR pump tlat is not in operation.
INSERT'3.9 10B (Lh) NOTE
/~T One required RHR loop may be removed from operation and 's J considered OPERABLE:
- a. To support-filling and draining the reactor cavity when aligned to, or during transitioning to or from, the refueling water storage tank provided the required RHR
' loop is capable of being realigned to the Reactor
-Coolant System (RCS): or
~
- b. To support required testing provided.the required RHR loop is capable of being realigned to the RCS.
-INSERT 3.910C (Au)
-NOTE .
..While this'LC0 is not met, entry into MODE 6 with the water level < 23 ft b above the top.of- the reactor vessel flange is not permitted.
L
): .
8/6/98 Revision I l l - 1
- 1 L _ - - - - - _ -
i
TREF #UELING OTERA gy 9.11 WATER LEVEL / BORON CONCENTRATION - STORAGE P00 ( c0 3.~7.14
\
llMITIN4 CONDITION FOR OPERATION L co ~3,n.1 5' O N
, l 3.9.11 At least 23 feet of water shall be maintained over the top of irradiated fuehassemblies seated in the storage racks. The dissolved oron f' concentration oNhe water in the storage pool shall be maintained a greater than or equal to 2000 ppm. 1 APPLICABILITY: Whenever irradiated fuel assemblies are in t storage pool.
I ACTION:
- a. With the water leve\ of the l requirements ove specification not satisfied, suspend all 'mqvement of fuel ssemblies and crane
- operations with loads in De fuel sto ge areas and restore the water level to within its limit W thin 4 urs.
- b. With the boron concentration r raments of the above specification not satisfled, suspend all mov t of fuel assemblies and crane operations with loads in the uti torage areas and immediately take action to restore the diss ved bor concentration to within its limit as soon as possibi . ~
c. l The provisions of Sp ification 3.0.3 ar not applicable. j
.O i
SURVEILLANCE REOU EMENTS
.)
4.9.11 The ater level in the storage pool shall be determined to b at least its minim required depth at least once per 7 days when irradiated f 1 l assembi s aie in the fuel storage pool. 4.9 1.a Bot on concentration in the storage pool shall be determined to be gr ater than or equal to 2000 ppm at least once per 48 hours.
; i Addussed k Su%%1 Sa Doc for .Sec Han 3 7 l
BRAIDWOOD - UNITS 1 & 2 3/4 9 13 AMENDMENT NO. 85 L w'._____- - - - -- _
~
DISCUSSION OF CHANGES TO CTS
, ITS SECTION 3.9 REFUELING OPERATIONS J^
ADMINISTRATIVE CHANGES (A) A3 All reformatting, renumbering, and editorial rewording are in accordance with the Westinghouse Standard Technical Specifications. NUREG-1431. During the development certain wording preferences or English language conventions were adopted. As a result, the Technical Specifications (TS) should be more readily readable, and therefore understandable, by plant operators and other users. During the reformatting, renumbering. and rewording process, no technical changes (either actual or interpretational) to the TS were made unless they were identified and justified. A2 CTS LC0 3.9.1 Footnote
- denotes requirements for defining MODE 6. CTS SR 4.9.1.1.a denotes a requirement that is adequately addressed by LCO 4.0.4 for surveillance testing prior to entering the " defined" MODE 6 (i.e.. unbolting the reactor head). NUREG-1431 does not contain references to these requirements, but rather reformats the ] presentation of these requirements. ITS Section 1.1 defines MODE 6 as " Required reactor vessel head closure bolts less than fully tensioned" and ITS SR 3.0.4 requires the LCO's Surveillance to be met within their specified Frequency prior to entry into a MODE or other specified condition in the Applicability. During this reformatting no technical m changes (either actual or interpretational) to the TS were made unless s
) they were identified and justified. This change is consistent with NUREG-1431.
A3 CTS LC0 3.9.1. Footnote #. Footnote **, and associated requirements ~ denoting boron concentrations limits for specific cycles have been deleted. The deleted boron concentration limits are for previous fuel cycles and are no longer applicable to operation of the units since the applicable cycles will have been completed. The change is editorial in nature and does not involve a technical change (either actual or interpretational) to the TS. This change is consistent with NUREG-1431. A, Not used. lA3 CTS LCO 3.9.4. rootnote
- and SR 4.9.4.2. Footnote # (Braidwood only) and associated references denoting the non-applicability of certain requirements prior to initial criticality on Cycle 1 have been deleted.
The deleted requirements are for previous cycles and are no longer applicable t.o operation of the units since the applicable cycles have been completed. The change is editorial in nature and does not involve a te'c hnical change (either actual or interpretational) to the'TS. This change is consistent with NUREG-1431.
, s, 3.9 1 l (l UNITS 1 &.2 7/17/98 Revision I
! BYRON /BRAIDWOOD
Tf
$ . DISCUSSION OF CHANGES TO CTS ITS SECTION 3.9 REFUELING OPERATIONS
( l c -- L b m. . A, nj 1CTT LC0 3.9.1 requires maintaining boron concentration of all filled portions'of the RCS and refueling canal. Clearly, it is the intent of l t-
$ CTS to include the refueling cavity when referencing the RCS in the LC0 te and Surveillance Requirements Sections. Therefore, ITS LC0 3.9.1 clarifies the " filled portions of the RCS" region by specifically denoting the " refueling cavity" area. 'This change is perceived as the
. intent of the CTS wording, is considered editorial in nature and does not involve a technical change (eitbar actual or interpretational) to the TS. This change is consistent with NUREG-1431.
Aw CTS LCO 3.9.1-provides. surveillance requirements for the unborated water source' isolation valves and associates the specific requirements-for these valves with.the requirements for maintaining boron concentration. NUREG-1431- reformats these requirements and provides . separate require'ments for the unboratec source isolation valves (ITS LC0 3.9.2). The CTS has been revised to add the specific LC0 and Action requirements. During this reformatting.no technical changes (either actual or interpretational) to the TS were made unless they were identified and' justified elsewhere. This change is consistent with l 'NUREG-1431.
.Au -CTS'LCO 3.9'.9. requires the Containment Purge Isolation System to be- -
OPERABLE 'during CORE ALTERATIONS or during movement of irradiated fuel h;% assemblies. In addition,- CTS Action 3.9.9.a requires closure of each of the aurge valves providing direct access from the containment atmosphere. to tie outside. atmosphere whenever the Containment Purge Isolation System is inoperable. NUREG-1431 reformats and combines these requirements into ITS.LC0 3.9.4 c for each )enetration providing direct:
. access from the containment atmosphere to tie outside atmosphere. The l
ITS requires the penetration to be closed by the isolation valve or
. capable of being closed by an operable Containment Purge. Isolation System. During this reformatting no technical changes (either actual or
. interpretational) to the TS were made unless they were identified and
- justi fied. This change is consistent with.NUREG-1431.
i BYRON /BRAIDWOOD UNITS 1 & 2 3.9 3 7/17/98 Revision I I l-O x_:__________. -_ _ . _
DISCUSSION OF CHANGES T0 CTS p ITS SECTION 3.9 . REFUELING OPERATIONS' (); Au CTS 3.9.1 ' Actions-are revised to include a Note precluding a MODE change
'into MODE 6 from MODE 5 while the LCO-isLnot met. In addition. CTS 3.9.8.2 Actions are revised to include a Note precluding a MODE ct.ange into. MODE 6 with the water level < 23 ft above the top of the reactor '
vessel flange while the LC0 is not met. 'These Notes are inserted in conjunction with a revision 1to LC0' 3.0.4 (refer to Section 3.0 DOC L3 ). As a' result of the~ change to LCO 3.0.4 all ITS Actions were evaluated for individual. acceptability of this change. Based on this evaluation where MODE change restrictions were determined to be required in MODES 5 and 6. or in MODES 1, 2. 3. and 4 during unit shutdown Notes containing
.the appropriate MODE change restrictions are added to the individual 1 Specifications. The Notes that are added.to ITS 3.9.1 and ITS 3.9.6 Actions are a result of this evaluation. Since the technical aspects of this change are addressed in Section 3.0 this-change is considered-administrative in this Section:
P 7-) M l q i i j i l. I
~
n js_/- BYRON /BRAIDWOOD - UNITS 1 & 2. 3.9 3a 7/17/98 Revision I
'i.
DISCUSSION OF CHANGES TO CTS
'V h ITS SECTION 3.9 REFUELING OPERATIONS LA, - CTS LCO 3.9.1 contains descriptive information regarding the regions required to maintain minimum ]oron concentration. This information is relocated to the-ITS Bases. The requirements of' ITS LC0 3.9.1. are
' adequate to ensure the minimum boron concentration is maintained within
-required limits. As a result. the relocated information is not necessary to be included in the TS to provide adequate protection of the public health and safety. The relocation of this information maintains the consistency with NUREG-1431. Any change'to this descriptive
'information will be made in accordance with the Bases Control Program described in ITS Section 5.5.
1 LA, CTS SR 4.9:1.2 contains details regarding the analysis used for determining boron concentration. LThese details of the method of performing the-Surveillance is relocated to the ITS Bases. The requirements of ITS LCO 3.9.1 and ITS SR 3.9.1.1 are adequate to ensure-the minimum boron concentration-is maintained within required limits. As a: result, the relocated details are not necessary to be included in the TS to provide adequate protection of the public health and safety. The relocation of these details maintains the consistency with NUREG-1431. Any change to these details will be made in accordance with
. the Bases Control Program described in ITS Section 5.5.
. LA, CTS LC0 3.9.2 contains details regarding the Operability requirements of -
1d the source range neutron- flux monitors.. The requirement that each monitor provide visual indication is relocated to the ITS Bases. These details of Operability are not necessary in the LCO. The definition of 0)erability and the requirements of ITS LC0 3.9.3 suffice. As such. t1ese details are not required to be in the TS to provide adequate 4 protection.of the aublic health and safety. The relocation of this , detail maintains t1e consistency with NUREG-1431. Any change to the { detail relocated to the Bases will be made in accordance with the Bases ] Control Program described in ITS Section 5.5.- LA w CTS SR 4.9.8.1 and SR 4.9.8.2 contain specific requirements for RCS
%go temperature for the performance of RHR loop testing. This requirement is relocated to the TRM. The requirements of ITS LCOs 3.9.5 and 3.9.6 Qt? are adequate to ensure the required residual heat removal loop (s) are 4
g l- maintained Operable and in operation. This requirement is not required . to be in the TS to provide adequate protection of the public health and 4h safety.. --The relocation of this requirement maintains the consistency i
.Q with NUREG-1431. Any change to this requirement will be made in accordance with 10 CFR 50.59. . /~)N E(~ BYRON /BRAIDWOOD UNITS 1 & 2 3.9-11 7/17/98 Revision I L. - . __ _ _ - - . - - - - - - - - = -
DISCUSSION OF CHANGES TO CTS ( i ITS SECTION 3.9 REFUELING OPERATIONS
.L2 CTS SR 4.9.4.1 and SR 4.9.9 require the containment purge isolation valves.to be demonstrated OPERABLE every 7 days by verifying that
-T containment isolation occurs on an actuation signal. ITS SR 3.9.4.2
- 4. requires the containment Jurge isolation valves to be demonstrated y OPERABLE every 18 months ay verifying that containment isolation occurs J' on an actuation signal. CTS:SR 4.9.4.1.b has been revised to relax the g Surveillance Frequency from 7-days to 18 months. This is acceptable .
since the instrumentation is highly reliable and since the containment k purge isolation system requires the performance of additional SRs to ensure operability of the system and the containment isolation function. The 18 month Frequency maintains consistency with other similar ESFAS instrumentation and valve testing requirements. The SRs for the actuating instrumentation are contained in LCO 3.3.6. These requirements include a 12 hour Channel Check. a 92 day Channe~.
-Operational Test, and an 18 month Channel Calibration and TADOT (for the
' Manual. Initiation function only). In addition to the instrumentation j requirements of LCO 3.3.6. each containment purge isolation valve 9 requires the verification of the isolation time in accordance with -the IST program.(SR 3.9.4.3). In addition. CTS SR 4.9.4.1.a requires j verification that the penetrations are in their " closed / isolated" conditions at.least once every 7 days. ITS SR 3.9.4.1 re containment penetrations to be in their " required status" every quires7 the days.
The ITS demonstrates that each containment penetration required to be
.C]- 94u isolated is indeed isolated and those purge valves that are open are not blocked from closing. This change is consistent with NUREG-1431.
L3 CTS SR 4.9.1.1.b requires the reactivity conditions to be determined prior to withdrawal of any full-length control rod in excess of 57 steps from its fully inserted position within the reactor vessel. ITS LCO 3.9.1 requires boron concentration to be maintained within the limit specified in- the COLR. The CTS has been revised to delete this conditional based surveillance. This change is acceptable since the intent of this surveillance is analytically encompassed within the boron-
. limit of the LCO. As denoted in the Bases for ITS LCO 3.9.1. the required boron concentration and the plant refueling procedures that
. verify the correct loading plan ensure that the k r, of the core will l remain s 0.95 during refueling operation. SincelheLCOrequiresthe '
boron concentration limit to be met prior to and during the a)plicable mode, this conditional based surveillance is not required. .Tlis change ; is consistent with NUREG-1431. j 1 I I O 8vRONe8RA1 0* UN nS 1 a 2 3.9 14 7/17/98 Rev4sion 1
DISCUSSION OF CHANGES TO CTS { ITS SECTION 3.9 . REFUELING OPERATIONS L4 Comed has conformed to the STS LC0'3.9.5' Note which allows the recuired RHR loop to be removed from operation for 1 hour per 8 hour perioc provided no operations are 3ermitted that would cause reduction of the RCS boron concentration. 'T1is Note allows other refueling operation
' activities such-as core mapping. valve testing. installation and removal
' :of temporary ' lighting.-tem)orary submarines or. any. other activity not
' involving any operations tlat would cause a boron reduction in the RCS I or a reactivity-change. The CTS also provides a footnote stating that the RHR loop may be removed from operation for up-to_1 hour per 8 hour
. period during-the performance of CORE ALTERATIONS in the vicinity of the reactor vessel hot legs, In converting to 115. and in-compliance wi.th ;
- o. the STS; the words. "during the-performance of CORE ALTERATIONS'in the ,
L -(/ vicinity of the vessel hot legs" were reinoved. The ITS definition of CORE ALTERATION states. " CORE ALTERATION shall be the movement of any T fuel, sources. 'or. reactivity control components, within the reactor
# vessel with the vessel head removed and fuel in the vessel." The ITS definition differs from the CTS definition in that equi) ment-in' the h vicinity of the hot' leg (such as lighting. temporary su) marines'. etc.)
.{ is no longer required to be classified as CORE ALTERATIONS in ITS.
Without the revised Note, certain operations which were. allowed under the CTS Note would no longer be allowed in ITS. The footnote in ITS LCO 3.9.5 is consistent with the ITS definition and the deletion of the words. "during the performance of CORE ALTERATIONS in the vicinity of
-Q ,
the vessel hot . leg" is appropriate. i 4 I
. BYRON /BRAIDWOOD UNITS 1 & 2 3.9 15 7/17/98 Revision I
DISCUSSION OF CHANGES TO CTS ITS SECTION 3.9 REFUELING OPERATIONS L3 CTS LC0 3.9.10 requires maintaining the reactor vessel water level limit during movement of control rods within the containment. ITS LCO 3.9.7 i requires maintaining the refueling cavity water level limit during CORE ALTERATIONS, except during the latching and unlatching of control rod drive shafts. The deletion of the CTS reactor vessel water level requirements for the movement of control rods is acceptable since movement of control rods (outside of a fuel assembly) within the reactor vessel during MODE 6 does not normally occur except .for unlatching and latching of the control rods. No safety analysis exists for a fuel handling accident due to a dropped control rod. In containment, control rods are moved within their associated fuel assemblies and therefore. the requirement to maintain water level will be applicable. The ITS definition of CORE ALTERATIONS specifically includes movement of control rods and is part of the Applicability of ITS LCO 3.9'.7. Therefore, a separate water level requirement for moving control rods is unnecessary. The CTS has also been revised to add the water level requirement exception since the function for latching and unlatching control rod drive shafts requires the movement of control rods. This change is acceptable since the core is fully loaded and the upper internals are installed during the evolution of latching and unlatching control rod drive shafts. In this configuration it is not Jossible to drop a fuel assembly or control rod into the core. Since t1e water level requirement is specifically associated with mitigating the release of Q fission products to the environment in the event of a dropped assembly into the core, it is not necessary to maintain the reactor vessel water level limit during latching or unlatching control rod drive shafts. This change is consistent with NUREG-1431. O BYRON /BRAIDWOOD - UNITS 1 & 2 3.9 15a 7/17/98 Revision A
1 j DISCUSSION OF CHANGES TO CTS j (aq , ITS SECTION 3.9 - REFUELING OPERATIONS ' m (lL. g CTS SR 4.9.4.1 and SR 4.9.9 requires verification that the containment purge valves actuate to the isolation sosition on an ESF test signal. q ITS SR 3.9.4.2 requires verification t1at the containment purge valves actuate to the isolation position on an actual or simulated actuation H signal. The CTS has been revised to allow either an actual or simulated 4 actuation signal for demonstrating performance of the SR. This is M acceptable since use of an actual signal instead of a simulated signal will not affect the performance of the associated components. There is no reason why an actual signal would preclude satisfactory verification or performance of an actuation logic test or channel operational test. Operability can be adequately demonstrated in either case since the associated components can not discriminate between actual or simulated signals. This is perceived as the intent of the CTS wording, and therefore, could be considered an administrative change. However. since this allowance is being specifically denoted in the SR, this change is discussed and justified as a less restrictive change. The intent of this change is to provide clarity and completeness in avoiding any misinterpretation. This change is consistent with NUREG-1431. L, CTS SR 4.9.4.1 and SR 4.9.9 require, within 100 hours prior to the start of CORE ALTERATION or movement of irradiated fuel within the containment, that the containment purge isolation valves be demonstrated operable and that the required containment. penetrations be verified (,l closed. ITS LC0 3.9.4 and SR 3.0.4 require the OPERABILITY of the containment purge isolation valves and the verification of required containment-penetrations prior to entering the mode of applicability (e.g., prior to core alterations or the movement of irradiated fuel within the containment). The CTS has been revised to delete the 100 hour requirement. This is acceptable since the ITS provides general rules for the application of all surveillance requirements in the TS. SR 3.0.4 establishes the requirement that all applicable SRs must be met before entry into a mode or other specified condition in the Applicability. In addition, the specific time frames and conditions necessary for meeting the SRs are specified in the Frequency. This change is consistent with NUREG-1431. I I ( , t V BYRON /BRAIDWOOD - UNITS 1 & 2 3.9 17 7/17/98 Revision I (
e Baron Concentration 3.9.1
- /'3 ;
~V 3.9 REFUELING OPERATIONS' 3.9.1' Boron' Concentration-.
LCO 3.9.I, Boron concentrations of the Reactor Coolant System, the refueling canal, and the refueling cavity shall .be maintained within the limit specified in the COLR. APPLICABILITY: MODE 6. 1 W....._...-s,e--._~~.-------.Nonc Hooc (, % s e w , w a Lct . > n., ,n .x ,, n m _ _ ,_ _3,,,m i
. ACTION w_ c- _Ot*_*,T.?_**.--
"f .. ___,,,,,
CONDITIONL REQUIRED ACTION C0fiPLETION TIME
'A. Boron concentration A.1 Suspend CORE Immediately not within limit.- ALTERATIONS.
AND Suspend positive . Immediately h A.2 reactivity additions. d. AND A.3. Initiate action to Immediately restore boron concentration to within limit. 4 1 J l SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY Verify boron concentration is within the 72 hours SR -3.9.1.1-limit specified in c g y . 3.9-1 Rev 1, 04/07/95 L .kOG STS , 12u) f L
Containment Penetrations 3.9.4 ] 3.9 REFUELING OPERATIONS 3.9.4 Containment Penetrations is
/on, Joor in et p.erwwe14'a. leek ele sed a+td ff" LCO 3.9.4 The containment penetrations shall be in the following ,
status:
- a. ,
k he equipment hatch C ;;; e g held in place by M 19 N 3 olts; _
- b. One door in(M) air lock closed; and
- c. Each penetration providing direct access from the containment atmosphere to the outside atmosphere either:
- 1. closed by a manual or automatic isolation valve, blind flange, or equivalent, or-
- 2. capable of being closed by an OPERABLE Containment tFiimcat smRgr11s01ation System.
(Inser+ 3 8-toAM APPLICABILITY:
& During CORE ALTERATIONS, ~
H *?
$}
During movement of irradiated fuel assemblies within Em eq containment. b E s ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.1 Suspend CORE Immediately A. -One or more containment. ALTERATIONS. penetrations not in required status. AND A.2 Suspend movement of Immediately irradiated fuel assemblies within containment. o U 3.9-6 Rev 1, 04/07/95 WOG STS Re/ r
RHR and Coolant Circulation-High Water Level 3.9.5 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. _ (e.onti nued) - A.4 Close' all containment- 4 hours penetrations
, providing direct-
-access from containment atmosphere to outside
, atmosphere. SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY
. SR '3.9.5.1- Verify one RHR loop is in operation and 12 hours circulating reactor coolant at a flow rate b( -- of 2 = ;; ogpm.
% w?
._ .. . CED .
34, r l-- i l( WOG STS 3.9-9 Rev 1, 04/07/95 REVr L I- .
RHR and Coolant Circulation-Low Water Level 3.9.6 o U 3.9 REFUELING OPERATIONS 3.9.6 Residual Heat Removal (RHR) and Coolant Circulation-Low Water Level LC0 3.9.6 Two RHR loops shall be OPERABLE, and or.e RHR loop shall be in operation. Noke- C . 'xmes:r 9A - to A APPLICABILITY: MODE 6 with the water level < 23 ft above the top of reactor vessel flange. h wa Lco a nor m 4m -...... . m t4eoc af er lever _. - h
- ACTIONS t'.R *h***J en%
.. h_e % g o U.e91 c yeGeL lig e o s nom+h th (1r%f$e __ .
CONDITION REQUIRED ACTION COMPLETION TIME A. JLess.th;n the requirec . A.1 Initiate action to Imediately au;;;t,er af """, l eep: i restore (r4eu-wed RHR
^PERABLE. / loops to OPERABLE One or more. hft,' -O
(. loop 4 ikofE CL d e..J N
. \. J Imediately A.2 Initiate action to establish 2 23 ft of water.above the top of reactor vessel flange.
B. No RHR-loop in B.1 Suspend operations Imediately operation. involving a reduction in reactor coolant boron concentration. AND (continued) A V WOG-STS 3.9-10 Rev 1, 04/07/95~
/ZW f s
RHR and Coolant Circulation-Low Water Level 3.9.6
~
ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME I B. (continued) B.2 Initiate action to Immediately j restore one RHR loop ) to operation. j AND B.3 Close all containment 4 hours penetrations providing direct l access from containment atmosphere to outside atmosphere. ex SURVEILLANCE RFOUIREMENTS b SURVEILLANCE FREQUENCY SR 3.9.6.1 Verify one RHR loop is in operation and 12 hours - circulating reactor coolant at a flo,w rate H g of 2 FNRRR gpm. gs D =% i i 7 days @ SR 3.9.6.2 Verify correct breaker alignment and e indicated power available to the required RHR.. pump that is not in opercticr.. . 1 ( y'Y U -
-WOG STS 3.9-11 Rev 1, 04/07/95 l
L ({EV I L_ _ - _ _ _ _ _ _ _ _ - _ _ __. .- A
L. ., . j' , L I . Refueling' Cavity W,ter. Level-
,c 3.9.7 )
[3.9 REFUELING OPERATIONS'
.. ?
n , _ ~3.9.7.M Refueling Cavity Water. Level' 1 a] j l T LC0 i3.9.7> ' Refueling cavity water level- shall be maintained ;t 23 ft (5 r,
.above the top of f reactor vessel flange.-
g. O' 4: ~ f' LAPPLICABILITYi During CORE ALTERATIONS, except during latchingL and. 'A 3
. .. ' unlatching of. control' rod drive shafts, _
During movement'of irradiated fuel assemblies within a
@4 C ,pf 1' containment.
- k.
y. JACTIONS: J- , CONDITIONi : REQUIRED ACTION COMPLETION TIME li ,
? A. Refueling 1 cavity water A.1: . Suspend COREL ' Immediately
'levelcuotlwithinl ALTERATIONS.
limit. .' 17 4 8E! t
' v~-
- A.2 : Suspend movement of? _ Immediately.
-irradiated fuel c assemblies within
-containment.
ll ; L .' A.3 ~ ~ Initiate ac . ^ ediately' D , ' restore r eH
, water level to.
within limit. lL f n
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J d WOG.STSI 3.9-12 Rev 1, 04/07/95
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.__._L_ _ _ .__[_... _ _ _ _ _ _ - . _ _ _ _ - - _
JUSTIFICATION FOR DIFFERENCES TO NUREG 1431 LCOS p%/ SECTION 3.9 - REFUELING OPERATIONS. BRACKETED ~ CHANGES (B)
'B r ITS LCO-3.9.3 was revised to delete theLoptional wording in the brackets since the design only contains two source range neutron flux monitors.
, : Bi - The brackets were removed and.the plant specific -value was added.
'M
- p o,g
. ( g,--GENERIC Cr CHANGES This change is consistent with NUREG-1431. (C)-
as modified by TSTF-21, y^ Revision 1. IC l 2
;Not used.
C3 - This change is consistent with-NUREG-1431, as modified by TSTF-20 (NRC g)2{ Approved). 1l Cr Not'used. Ci This change'.is consistent with NUREG-1431, as modified by TSTF-96 (NRC
' Approved) n U-4
+
-e I
( _,m
' d. )( BYRON /BRAIDWOOD UNITS 1 & 2 3.9 1 7/17/98 Revision I i ,-4. . . - . . . -
1 JUSTIFICATION FOR DIFFERENCES TO NUREG 1431 LCOS SECTION 3.9 - REFUELING OPERATIONS PLANT SPECIFIC CHANGES-(P) Pc During the development certain wording preferences. English language
. conventions. reformatting. renumbering, or editorial rewording consistent with plant specific nomenclature were adopted. As a result.
the Technical Specifications (TS) should be more readily readable by. _ and therefore understandable to. plant operators and other users. . During this reformatting, renumbering, and rewording process, no
~ technical changes-(either actual or interpretational) were made to the
.TS unless they were identified and justified.
Pr ITS LC0 3.9.4 and associated Bases was revised to allow an exception to
.the LC0 by the addition of a Note. This change is consistent with the current licensing basis as presented in CTS LCO 3.9.4 and allows. with the fuel handling building. exhaust filter system capable of maintaining a negative pressure. the equipment hatch to be not intact or both equipment hatch air lock doors to be opened. . In this condition, the containment and fuel handling building atmospheres are coupled. In the event of a fuel handling accident within the containment, any release of radioactivity to the containment may be communicated to the fuel handling building. Requiring the fuel handling building exhaust filter system to be capable _of maintaining a negative ]ressure_with respect to. 4 N
the outside atmosphere precludes a release to t1e environment beyond the i T e) V d g assumptions of the safety analysis. Consistent with the Byron and Braidwood hatch design, the hatch is only referred to as being " held in rd place" versus " closed." l p N kP 3 NUREG LC0 3.9.4 and the associa.ted Surveillance Requirements section of the Bases were revised to add a new surveillance (ITS SR 3.9.4.3). This SR requires the verification of each required containment purge valve isolation time consistent with the IST program. The isolation times L ensure that each containment purge supply and exhaust valve. in l penetrations which provide direct access from the containtnent atmosphere L' to the outside atmosphere, is capable of closing within the assumptions i L of the safety analysis following a fuel handling accident. l P, .A Note was added to NUREG SR 3.9.6.1 by TSTF-21. Revision 1. -This Note which allows the requirement for one RHR loop to be removed from , operation and consicered OPERABLE for.certain evolutions, has been ! repositioned following the LCO. This change enhances the presentation i of the requirements for the Note and eliminates the potential for misinterpretation. During this re)ositioning. the Note has been reformatted consistent with other _C0 notes. This reformatting did not involve any technical changes. 0; mRON/BRAIDWOOD UNITS 1 & 2 3.9 2 7/17/98 Revision I 4
JUSTIFICATION FOR DIFFERENCES TO NUREG 1431 LCOS SECTION 3.9 REFUELING OPERATIONS P3 Condition A of ITS LCO 3.9.6.was revised to clarify the description of the Condition and the associated Required Action. This change eliminates the potential for misinterpretation of the Action requirements and provides consistency with the style and format of other similar ITS Action requirements. This is an additional enhancement only and does not involve a technical issue. P. This Note was added for consistency and compliance with NUREG-1431. LC0 3.0.4. O O - evR0Ne8RAIow000 . uN1TS 1 a 2 3.9 3 7/17/98 Rev4s4en 1
Baron Concentration B 3.9.1 g V B 3.9 REFUELING OPERATIONS B 3.9.1 Boron Concentration
%ak- Are 4ybrobcAly coaf leA
+v %e rd.ador core.
BASES' knad portio ts) BACKGROUND h The _ limit on the boron concentration of the Reactor Coolant
/3ystem (KLd), tAe refueling canal, and the refueling cavity a
\_durino refuelino.fensures heO the re tor remains suberitical during MODE 6. Refueling boron concentration is
@ .f the these soluble boron velc;.es having concentration in the direct :::::: to th /=cl:nt rec ter"==chre of TJ idur'n; refec? hg.f Lis mova m p.a o m a.,,, g The soluble boron concentration offsets the core reactivity and is measured by chemical analysis of a representative sample of the coolant in each of the volumes. The refueling poronconcentrationlimitisspecifiedintheCOLR. GMFf acecu e: enor e specified boron concentration 'i4Mr
" to maintain an overall core reactivity of kg s 0.9S during fuel handling, with control rods and fuel assemblies assumed Q to be in the most adverse configuration h reactivity),hlla::d h ?!=t ;r=edcre d(least negative 2 x)
GDC 26 of 10 CFR 50, Appendix A, requires that two independent reactivity control systems of different design principles be provided (Ref.1). One of these systems must be capable of holding the reactor core subcritical under cold conditions. The Chemical and Volume Control System (CVCS) is the system capable of maintaining the reactor subtritical in. cold conditions by maintaining the boron concentration. t (f M.2)} i
~ The reactor is brought to shutdown conditions before beginning operations to open the reactor vessel for refueling.
the ves.sel head is unbolted ( had th:After
,, s;ca thet] removed.
RCS is_ cooled@ and de p % . 9 - . _ , , .y - , . 9 The e f= li= c:=
'5 and the) ref'Jeling cavity @ then flooded with borated water from the refueling water storage tank through the open reactor vessel by gravity feeding or by the use of the Residual Heat Removal (RHR) System pumps.
The pumping action of the RHR System in the RCS and the natural circulation due to thermal driving heads in the_ reactor vessel and refue' ling cavityjhix th; :.dd:d 1 tc-en c e n t r: t e d bont-eew-w+t&t h e = t e r--h-t h e r e f = l i n g i r~ $> ( (,N) f enmre adeguate. 4*iq of 4h kauted W- , { (continued) ' WOG STS B 3.9-1 Rev 1, 04/07/95 Ri:V E L _ _ - _ _ -
Baron Concentration B 3.9.1 BASES BACKGROUND tral ) The RHR System is in operation during refueling (see (continued) LC0 3.9.5, " Residual Heat Removal (RHR) and Coolant Circulation-High Water Level," and LC0 3.9.6, " Residual h Heat Removal (RHR) and Coolant Circulation-Low Water Level")'to provide forced circulation in the RCS and assist j inmaintainingtheboronconcentratiogintheRCS,the refueling canal, and the refueling cavity above the COLR I limit. j l l APPLICABLE During refueling operations, the reactivity condition of the l SAFETY ANALYSES core is-consistent with the initial conditions' assumed for
.the boron dilution accident in the accident analysis %nd is MN
@ conservative for MODE 6. The boron concentration limit .l specified in the COLR is based on the core reactivity at the i beginning of each fuel cycle (the end of refueling) and includes an uncertainty allowance.
1 The required boron concentration and the plant refuel.ing ) procedures that verify the correct fuel loading plan j p (including full core mapping). ensure that the k g of the ! Q' @v core will remain s 0.95 during the refueling oper,ation. 44 : Hence, at least a 5%Ak/k margin of safety is established during refueling 3J. oc e, Op, in.p;nd pg)
. Durina refueling,_stne tog Wwater Rc4volume in the spent fuel pool, m j g- g> the transferM, the refueling canal, the refueling g Of. cavity, and the reactor vessel for.m a single mass. As a b result, the soluble boron concentration is relatively the
, same in each of these volumes.
boron dilution accident analyzed occurs in o The limiting $ . A detailed discussion of this event is
@ W MODE
.provided in Bases B 3.1
- < :CC m. "
5 (Ref.
" SHUTDOWN MARGIN (SDM)G -i-;.
Jhe RCS boron concentration satisfies Criterion 2 of C_c "d Molia =:nr]. '
"locfG $O.%CcQfid i
att 4dak porHon oC LC0 The LC0 requi es that a minimum boron concentration be maintained in the RCS, the refueling canal, and the p - refueling cavit while in MODE 6. The baron concentration limit specified in the COLR ensures that a core gk , of i ( ( ,hv are by(rnbcally ceupleA 4eh reuk cart,] , (continued) WOG STS B 3.9-2 Rev 1, 04/07/95 kO/I o
I 1 i Boron Concentration ! B 3.9.1 j n
' BASES i
i, LCO. s 0.95~is maintained during fuel handling operations. (continued) ' Violation of the LCO could lead to an inadvertent criticality during MODE 6. APPLICABILITY This LCO is applicable in MODE 6 to ensure that the fuel in the reactor vessel will remain subcritical._ The required-baron concentration ensures a k,,, s 0.95. f ^.bric ".00E 5, It4W rLcc 3.1.1, " MU :uem ""c F ge".) , 200
- F," =d j
BSA-54 1CC 2. : . 2. "5 HUT 000" "90!" (50")' ' TM, s: 200"r," enture that an adequate amount of negative reactivity is available Pg to shut down the reactor and maintain it subcritical.
' (rnaer 6 3.9-39 ---> h ACTIONS g A.I Vk !
Continuation of CORE ALTERATIONS or positive reactivity " additions (including actions to reduce baron concentration) g is contingent upon maintaining the unit in compliance with N (3 U the-LCO. If the boron concentration of any coolant volume 3
-kJ g g g ,y , , ,p >T in.the RCS, the refueling canal, or the refueling cavity is #
less than its limit, 11 operations involving CORE - ALTERATIONS d[P positi reactivity additions must be
- h- suspended mmediately. gcgT T Suspension of CORE ALTERATIONS and positive reactivity 4 additions shall not preclude moving a component to a safe position. [
I dr norweeiw lcoofdeen e e4 4he ecofant volums k lor
. e purpose of sp+em +empora.cc c.,,wt.
In addition to immediately suspending CORE ALTERATIONS positive reactivity additions, LeermenMo restore the* concentrationmustbeinitiatedimmediately.]*eiog berord Crb.co. are e c4My mal 4H AM*'P 4 Nh V Or deter-4ain the ecu* red combiaatie9 0f boration- flow Nd rate and concentration ne unicue Occier Sasir EventYfnust be satisfied. The only requirement is to restore the boron concentration.to its required value as soon as possible. In L order to. raise the boron concentration as soon as possible, the operator should begin boration with the best source available for unit conditions. 7y q) - (cont' WOG STS B 3.9-3 Rev 1, O. (-
- - l
- e. . . # ' BASES' INSERT (S)~
f]v. SECTION 3.9 Bases 3.9.1 INSERT B 3.9 3A (Pw) In MODES 1-and 2 with k rr a 1.0. LCO 3.1.4. " Rod Group Alignment Limits." 4LC0 3.1.5. " Shutdown Ba,nk. Insertion Limits." ana LCO 3.1.6. " Control Bank
.. Insertion' Limits." ensure an adequate amount of negative reactivity is available to shutdown the reactor. In MODE 2 with k,,, < 1.0 and MODES 3. 4 and 5. LCO '3.1.1 " SHUTDOWN ~ MARGIN -(SDM). " ensures INSERT B 3.9 3B ( P,)
an inadvertent.' criticality may. occur'due to an incorrect fuel loading. To minimize the potential .of an inadvertent criticality resulting from a fuel
-loading error.
[') . INSERT B 3.9 3CI (P3 )- The ACTIONS lare modified by- a Note stating that entry into the MODE 6..from MODE 5.is not' permitted while-the LCO is not met. .This is an' exception.to-
. lLCO 3.0.4 and precludes detensioning the head when'.the refueling' boron- .
- concentration . limit specified in.the COLR-'is not met.
L fy p V - 7/21/98 Revision I 5 _ .._..m___.u-__m_ -___._. ___.-_.-___.._.mm.m__ -_ - _ -
m .q
^
Unborated Water Source Isolation Valves B 3.9.2 V B 3.9 REFUELING OPERATIONS B 3.9.2 Unborated. Water Source Isolation Valves BASES
-BACKGROUND During MODE 6 operations, all. isolation valves for reactor' makeup water sources containing unborated water that are 3 connected to the Reactor Coolant System (RCS) must be closed
((cVittB,uM28, to prevent unplanned boron dilution of the reactor coolant. CVf% dH5 The isolation valves $must be secured in the closed position. ask cVt429),6 The Chemical and Volume Control System is capable of supplying borated and unborated water to the RCS through
. h-. 'various_ flow. paths. Since a positive reactivity addition made by reducing the boron concentration is inappropriate during. MODE 6, isolation of all unborated water sources prevents an unplanned boron dilution.
h-(Inser F 6 M-sD b GncoAelled PPL1 CABLE The possibility of an paeveetedoron dilution event A SAFETY? ANALYSES (Ref. 1) occurring during MODE 6 refueling operations is V i precluded by' adherence to this LCO, which requires that
-potential dilution sources be isolated. Closing -. the required valves during refueling operations prevents the flow of unborated water to the filled portion o.f the RCS. ,
The valves are used to isolate unborated water sources. ' These valves. have the potential to indirectly allow dilution i of.the RCS boron concentration in MODE 6. By isolating i L unborated water sources, a safety. analysis for an uncontrolled boron-dilution accident in accordance with the Standard Review Plan (Ref. 2) is not required for MODE 6. l gds 3 f', m" .. "_ c ' * = .= = = = = = = = t = ~ ' ' = = = ' " r * *' ' ' '
- r ' " ' ' -
- m _
QG~ olFp{so *SY(e51Ybb . O # N ** **" ""#*Y
.tCO This LCO requires that flow paths to the RCS from unborated !
water sources be isolated to prevent unplanned boron dilution during MODE 6 and thus avoid a reduction in SDM. l ;
^
N j 6 i
+
g H i IN (f l (continued) WOG STS. ~ B 3.9-5 Rev 1, 04/07/95 Red.T.
- ..e
; BASES-INSERT (S)
W SECTION 3.9
/d .
Bases 3.9.2
': INSERT B 3.9 5A -(P3 )-
.The Refueling Water. Storage Tank'(RWST) is assumed to be a boration source.
With the RWST boron concentration not satisfying these assumptions.~the RWST
; becomes a petential' dilution source and' valves CV112D and CV112E are c considered unborated water source isolation valves. These valves must be
.. secured in the closed position.
l .1 L
+
INSERT B 3.9 5B l- , Deleted in: Revision I. n L: l 9 p; r by ll - 1 L , l L I I. 1e L
..d ri.-
NJ . 7/21/98 Revision I T. f 9 k._ ___ _ _ _ . _ _ . _ _ _ _ _ _ _ . _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _
Unborated Water Source Isolation Valves 6 3.9.2 O V. ' BASES ACTIONS
@( AL 4.Z;and A3 (can+Ef 4)}
Due to the potential of having diluted the boron concentration of the reactor coolant, SR 3;9.1.1
-(verification of baron concentration) must be performed whenever Condition A'is entered to_ demonstrate that the
' required baron ~ concentration exists. The Completion Time of 4 hours is sufficient to obtain and analyze a reactor
@- Lcoolant sample for baron concentration.
SURVEILLANCE .SR 3.9.2.1 Q ( c.vms,us4n,aset,asog awl cVene } 1 m REQUIREMENTS o These valves are to be secured closed to isolate possible qY 4'[' dilution paths. The likelihood of a significant reduction in the boron concentration during MODE 6 operations is g N/ remote due to the large mass of barated water in the f((F et,w.w refueling cavity and the fact that all unborated water y sources are isolated, precluding a dilution. The boron g n tww. vuiuA .f concentration is checked every 72 hours during MODE 6 under y~- j y,%" M " SR 3.9.1.1. This = = : Pl:nrademonstrates that she valves i (8 " d P are4 closed +through a system walkdown. The 31 day Frequency 3g is based on engineering judgment and is considered reasonable in view of other administ tive controls that n - will ensure that the valve opening is an'unlikely EV possibility, p ,g j. (ex.s g REFERENCES 1. h @FSAR, Section "5.2.O.
- 2. NUREG-0800, Section 15.4.6.
l M h WOG STS , B 3.9-7 Rev 1, 04/07/95 g g .
Nuclear Instrumentation B 3.9.3 D
'() BASES ACTIONS 'Oktaka B.2 (continued) probability of a change in core reactivity during this time period.
3.9.3.1
~ SURVEILLANCE SR REQUIREMENTS SR 3.9.3.1 is the performance of a CHANNEL CHECK, which is a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that the two indication channels should be
, consistent with core conditions. Changes in fuel-loading and core geometry can result in significant differences between source- range channels, but each channel should be consistent with.its local conditions.
The Frequency of 12 hours is consistent with the CHANNEL CHECK Frequency specified similarly for the same instruments in LC0 3.3.1. fh qj SR 3.9.3.2 SR 3.9.3.2 is the performance of a CHANNEL CALIBRATION every 18 months. This SR is modified by a Note stating' that neutron detectors are excluded from the CHANNEL CALIBRATION. 2 The CHANNEL CALIBRATION for the source. range neutron flux-monitors consists of obtaining the detector @ n= ;
@4 k t s u siE33iiih discriminator curves, evaluating those curves, and "
comparing the curves to the manufacturer's data. The 18 month Frequency is based on the need to perform this 'd i Surveillance under the conditions that apply during a plant outage. Operating experience has shown these components ff ' usually pass the Surveillance when performed at the .18 month Frequency. REFERENCES h gP. 1, .u f54% Tal,le. 7,5'-A 10 CFR 50, Appendix A, GDC 13, GDC 26, GDC 28, and GDC.29. L h @ @ SAR, Section c n. .'O. /C 4.6 m k.} ' L WOG STS B 3.9-10 Rev 1, 04/07/95 V RGI T L e
Containment Penetrations B 3.9.4 fs U ' BASES SURVEILLANCE SR 3.9.4.1 (continued) REQUIREMENTS
-g- demonstrate that each valve operator has motive powe which~
will ensure that each valve is capable of being close( by an OPERABLE automatic p6ntainment purna "d cuhau3t fiolation i signal. [vh.m. % J
~
The Surveillance is performed every 7 days during CORE ALTERATIONS or movement of irradiated fuel' assemblies within i containment. The Surveillance interval is selected to be commensurate with the normal duration of time to complete fuel handling operations.f A :crccill:n:: before th: : tart 7 f;f refueliaij cperitiens will prc'eid; tu er thre
;;r;;ilhnt: verification; durin; th: ;;1 5 ble ;;ried '--J tFi: LCO.I As such, this Surveillance ensures that a postulated fuel handling accident that releases fission product radioactivity within the containment willfnot result in a releas; af fission product radioactivity to the environment.
p u SR 3 . 9 .' 4 . 2 - P, ' EnynuncJ 5.wy Fe. be, A,.h n ,n Sg #,m , (teureA ! This Surveillance. demonstrates that each containment purge l Q hre e r :urtivalve actuates to its isolation position on
- enn:C 1"ti: tion or en)an actual or simulated high radiation signal. The 18 month Frequency maintains J consistency with other similar JE5f*5 instrumentation and
. val In LCO 3.3.6, the Containment
' Ventilanmburve testing =d Shaus requirements.
S Isolation instrumentation requires a CHANNEL CHECK every 12' hours and a COT every 92 days to pj.
, ensure the channel OPERABILITY during refueling operations. p Every 18 months a CHANNEL CALIBRATION is performed. -The-e.,+-- --+..o s
w .mennnen +w4e so-n -;7at;g :.f e ;,
, 1A mnn+be , deirinj enfi_ig]ig , gn 3 gTqggrorn TFCT MtllTS,
~ SR-3. .;J demonstrates that the isolation time of each h vaive is in accordance with the Inservice Testing Program-g SM3 requirements. These Surveillance performed during MODE 6 oc will ensure that the valves are capable of closing after a
. postulated fuel handling accident to limit 'a release of fission product radioactivity from the containment. i Pa n znsc-ex
(): %JA-ICA (continued) WOG STS B 3.9-15 Rev 1, 04/07/95 h HEY 1 l. e _ _. ___- - _ _ .
RHR and Coolant Circulation-High Water Level l { B 3.9.5 D.n- BASES
-APFLICABLE ,mJuc .un. Thcccf;rc. th: ""' Sy:t = i: retair.:d :: :)
SAFETY ANALYSES Sp :ific;tfor,. / (continued) LCO Only one RHR loop is required for decay heat removal in MODE 6, with th~e water level 123 ft above the top of the
. reactor vessel flangeW Onn :n: """ 1::: 1: recuired to be;
.DPC"aCLEAbecause the volume of water above the reactor '
' vessel flange provides backup clocav hont removal capability.
(1tt-4enso /ne RHR loop 1 be FPERA h ndd n operation 4to d provide: h *"eM [ c.
- a. Removal of decay heat; g
- b. Mixing of borated coolant to minimize the possibil'ity D of criticality; and "
- c. Indication of reactor coolant temperature.
fm An OPERABLE RHR loop includes an RHR pump, a heat exchanger,
- d. valves, piping, instruments, and controls to ensure an OPERABLE flow pathend te d:ter-Me the 1er crd ter;:r:turca .
The ' flow path starts in one of the RCS hot legs and is returned to the RCS cold-legs. 4 The LC0 is modified by a Note that allows the required operating RHR loop to be removed from service for up to I hour per 8 hour period, provided no operations are permitted that would cause a reduction of the RCS boron concentration. Baron concentration reduction is prohibited because uniform concentration distribution cannot be ensured without forced circ 61ation. This permits operations such as co.re mapping or alterations in the vicinity of the reactor vessel hot leg nozzles and RCS to RHR isolation valve ' testing. During this I hour period, decay heat is removed l by natural convection to the large mass of water in the l refueling cavity. l fed rsnig # %4 berated cwt-) APP!ICABILITY, One RHR loop must be OPERABLE and in operation in MODE 6, i with the water level 2 23 ft above the top of the reactor vessel flange, to provide decay heat removal. The 23 ft water level was. selected because it corresponds to the 23 ft O
'V .
(continued) WOG STS- B 3.9-18 Rev 1, 04/07/95 R:aJ r
L RHR and Coolan' Circulation-Low Water Level
-h -
B 3.9.6 BASES 1
. (%n ahi+1o A _
- LCO ' ""'" "~a one loop iiiB H must be in operation in order l(continued) to.prov.ide: ,
1
- a. ' Removal ~of-decay heat; Hso
.b. . Mixing.of borated coolant to minimize the possibility-
-of criticality; and g1 g ( m-c.- Indication of reactor coolant temperature.
. An OPERABLE RHR' loop consists of an RHR pump,'a heat "
exchanger, valves,.. piping, instruments and controls'to ,
, ensure .an OPERAB_LE flow path. n: :: det:r- H: th: let e n d ',
n::::r:t:r:J The flow path starts in.one of the RCS hot '
' legs and is returned to _the RCS cold legs. [2m,, e a 3.e,.nal-O APPLICABILITY.: Two RHR loops 'are required to be OPERABLE, and one RHR loop must'be in operation in MODE 6, with the water level < 23 ft j g
<Neth. ,
J above the ' top of the reactor ' vessel _ flange, _ to provide decay _ heat removal +. Requirements for the RHR System in teHHw , V- ~
'7, W "* */ y g :r ::v: ecd by LCO: " 5 : tier 2.4, Per ter C 1:nt
'.nd
; Suti: 2.5, E :r;:n:', C:r: C::l' : )
*% O
,.,ar,(".C),f,]RHR1 O:t::: 'EC:s cop.requirementsinMODE.6withthe=
A water leve1- 2 Z3 ft are: located in LCO 3.9.5, " Residual Heat Removal (RHR) .and Coolant Circulation-High Water: Level." k. s- ,, > \'k
' g' ' @w' S.Po % =Ldat el '"h~Wf the harA+.A yu %,.3 wtant -%,,Gre ACTIONS A.1 and'A.2 L duay yJ (w% me oors @$f 1 :: (TE,ouse, /N) v -
than the recuired ale ua RHR 1 cop 1 action shall be immediately initiated and continued until
' h. M """ 1;;p i:2 restored to OPERABLE status (ed-+e-eeece%m) or until 2' 23 ft of water level is established above the reactor vessel flange. When the water level is 2 23 ft above the reactor vessel flange, the Applicability changes to that of.LC0 3.9.5, and only one RHR loop is_ required to be OPERABLE and in operation. An immediate Completion Time is necessary'for an operator to initiate corrective actions.
%4 re9 d rei Mer of RHA lect * < reg (continued)
..WOG STS B 3.9-22 Rev 1, 04/07/95 WX
i \ cJ., BASES INSERT (S) jep; SECTION 3.9-v Bases 3 9.6-IINSERTB3i9-22A (Pa) l1'.42, 3,'4, and'5 are covered by LCO 3.4.6. "RCS Loo)s - MODE 4." LCO 3.4.7, "RCS; Loops - MODE 5. Loops Filled.":LCO 3.4.8. "RCS _ oops - MODE 5.; Loops Not:
. Filled " LC0 3.5.2, "ECCS - Operating." and LC0 3.5.3, 'ECCS - Shutdown. "
4 INSERT B~3.9 22B (C 2 and P3)
- However. -the LCO.is modified by a Note that 3ermits-the required RHRL loop to
.be removed from operation and considered OPERABLE when-aligned to, or during transitioning to or from, the' Refueling-Water Storage Tank (RWST) to support
- filling or draining' thel refueling cavity. or~ to support required' testing. 'if capable of'being realigned to the RCS.
[X-)' INSERT B'3.9 22C -(P25 ) The! ACTIONS are modified by a Note stating'that entry:into the' Applicability
'is not permitted while the LC0 is not met. This-is an exception to LCO 3.0.4
- and.precludesitransition into' MODE ~ 6 with water level < 23 ft while the LCO is not; met.
o l' 4 f
=
9 ()l . 7/21/98 Revision I
{hMNhCavity Water
- r. fy N ,l B 3.9 REFUELING OPERATIONS r
B 3.9.7 ~ Refueling Cavity Water Level
}
BASES- # # " BACKGROUND The movement of irradiated fuel assemblie or performance of CORE . ALTERATIONS, except during latching and nlatching of control rod drive shafts, within containment requires a minimum water level lof 23 ft above the top of-the reactor cs .e aw u2 Ressel flan e. L" r'n: reft:li ne. thir -4 a"4 insufficient
)
( u g)._ --r wa t ei" inJthe ce ni- (
; reto 'n; c r :', f;d ?
h, (trrr?:r err.2 refueling cavit1 ead ceat fuel ;::k ( L rrici;r.t .cter i ::c 2ry3to retain iodine fission product activity (1- t ic r:t f i- th :=nt Nt fuel handling accident,(Refs. I and 2). Sufficient iodine crum.co) - g c,nu,gattivity would be retained.to limit offsite dos ~es- from the
. accident to < 25% of 10 CFR-100 limits, as provided by the guidance of Reference'3.
. APPLICABLE During CORE ALTERATIONS and movement of irradiated fuel ~
(;q'f SAFETY ANALYSES assemblies,. the water level in the cutu;;mc cana; a~ tw refueling cavity is an initial. condition design parameter in
-the' analysis of a fuel. handling accident in containment,;as postulated by Regulatory Guide l.25 (Ref.1). : A minimum. .J water level of 23 ft (Regulatory Position C.I.c of Ref.1) 'q ? {
1
' allows a decontamination factor. of 100 (Regulatory 3h l Position C.Lg of Ref.1) to be 'used in the accident *4*
analysis for iodine. This relates to the assumption that 99% of the total iodine released from the pellet to cladding g
% gap of all the dropped fuel assembly rods is retained by the refueling cavity water. The fuel pellet to claddingja3 is ( )
T assumed to contain 10% of the total fuel rod iodine Cane moWe. *C (4 .O. ' era && & inventoryt(Ref. . M" # The fuel handling accident analysis inside containment is t- J described in Reference 2. With a minimum water level of 23 ft and a minimum decay time of 100 hours prior to fuel ,
' handling, the analysis and test programs demonstrate that l the iodine release due to a postulated fuel handling '
accident is adequately captured by the water and offsite doses are' maintained within allowable limits (Refs. 4 1
-and 5). ' ~V (continued) 1 WOG STS. B 3.9-25 -
Rev 1, 04/07/95 Rev I
i P 1JUSTIFICATIONFORDIFFERENCESTONUREG1431 BASES-f~y .SECTION 3.9 REFUELING OPERATIONS'- A/
~
!BRACKETEDCHANGES(B) , )
B{ .The brack'ets were. removed and the plant specific value was-added.
. B, - .LThe brackets ;were removed and the_ plant specific reference was added.
GENERIC CHANGES-(C)- 1
.Cp - This change'is consistent with-NUREG-1431, as modified by. TSTF-20-(NRC
- Approved).
~
Cf - This change is consistent with NUREG-1431, as; modified by TSTF-21.
- Revision 1.
lC.3' This change to.
Reference:
10 CFR 50.36(c)(2)(ii), is consistent with
- NUREG-1431, 'as modified by ~an editorial change submitted to and: approved
~~
by the.NRC, Ci" LThis change -is consistent with NUREG-1431, as modified by. TSTF-96'(NRC _ Approved).
;w l'
-C 3- LThisichange.is consistent with NUREG-1431. 'as modified by TSTF-23.
, [j"i Revision 1 (NRC. Approved). The appropriate bracketed-information for.
~ Byron and Braidwood is utilized- .
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;s:
(96 ,)F BYRON /BRAIDWOOD - UNITS 1 & 2 3.9 1 ~ 7/17/98 Revision I
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W S- ~ JUSTIFICATION FOR DIFFERENCES T0'NUREG 1431 BASES
.SECTION 3.9 REFUELING OPERATIONS- l Q)4 .m- ~
. 1 The
Background:
and Surveillance Requirements sections of the Bases for-4l%Pg.' 'ITS.LC0 3.9.2 were revised'to include the valve designations for those 1 J
-valves which must be verified closed in order to ~ ensure unborated water
. sources are isolated. This change is consistent with the current clicensing basis as presented in CTS SR 4.9.1.3. CTS LCO 3.1.2.5 and LCO 3.1.2.6. 'This' is an editorial' enhancement only and .dces not. involve a technical issue.
% P. The Surveillance Requirements sections of the Bases for.ITS LCO 3.9.2..
- 6 .were. revised to include the methods which can be used to-verify that the.
% unborated water. source isolation valves are closed and secured in 4 -position. This change is consistent with the current licensing basis as presented in CTS SR 4.9,1.3. This is an editorial enhancement only and does not involve a technical issue.
P, The Bases for LC0 3.9.1 was revised to clarify the:descri) tion of the.
. refueling operation since the LCO is applicable only to t1e filled portions of the RCS. the refueling canal.- and the refueling cavity that
.are' hydraulically coupled to the reactor core. The boron concentration
.is' established to ensure that the kg of the reactor core remains s 0.95
'during refueling operations. This infers that the boron concentration
. limit is . applicable'only for the reactor core or those areas in direct communication with the reactor core. -This-is an editorial enhancement
.h only and'does.not-involve a. technical change.
P, The Background section of the Bases for NUREG LC0 3.9.3 was revised to
-indicate that the use of portable ~ detectors is allowed. This change.is-consistent with. current _ practice and with NUREG-1430 and NUREG-1432.
- P, The Ba'ckground section of the Bases for:ITS LC0 3.9.-4 was revised to clarify specific design features and system configurations. This is an
. editorial enhancement only and does not involve a technical issue.
Pg Not used. Pii .The Background section of the Bases for ITS.LC0 3.9.4 was revised to
. clarify the specific provisions used in approving equivalent isolation methods. This change eliminates the potential for misinterpretation of
.the requirements for approving'these methods and provides consistency.in the application and understanding of these requirements. The inappropriate reference to a GPU Safety Evaluation..is also omitted.
'This is an editorial enhancement'only and does-not involve'a technical Eissue.
1/7 Q ' BYRON /BRAIDWOOD UNITS 1 & 2 3,9 3 7/17/98 Revision I p' L
JUSTIFICATION FOR DIFFERENCES TO NUREG 1431 BASES ( SECTION 3.9 REFUELING OPERATIONS P 23 The Applicable' Safety: Analyses section of the Bases for- NUREG 3.9.2 ! incorrectly refers to the RCS boron concentration (which is addressed in i LCO 3.9.1. E The statement-is revised to refer to the unborated water ~ source: isolation valves which are the subject of LC0 3.9.2. i
~
- P,c The Bases. for LCO 3.9.6 is revised to reflect changes made to LC0 3.9.6 Specifically. the Note to SR. 3.9.6.1, added by TSTF-21. Revision 1 is :
-positioned followinglthe LC0'and is reformatted consistent with other-LC0 notes. -This is an editorial enhancement only.and does not involve any technical changes.
P,3 This Note was added for consistency and compliance with NUREG-1431. l LCO 3.0.4. P
~ 26' The Bases-for SR 3.9.4.2 is revised by deleting the: sentence. "The
. system actuation response time is demonstrated every 18 months. during refueling, on a STAGGERED TEST BASIS." These valves will continued to.
be tested on an 18 month frequency but our current licensing basis does not' allow the flexibility of a STAGGERED TEST BASIS. This change'is consistent with Byron and Braidwood current licensing basis. D v i i I L ,
,V BYRON /BRAIDWOOD UNITS 1 & 2 3.9 6 7/1.7/98 Revision I
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